WorldWideScience

Sample records for reactor internals aging

  1. Boiling-Water Reactor internals aging degradation study. Phase 1

    Energy Technology Data Exchange (ETDEWEB)

    Luk, K.H. [Oak Ridge National Lab., TN (United States)

    1993-09-01

    This report documents the results of an aging assessment study for boiling water reactor (BWR) internals. Major stressors for BWR internals are related to unsteady hydrodynamic forces generated by the primary coolant flow in the reactor vessel. Welding and cold-working, dissolved oxygen and impurities in the coolant, applied loads and exposures to fast neutron fluxes are other important stressors. Based on results of a component failure information survey, stress corrosion cracking (SCC) and fatigue are identified as the two major aging-related degradation mechanisms for BWR internals. Significant reported failures include SCC in jet-pump holddown beams, in-core neutron flux monitor dry tubes and core spray spargers. Fatigue failures were detected in feedwater spargers. The implementation of a plant Hydrogen Water Chemistry (HWC) program is considered as a promising method for controlling SCC problems in BWR. More operating data are needed to evaluate its effectiveness for internal components. Long-term fast neutron irradiation effects and high-cycle fatigue in a corrosive environment are uncertainty factors in the aging assessment process. BWR internals are examined by visual inspections and the method is access limited. The presence of a large water gap and an absence of ex-core neutron flux monitors may handicap the use of advanced inspection methods, such as neutron noise vibration measurements, for BWR.

  2. Pressurized-water reactor internals aging degradation study. Phase 1

    Energy Technology Data Exchange (ETDEWEB)

    Luk, K.H. [Oak Ridge National Lab., TN (United States)

    1993-09-01

    This report documents the results of a Phase I study on the effects of aging degradations on pr internals. Primary stressers for internals an generated by the primary coolant flow in the they include unsteady hydrodynamic forces and pump-generated pressure pulsations. Other stressors are applied loads, manufacturing processes, impurities in the coolant and exposures to fast neutron fluxes. A survey of reported aging-related failure information indicates that fatigue, stress corrosion cracking (SCC) and mechanical wear are the three major aging-related degradation mechanisms for PWR internals. Significant reported failures include thermal shield flow-induced vibration problems, SCC in guide tube support pins and core support structure bolts, fatigue-induced core baffle water-jet impingement problems and excess wear in flux thimbles. Many of the reported problems have been resolved by accepted engineering practices. Uncertainties remain in the assessment of long-term neutron irradiation effects and environmental factors in high-cycle fatigue failures. Reactor internals are examined by visual inspections and the technique is access limited. Improved inspection methods, especially one with an early failure detection capability, can enhance the safety and efficiency of reactor operations.

  3. 77 FR 23513 - Updated Aging Management Criteria for Reactor Vessel Internal Components of Pressurized Water...

    Science.gov (United States)

    2012-04-19

    ... COMMISSION Updated Aging Management Criteria for Reactor Vessel Internal Components of Pressurized Water... Management Criteria for PWR Reactor Vessel Internal Components.'' The original notice provided the ADAMS... published a notice requesting public comments on draft LR-ISG-2011-04, ``Updated Aging Management...

  4. 77 FR 16270 - Updated Aging Management Criteria for Reactor Vessel Internal Components of Pressurized Water...

    Science.gov (United States)

    2012-03-20

    ... COMMISSION Updated Aging Management Criteria for Reactor Vessel Internal Components of Pressurized Water... license renewal interim staff guidance (LR-ISG), LR-ISG-2011-04, ``Updated Aging Management Criteria for... Aging Lessons Learned (GALL) Report for the aging management of stainless steel structures...

  5. Structural integrity and management of aging in internal components of BWR reactors; Integridad estructural y manejo del envejecimiento en componentes internos de reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Arganis J, C.R. [Instituto Nacional de Investigaciones Nucleares, Km 36.5 Carretera Mexico, Toluca Salazar Edo. de Mexico (Mexico)]. E-mail: craj@nuclear.inin.mx

    2004-07-01

    Presently work the bases to apply structural integrity and the handling of the aging of internal components of the pressure vessel of boiling water reactors of water are revised and is carried out an example of structural integrity in the horizontal welding H4 of the encircling one of the core of a reactor, taking data reported in the literature. It is also revised what is required to carry out the handling program or conduct of the aging (AMP). (Author)

  6. Semi-Annual Report on Work Supporting the International Forum for Reactor Aging Management (IFRAM)

    Energy Technology Data Exchange (ETDEWEB)

    Bond, Leonard J.; Brenchley, David L.

    2011-11-30

    During the first six months of this project, Pacific Northwest National Laboratory has provided planning and leadership support for the establishment of the International Forum for Reactor Aging Management (IFRAM). This entailed facilitating the efforts of the Global Steering Committee to prepare the charter, operating guidelines, and other documents for IFRAM. It also included making plans for the Inaugural meeting and facilitating its success. This meeting was held on August 4 5, 2011, in Colorado Springs, Colorado. Representatives from Asia, Europe, and the United States met to share information on reactor aging management and to make plans for the future. Professor Tetsuo Shoji was elected chairperson of the Leadership Council. This kick-off event transformed the dream of an international forum into a reality. On August 4-5, 2011, IFRAM began to achieve its mission. The work completed successfully during this period was built upon important previous efforts. This included the development of a proposal for establishing IFRAM and engaging experts in Asia and Europe. The proposal was presented at Engagement workshops in Seoul, Korea (October 2009) and Petten, The Netherlands (May 2010). Participants in both groups demonstrated strong interest in the establishment of IFRAM. Therefore, the Global Steering Committee was formed to plan and carry out the start-up of IFRAM in 2011. This report builds on the initial activities and documents the results of activities over the last six months.

  7. Proceedings of the Inaugural Meeting of the International Forum for Reactor Aging Management (IFRAM)

    Energy Technology Data Exchange (ETDEWEB)

    Bond, Leonard J.; Brenchley, David L.

    2011-09-01

    In almost all countries with nuclear power plants (NPPs), regulatory authorities and the nuclear industry are looking at some form of extended operating periods. To support life extension activities it is necessary to ensure the continued safety and reliability of system, structures, and components, and the component materials. Internationally, a variety of individual national and international activities have been initiated including Plant Life Management through the International Atomic Energy Agency, Electric Power Research Institute’s Long Term Operation program, and various national programs in managing materials degradation and related topics. The U.S. Nuclear Regulatory Commission (NRC) engaged the international community in workshops in 2005-2006 to identify research needs and to collect information in an expert panel report on Proactive Management of Materials Degradation (PMMD), which was reported in NUREG/CR-6923. These results are also available via an Information Tool on the internet at http://pmmd.pnl.gov. This information builds on the extensive compilations known as the GALL Report (Generic Aging Lessons Learned, NUREG-1801, Vols. 1 and 2). Pacific Northwest National Laboratory (PNNL) recently issued a report on the review of various international activities in PMMD (PNNL-17779). There have also been initiatives by Electricite de France, Tokyo Electric Power Company, EPRI, and others to establish a "Materials Aging Institute." Within the materials degradation research community there are also networks and technical meetings focused on some elements of PMMD. In spite of all these efforts, there is currently no forum to bring together these diverse activities and provide coordinated information exchange and prioritization of materials aging management/PMMD topics. It is believed that the International Forum for Reactor Aging Management (IFRAM) would be a good way to achieve this goal and help develop new approaches for ensuring continued safe

  8. A study on the fault diagnostic techniques for reactor internal structures using neutron noise analysis

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Tae Ryong; Jeong, Seong Ho; Park, Jin Ho; Park, Jin Suk [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-08-01

    The unfavorable phenomena, such as flow induced vibration and aging process in reactor internals, cause degradation of structural integrity and may result in loosing some mechanical binding components which might impact other equipments and components or cause flow blockage. Since these malfunctions and potential failures change reactor noise signal, it is necessary to analyze reactor noise signal for early fault diagnosis in the point of few of safety and plant economics. The objectives of this study are to establish fault diagnostic and TS(thermal shield), and to develop a data acquisition and signal processing software system. In the first year of this study, an analysis technique for the reactor internal vibration using the reactor noise was proposed. With the technique proposed and the reactor noise signals (ex-core neutron and acceleration), the dynamic characteristics of Ulchin-1 reactor internals were obtained, and compared with those of Tricastin-1 which is the prototype of Ulchin-1. In the second year, a PC-based expert system for reactor internals fault diagnosis is developed, which included data acquisition, signal processing, feature extraction function, and represented diagnostic knowledge by the IF-THEN rule. To know the effect of the faults, the reactor internals of Ulchin-1 is modeled using FEM and simulated with an artificial defect given in the hold-down spring. Trend in the dynamic characteristics of reactor internals is also observed during one fuel cycle to know the effect of boron concentration. 100 figs, 7 tabs, 18 refs. (Author).

  9. Verification of Remote Inspection Techniques for Reactor Internal Structures of Liquid Metal Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Joo, Young Sang; Lee, Jae Han

    2007-02-15

    The reactor internal structures and components of a liquid metal reactor (LMR) are submerged in hot sodium of reactor vessel. The division 3 of ASME code section XI specifies the visual inspection as major in-service inspection (ISI) methods of reactor internal structures and components. Reactor internals of LMR can not be visually examined due to opaque liquid sodium. The under-sodium viewing techniques using an ultrasonic wave should be applied for the visual inspection of reactor internals. Recently, an ultrasonic waveguide sensor with a strip plate has been developed for an application to the under-sodium inspection. In this study, visualization technique, ranging technique and monitoring technique have been suggested for the remote inspection of reactor internals by using the waveguide sensor. The feasibility of these remote inspection techniques using ultrasonic waveguide sensor has been evaluated by an experimental verification.

  10. Verification of Remote Inspection Techniques for Reactor Internal Structures of Liquid Metal Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Joo, Young Sang; Lee, Jae Han

    2007-02-15

    The reactor internal structures and components of a liquid metal reactor (LMR) are submerged in hot sodium of reactor vessel. The division 3 of ASME code section XI specifies the visual inspection as major in-service inspection (ISI) methods of reactor internal structures and components. Reactor internals of LMR can not be visually examined due to opaque liquid sodium. The under-sodium viewing techniques using an ultrasonic wave should be applied for the visual inspection of reactor internals. Recently, an ultrasonic waveguide sensor with a strip plate has been developed for an application to the under-sodium inspection. In this study, visualization technique, ranging technique and monitoring technique have been suggested for the remote inspection of reactor internals by using the waveguide sensor. The feasibility of these remote inspection techniques using ultrasonic waveguide sensor has been evaluated by an experimental verification.

  11. Internal Combustion Engines as Fluidized Bed Reactors

    Science.gov (United States)

    Lavich, Zoe; Taie, Zachary; Menon, Shyam; Beckwith, Walter; Daly, Shane; Halliday, Devin; Hagen, Christopher

    2016-11-01

    Using an internal combustion engine as a chemical reactor could provide high throughput, high chemical conversion efficiency, and reactant/product handling benefits. For processes requiring a solid catalyst, the ability to develop a fluidized bed within the engine cylinder would allow efficient processing of large volumes of fluid. This work examines the fluidization behavior of particles in a cylinder of an internal combustion engine at various engine speeds. For 40 micron silica gel particles in a modified Megatech Mark III transparent combustion engine, calculations indicate that a maximum engine speed of about 60.8 RPM would result in fluidization. At higher speeds, the fluidization behavior is expected to deteriorate. Experiments gave qualitative confirmation of the analytical predictions, as a speed of 48 RPM resulted in fluidized behavior, while a speed of 171 RPM did not. The investigation shows that under certain conditions a fluidized bed can be obtained within an engine cylinder. Corresponding Author.

  12. Ageing management experience at NUR Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Melllal, Sabrina; Rezig, Mohamed; Zamoun, Rachid; Ameur, Azeddin [Nuclear Research Center of Draria, Algiers (Algeria)

    2013-07-01

    NUR is a 1 MW, open pool reactor moderated and cooled by light water. It was commissioned in 1989. NUR is used for education and training in Nuclear Engineering and related topics for COMENA and National Scientific Community. It is also used to perform R and D works and services at national and regional levels. In this presentation, we describe the methodology and the main development activities related to the ageing management at NUR reactor. These activities include inspection actions and development actions to introduce modifications, to solve obsolescence issues in view to implement the required preventive and curative maintenance programs and to improve the performances of the installation. These actions involved mainly the Operation Assistance System of the Reactor (OAS), the secondary cooling loop, the cooling tower. A new OAS using a new technology and having more possibilities than the older one was introduced in the control system of the reactor. The OAS hardware structure, software structure and the main functions performed are presented. The second loop is entirely refurbished. Two new cooling towers are installed and connected to the main heat exchanger with new piping and valves. The architecture of this new installation is described and the performance assessed. Other actions which involve auxiliary systems like emergency electrical system, air pneumatic system and automatic fire extinguishing are presented.

  13. Related activities on management of ageing of Dalat Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Pham Van Lam [Reactor Dept., Nuclear Research Institute, Dalat (Viet Nam)

    1998-10-01

    The Dalat Nuclear Research Reactor (DNRR) is a pool type research reactor which was reconstructed in 1982 from the previous 250 kW TRIGA-MARK II reactor. The reactor core, the control and instrumentation system, the primary and secondary cooling systems as well as other associated systems were newly designed and installed. The renovated reactor reached its initial criticality in November 1983 and attained its nominal power of 500 kW in February 1984. Since then DNRR has been operated safely. Retained structures of the former reactor such as the reactor aluminum tank, the graphite reflector, the thermal column, the horizontal beam tubes and the radiation concrete shielding are 35 years old. During the recent years, in-service inspection has been carried out, the reactor control and instrumentation system were renovated due to ageing and obsolescence of its components, reactor general inspection and refurbishment were performed. Efforts are being made to cope with ageing of old reactor components to maintain safe operation of the DNRR. (author)

  14. Upper internals arrangement for a pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Singleton, Norman R; Altman, David A; Yu, Ching; Rex, James A; Forsyth, David R

    2013-07-09

    In a pressurized water reactor with all of the in-core instrumentation gaining access to the core through the reactor head, each fuel assembly in which the instrumentation is introduced is aligned with an upper internals instrumentation guide-way. In the elevations above the upper internals upper support assembly, the instrumentation is protected and aligned by upper mounted instrumentation columns that are part of the instrumentation guide-way and extend from the upper support assembly towards the reactor head in hue with a corresponding head penetration. The upper mounted instrumentation columns are supported laterally at one end by an upper guide tube and at the other end by the upper support plate.

  15. Remote Inspection Techniques for Reactor Internals of Liquid Metal Reactor by using Ultrasonic Waveguide Sensor

    Energy Technology Data Exchange (ETDEWEB)

    Joo, Young Sang; Kim, Seok Hun; Lee, Jae Han

    2006-02-15

    The primary components such as a reactor core, heat exchangers, pumps and internal structures of a liquid metal reactor (LMR) are submerged in hot sodium of reactor vessel. The division 3 of ASME code section XI specifies the visual inspection and continuous monitoring as major in-service inspection (ISI) methods of reactor internal structures. Reactor core and internal structures of LMR can not be visually examined due to an opaque liquid sodium. The under-sodium viewing and remote inspection techniques by using an ultrasonic wave should be applied for the in-service inspection of reactor internals. The remote inspection techniques using ultrasonic wave have been developed and applied for the visualization and ISI of reactor internals. The under sodium viewing technique has a limitation for the application of LMR due to the high temperature and irradiation environment. In this study, an ultrasonic waveguide sensor with a strip plate has been developed for an application to the under-sodium viewing and remote inspection. The Lamb wave propagation of a waveguide sensor has been analyzed and the zero-order antisymmetric A{sub 0} plate wave was selected as the application mode of the sensor. The A{sub 0} plate wave can be propagated in the dispersive low frequency range by using a liquid wedge clamped to the waveguide. A new technique is presented which is capable of steering the radiation beam angle of a waveguide sensor without a mechanical movement of the sensor assembly. The steering function of the ultrasonic radiation beam can be achieved by a frequency tuning method of the excitation pulse in the dispersive range of the A{sub 0} mode. The technique provides an opportunity to overcome the scanning limitation of a waveguide sensor. The beam steering function has been evaluated by an experimental verification. The ultrasonic C-scanning experiments are performed in water and the feasibility of the ultrasonic waveguide sensor has been verified. The various remote

  16. Degradation of fastener in reactor internal of PWR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, D. W.; Ryu, W. S.; Jang, J. S.; Kim, S. H.; Kim, W. G.; Chung, M. K.; Han, C. H

    2000-03-01

    Main component degraded in reactor internal structure of PWR is fastener such as bolts, stud, cap screw, and pins. The failure of these components may damage nuclear fuel and limits the operation of nuclear reactor. In foreign reactors operated more than 10 years, an increasing number of incidents of degraded thread fasteners have been reported. The degradation of these components impair the integrity of reactor internal structure and limit the life extension of nuclear power plant. To solve the problem of fastener failure, the incidents of failure and main mechanisms should be investigated. the purpose of this state-of-the -art report is to investigate the failure incidents and mechanisms of fastener in foreign and domestic PWR and make a guide to select a proper materials. There is no intent to describe each event in detail in this report. This report covers the failures of fastener and damage mechanisms reported by the licensees of operating nuclear power plants and the applications of plants constructed after 1964. This information is derived from pertinent licensee event report, reportable occurrence reports, operating reactor event memoranda, failure analysis reports, and other relevant documents. (author)

  17. Use of MCNP for characterization of reactor vessel internals waste from decommissioned nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Love, E.F.; Pauley, K.A.; Reid, B.D.

    1995-09-01

    This study describes the use of the Monte Carlo Neutron-Photon (MCNP) code for determining activation levels of irradiated reactor vessel internals hardware. The purpose of the analysis is to produce data for the Department of Energy`s Greater-Than-Class C Low-Level Radioactive Waste Program. An MCNP model was developed to analyze the Yankee Rowe reactor facility. The model incorporates reactor geometry, material compositions, and operating history data acquired from Yankee Atomic Electric Company. In addition to the base activation analysis, parametric studies were performed to determine the sensitivity of activation to specific parameters. A component sampling plan was also developed to validate the model results, although the plan was not implemented. The calculations for the Yankee Rowe reactor predict that only the core baffle and the core support plates will be activated to levels above the Class C limits. The parametric calculations show, however, that the large uncertainties in the material compositions could cause errors in the estimates that could also increase the estimated activation level of the core barrel to above the Class C limits. Extrapolation of the results to other reactor facilities indicates that in addition to the baffle and support plates, core barrels may also be activated to above Class C limits; however the classification will depend on the specific operating conditions of the reactor and the specific material compositions of the metal, as well as the use of allowable concentration averaging practices in packaging and classifying the waste.

  18. Inspection and evaluation guidelines for light water reactor internals

    Energy Technology Data Exchange (ETDEWEB)

    Iizuka, N. [Tokyo Electric Power Co., Inc. (Japan); Taniguchi, K. [Kansai Electric Power Co., Inc., Osaka (Japan); Yoshinaga, T. [Japan Atomic Power Co., Tokyo (Japan)

    2002-12-01

    On February, 2000, in the Engineering Society of Thermal and Nuclear Power Generation, the 'Investigation Group on Inspection and Evaluation Guidelines for Nuclear Reactor Internals' was established. This group was started at moments of some damage cases on reactor internals on BWRs and PWRs in Japan and foreign countries and of finding out cracks based on a number of SCC (stress corrosion cracking) at Inconel alloy weldings of a shroud support of BWR internals in Tsuruga Nuclear Power Station Unit-1 of the Japan Nuclear Power Generation Co., Ltd. on December, 1999. Under these conditions, this group made some guidelines for rational inspection with clear technical foundation, and so on as well as arrangements on structural functions, importance at safety, and so on of the reactor internals, promoted some investigations aiming at wide general proposal on how to carry out future internal inspections on LWR in Japan, and completed almost all of the investigations on March, 2002. Here were described basic indications of the guideline development and summaries of the developed guideline. (G.K.)

  19. Development of the ageing management database of PUSPATI TRIGA reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ramli, Nurhayati, E-mail: nurhayati@nm.gov.my; Tom, Phongsakorn Prak; Husain, Nurfazila; Farid, Mohd Fairus Abd; Ramli, Shaharum [Reactor Technology Centre, Malaysian Nuclear Agency, MOSTI, Bangi, 43000 Kajang, Selangor (Malaysia); Maskin, Mazleha [Science Program, Faculty of Science and Technology, Universiti Kebangsaan Malaysia, Selangor (Malaysia); Adnan, Amirul Syazwan; Abidin, Nurul Husna Zainal [Faculty of Petroleum and Renewable Energy Engineering, Universiti Teknologi Malaysia (Malaysia)

    2016-01-22

    Since its first criticality in 1982, PUSPATI TRIGA Reactor (RTP) has been operated for more than 30 years. As RTP become older, ageing problems have been seen to be the prominent issues. In addressing the ageing issues, an Ageing Management (AgeM) database for managing related ageing matters was systematically developed. This paper presents the development of AgeM database taking into account all RTP major Systems, Structures and Components (SSCs) and ageing mechanism of these SSCs through the system surveillance program.

  20. The Jules Horowitz Reactor - A new High Performance European Material Testing Reactor open to International Users Present Status and Objectives

    Energy Technology Data Exchange (ETDEWEB)

    Iracane, Daniel; Bignan, Gilles [CEA Atomic Energy Commission Saclay Batiment 121- 91191 Gif Sur Yvette (France); Lindbaeck, Jan-Erik; Blomgren, Jan [VATTENFALL AB Nuclear Power Jaemtlandsgatan 99 SE-16287 Stockholm (Sweden)

    2010-07-01

    infrastructure to perform screening, qualification and safety experiments on material and fuel behaviour under irradiation. It is a water-cooled reactor to provide the necessary flexibility and accessibility for managing several highly instrumented experiments, reproducing different reactor environments (water, gas or liquid-metal loops), generating transient regimes (of key importance for safety). The design work of the JHR experimental capacity is driven by identified and expected future experimental needs. Since a few years CEA has started building up a comprehensive scientific workforce with the help of domestic and international partners in order to prepare an up-to-date experimental capacity for JHR. This workforce, gathering a scientist community (young and seniors) is also of primary importance for education and training. One of the way to deal with this topic is to build an International Joint Program as requested by the JHR consortium agreement addressing priorities common to a large community sharing the produced information within a Joint Data Basis. This joint program is called the Jules Horowitz International Programme (JHIP), and has been conceived with the strategic scope to address fuel and materials issues of common interest that are keys for operating plants and future NPP (mainly focused on LWR) and will be implemented with the support of OECD/NEA as a secretariat. The overall objective of the proposed program is to increase the understanding of the mechanisms that govern fuel reliability and safety throughout the entire fuel service time and to assess design improvements aimed at enhancing the flexible, reliable and safe operation of nuclear fuel. Vattenfall decided to join the JHR Consortium in 2008. The strategy Vattenfall has for using the JHR Consortium membership has the ultimate target to mainly support long-term operation of the existing Gen-II reactors and those Gen-III reactors replacing the ageing fleet and meeting a growing energy demand. (authors)

  1. 3. International conference on catalysis in membrane reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-09-01

    The 3. International Conference on Catalysis in Membrane Reactors, Copenhagen, Denmark, is a continuation of the previous conferences held in Villeurbanne 1994 and Moscow 1996 and will deal with the rapid developments taking place within membranes with emphasis on membrane catalysis. The approx. 80 contributions in form of plenary lectures and posters discuss hydrogen production, methane reforming into syngas, selectivity and specificity of various membranes etc. The conference is organised by the Danish Catalytic Society under the Danish Society for Chemical Engineering. (EG)

  2. The International Reactor Physics Experiment Evaluation Project (IRPHEP)

    Energy Technology Data Exchange (ETDEWEB)

    J. Blair Briggs; Enrico Sartori; Lori Scott

    2006-09-01

    Since the beginning of the Nuclear Power industry, numerous experiments concerned with nuclear energy and technology have been performed at different research laboratories, worldwide. These experiments required a large investment in terms of infrastructure, expertise, and cost; however, many were performed without a high degree of attention to archival of results for future use. The degree and quality of documentation varies greatly. There is an urgent need to preserve integral reactor physics experimental data, including measurement methods, techniques, and separate or special effects data for nuclear energy and technology applications and the knowledge and competence contained therein. If the data are compromised, it is unlikely that any of these experiments will be repeated again in the future. The International Reactor Physics Evaluation Project (IRPhEP) was initiated, as a pilot activity in 1999 by the by the Organization of Economic Cooperation and Development (OECD) Nuclear Energy Agency (NEA) Nuclear Science Committee (NSC). The project was endorsed as an official activity of the NSC in June of 2003. The purpose of the IRPhEP is to provide an extensively peer reviewed set of reactor physics related integral benchmark data that can be used by reactor designers and safety analysts to validate the analytical tools used to design next generation reactors and establish the safety basis for operation of these reactors. A short history of the IRPhEP is presented and its purposes are discussed in this paper. Accomplishments of the IRPhEP, including the first publication of the IRPhEP Handbook, are highlighted and the future of the project outlined.

  3. Prevention of ageing of research reactors by design

    Energy Technology Data Exchange (ETDEWEB)

    Boado, J. [INVAP, Bariloche (Argentina); Lolich, J. [INVAP, Bariloche (Argentina)

    1995-12-31

    However, it is our experience as designers and builders of research reactors, that the most important cause of ageing of any experimental installation, is the loss of motivation of the personnel involved in the operation and maintenance, when the objectives for the utilisation of the facility change or research programs are abandoned for whatever reason. We therefore have endeavoured to design and construct research reactors with several engineering features and with an untraditional approach to the training of the future operator of the facility. During all phases of the design and construction of the reactor, we develop in the future operator of the facility the capacity, not only to operate it properly, but to innovate and to adapt the installation to the daily operating problems due to new requirements and options that might not have been foreseen when the facility was ordered. The versatility of the operator is thus a further guarantee against ageing by abandonment. (orig./HP)

  4. 8th International School of Fusion Reactor Technology "Ettore Majorana"

    CERN Document Server

    Leotta, G G; Muon-catalyzed fusion and fusion with polarized nuclei

    1988-01-01

    The International School of Fusion Reactor Technology started its courses 15 years ago and since then has mantained a biennial pace. Generally, each course has developed the subject which was announced in advance at the closing of the previous course. The subject to which the present proceedings refer was chosen in violation of that rule so as to satisfy the recent and diffuse interest in cold fusion among the main European laboratories involved in controlled thermonuclear research (CTR). In the second half of 1986 we started to prepare a workshop aimed at assessing the state of the art and possibly of the perspectives of muon- catalyzed fusion. Research in this field has recently produced exciting experimental results open to important practical applications. We thought it worthwhile to consider also the beneficial effects and problems of the polarization ofthe nuclei in both cold and thermonuclear fusion. In preparing the 8th Course on Fusion Reactor Technology, it was necessary to abandon the tradi...

  5. IAEA designated international centre based on research reactors (ICERR)

    Energy Technology Data Exchange (ETDEWEB)

    Di Tigliole, Andrea Borio; Bradley, Edward; Khoroshev, Mikhail; Marshall, Frances; Morris, Charles; Tozser, Sandor [International Atomic Energy Agency, Vienna (Austria). Dept. of Nuclear Energy

    2016-04-15

    International activities in the back end of the research reactor (RR) fuel cycle have so far been dominated by the programmes of acceptance of highly-enriched uranium (HEU) spent nuclear fuel (SNF) by the country where it was originally enriched. These programmes will soon have achieved their goals. However, the needs of the nuclear community dictate that the majority of the research reactors continues to operate using low enriched uranium (LEU) fuel in order to meet the varied mission objectives. As a result, inventories of LEU SNF will continue to be created and the back end solution of RR SNF remains a critical issue. In view of this fact, the IAEA drew up a report presenting available reprocessing and recycling services for RR SNF.

  6. Aging of reactor vessels in LWR type reactors; Envejecimiento de la vasija y de los internos del nuclear de los reactores tipo LWR

    Energy Technology Data Exchange (ETDEWEB)

    Gomez Briceno, D.; Lapena, J.; Serrano, M.

    2004-07-01

    Most of the degradation mechanisms of nuclear components were not included on the design so they have to be treated a posteriori, and that imply a loss of capacity. In this paper the state of the art on the reactor pressure vessel neutron embrittlement and on the irradiation assisted stress corrosion cracking that affects internal components, are explained. Special attention is devoted on the influence of the neutron fluence on IASCC process, on the material alterations promoted by irradiation and their consequences on the susceptibility to this phenomenon. Regarding the reactor pressure vessel degradation, this paper discuss the application of the Master Curve on the structural integrity evaluation of the vessel. Other aspects related to further developments are also mentioned and the importance of a good materials ageing management on the operation of the plant is pointed out. (Author) 12 refs.

  7. Reactors

    CERN Document Server

    International Electrotechnical Commission. Geneva

    1988-01-01

    This standard applies to the following types of reactors: shunt reactors, current-limiting reactors including neutral-earthing reactors, damping reactors, tuning (filter) reactors, earthing transformers (neutral couplers), arc-suppression reactors, smoothing reactors, with the exception of the following reactors: small reactors with a rating generally less than 2 kvar single-phase and 10 kvar three-phase, reactors for special purposes such as high-frequency line traps or reactors mounted on rolling stock.

  8. Aging study of boiling water reactor high pressure injection systems

    Energy Technology Data Exchange (ETDEWEB)

    Conley, D.A.; Edson, J.L.; Fineman, C.F. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States)

    1995-03-01

    The purpose of high pressure injection systems is to maintain an adequate coolant level in reactor pressure vessels, so that the fuel cladding temperature does not exceed 1,200{degrees}C (2,200{degrees}F), and to permit plant shutdown during a variety of design basis loss-of-coolant accidents. This report presents the results of a study on aging performed for high pressure injection systems of boiling water reactor plants in the United States. The purpose of the study was to identify and evaluate the effects of aging and the effectiveness of testing and maintenance in detecting and mitigating aging degradation. Guidelines from the United States Nuclear Regulatory Commission`s Nuclear Plant Aging Research Program were used in performing the aging study. Review and analysis of the failures reported in databases such as Nuclear Power Experience, Licensee Event Reports, and the Nuclear Plant Reliability Data System, along with plant-specific maintenance records databases, are included in this report to provide the information required to identify aging stressors, failure modes, and failure causes. Several probabilistic risk assessments were reviewed to identify risk-significant components in high pressure injection systems. Testing, maintenance, specific safety issues, and codes and standards are also discussed.

  9. Aging management program of the reactor building concrete at Point Lepreau Generating Station

    Science.gov (United States)

    Aldea, C.-M.; Shenton, B.; Demerchant, M. M.; Gendron, T.

    2011-04-01

    In order for New Brunswick Power Nuclear (NBPN) to control the risks of degradation of the concrete reactor building at the Point Lepreau Generating Station (PLGS) the development of an aging management plan (AMP) was initiated. The intention of this plan was to determine the requirements for specific structural components of concrete of the reactor building that require regular inspection and maintenance to ensure the safe and reliable operation of the plant. The document is currently in draft form and presents an integrated methodology for the application of an AMP for the concrete of the reactor building. The current AMP addresses the reactor building structure and various components, such as joint sealant and liners that are integral to the structure. It does not include internal components housed within the structure. This paper provides background information regarding the document developed and the strategy developed to manage potential degradation of the concrete of the reactor building, as well as specific programs and preventive and corrective maintenance activities initiated.

  10. Aging management program of the reactor building concrete at Point Lepreau Generating Station

    Directory of Open Access Journals (Sweden)

    Gendron T.

    2011-04-01

    Full Text Available In order for New Brunswick Power Nuclear (NBPN to control the risks of degradation of the concrete reactor building at the Point Lepreau Generating Station (PLGS the development of an aging management plan (AMP was initiated. The intention of this plan was to determine the requirements for specific structural components of concrete of the reactor building that require regular inspection and maintenance to ensure the safe and reliable operation of the plant. The document is currently in draft form and presents an integrated methodology for the application of an AMP for the concrete of the reactor building. The current AMP addresses the reactor building structure and various components, such as joint sealant and liners that are integral to the structure. It does not include internal components housed within the structure. This paper provides background information regarding the document developed and the strategy developed to manage potential degradation of the concrete of the reactor building, as well as specific programs and preventive and corrective maintenance activities initiated.

  11. Aging management at G.A. Siwabessy Multipurpose Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jupiter Sitorus Pane; Taswanda Taryo; Alim Tarigan; Alfahari Mardi; Koes Indra

    1998-10-01

    Aging phenomena was raised as an important issue at GAS-MPR after having some component degradation reported within its ten years operational record. This phenomena was anticipated by developing a program of aging management, where its purpose is to prevent an enlargement of degradation of the systems and components and failure. The aging managements include identification of system and component which have aging process enlargement and impact to the reactor operational safety, developing operational operation and maintenance practices, developing operational and maintenance data base, developing predictive maintenance program, and identifying and studying method and technology for inspection, testing, and surveillance. The result of studying on method and technology of testing, surveillance, and diagnostic conclude that FFT based noise spectrum analysis can be used to detect an anomaly of system and component performance. This analysis will be developed in the future using Auto Regressive Method, Kalman Filter, and Neural Network. (author)

  12. SAFETY ANALYSIS METHODOLOGY FOR AGED CANDU® 6 NUCLEAR REACTORS

    Directory of Open Access Journals (Sweden)

    WOLFGANG HARTMANN

    2013-10-01

    Full Text Available This paper deals with the Safety Analysis for CANDU® 6 nuclear reactors as affected by main Heat Transport System (HTS aging. Operational and aging related changes of the HTS throughout its lifetime may lead to restrictions in certain safety system settings and hence some restriction in performance under certain conditions. A step in confirming safe reactor operation is the tracking of relevant data and their corresponding interpretation by the use of appropriate thermalhydraulic analytic models. Safety analyses ranging from the assessment of safety limits associated with the prevention of intermittent fuel sheath dryout for a slow Loss of Regulation (LOR analysis and fission gas release after a fuel failure are summarized. Specifically for fission gas release, the thermalhydraulic analysis for a fresh core and an 11 Effective Full Power Years (EFPY aged core was summarized, leading to the most severe stagnation break sizes for the inlet feeder break and the channel failure time. Associated coolant conditions provide the input data for fuel analyses. Based on the thermalhydraulic data, the fission product inventory under normal operating conditions may be calculated for both fresh and aged cores, and the fission gas release may be evaluated during the transient. This analysis plays a major role in determining possible radiation doses to the public after postulated accidents have occurred.

  13. 1st International School of Fusion Reactor Technology "Ettore Majorana"

    CERN Document Server

    Knoepfel, Heinz; Safety, Environmental Impact and Economic Prospects of Nuclear Fusion

    1990-01-01

    This book contains the lectures and the concluding discussion of the "Seminar on Safety, Environmental Impact, and Economic Prospects of Nuclear Fusion", which was held at Erice, August 6-12, 1989. In selecting the contributions to this 9th meeting held by the International School of Fusion Reactor Technology at the E. Majorana Center for Scientific Cul­ ture in Erice, we tried to provide a comprehensive coverage of the many interre­ lated and interdisciplinary aspects of what ultimately turns out to be the global acceptance criteria of our society with respect to controlled nuclear fusion. Consequently, this edited collection of the papers presented should provide an overview of these issues. We thus hope that this book, with its extensive subject index, will also be of interest and help to nonfusion specialists and, in general, to those who from curiosity or by assignment are required to be informed on these as­ pects of fusion energy.

  14. Aging assessment of PWR (Pressurized Water Reactor) Auxiliary Feedwater Systems

    Energy Technology Data Exchange (ETDEWEB)

    Casada, D.A.

    1988-01-01

    In support of the Nuclear Regulatory Commission's Nuclear Plant Aging Research (NPAR) Program, Oak Ridge National Laboratory is conducting a review of Pressurized Water Reactor Auxiliary Feedwater Systems. Two of the objectives of the NPAR Program are to identify failure modes and causes and identify methods to detect and track degradation. In Phase I of the Auxiliary Feedwater System study, a detailed review of system design and operating and surveillance practices at a reference plant is being conducted to determine failure modes and to provide an indication of the ability of current monitoring methods to detect system degradation. The extent to which current practices are contributing to aging and service wear related degradation is also being assessed. This paper provides a description of the study approach, examples of results, and some interim observations and conclusions. 1 fig., 1 tab.

  15. A novel airlift reactor enhanced by funnel internals and hydrodynamics prediction by the CFD method.

    Science.gov (United States)

    Zhang, Tao; Wei, Chaohai; Feng, Chunhua; Zhu, Jialiang

    2012-01-01

    Airlift reactors have been used widely in many industrial processes, but little work has been conducted on such reactors integrated with internals. In this study, a novel airlift reactor with a funnel internal was developed to achieve better flow conditions and advantages in biological processes. The CFD (computational fluid dynamics) simulation method was employed to investigate the effect of the funnel internals on hydrodynamic properties in the reactor. A CFD model was developed for gas-liquid two-phase flow simulation in a bench-scale reactor. Grid-independent simulation results were verified with global-scale experimental data. The results showed that the local or global gas holdup could be enhanced by 15% and that turbulent kinetic energy could be reduced by a maximum of 7.8% when the superficial gas velocity was 1 cm/s. These features are beneficial for applications in stress-sensitive biological processes.

  16. IGORR-IV -- Proceedings of the fourth meeting of the International Group on Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rosenbalm, K.F. [comp.

    1995-12-31

    The International Group on Research Reactors was formed to facilitate the sharing of knowledge and experience among those institutions and individuals who are actively working to design, build, and promote new research reactors or to make significant upgrades to existing facilities. Twenty-nine papers were presented in five sessions and written versions of the papers or hard copies of the vugraphs used are published in these proceedings. The five sessions were: (1) Operating Research Reactors and Facility Upgrades; (2) Research Reactors in Design and Construction; (3) ANS Closeout Activities; (4) and (5) Research, Development, and Analysis Results.

  17. Segmentation of the internal of the Reactor Jose Cabrera NPP; Segmentacion de los Internos del Reactor CN Jose Cabrera

    Energy Technology Data Exchange (ETDEWEB)

    Rodriguez silva, M.; Borque Linan, J.

    2013-07-01

    The Plan of dismantling and decommissioning of the Jose Cabrera NPP represents the first total dismantling of a nuclear power station in Spain (level 3 of the IAEA). Complete disassembly of the different components of the primary circuit (internal reactor vessel, pusher, generator of steam, etc.) represents a differential activity against previous projects of dismantling The segmentation of the inmates of the reactor under water using tele operators cutting tools in the spent fuel pit, has been a challenge from the technological point of view as well as a critical activity in the framework of the radiological dis-assemblies associated to the Plan of dismantling and Decommissioning of the Jose Cabrera NPP.

  18. International working group on gas-cooled reactors. Summary report

    Energy Technology Data Exchange (ETDEWEB)

    1981-01-15

    The purpose of the meeting was to provide a forum for exchange of information on safety and licensing aspects for gas-cooled reactors in order to provide comprehensive review of the present status and of directions for future applications and development. Contributions were made concerning the operating experience of the Fort St. Vrain (FSV) HTGR Power Plant in the United States of America, the experimental power station Arbeitsgemeinschaft Versuchsreaktor (AVR) in the Federal Republic of Germany, and the CO/sub 2/-cooled reactors in the United Kingdom such as Hunterson B and Hinkley Point B. The experience gained at each of these reactors has proved the high safety potential of Gas-cooled Reactor Power Plants.

  19. Review of the International Atomic Energy Agency International database on reactor pressure vessel materials and US Nuclear Regulatory Commission/Oak Ridge National Laboratory embrittlement data base

    Energy Technology Data Exchange (ETDEWEB)

    Wang, J.A.; Kam, F.B.K.

    1998-02-01

    The International Atomic Energy Agency (IAEA) has supported neutron radiation effects information exchange through meetings and conferences since the mid-1960s. Through an International Working Group on Reliability of Reactor Pressure Components, information exchange and research activities were fostered through the Coordinated Research Program (CRP) sponsored by the IAEA. The final CRP meeting was held in November 1993, where it was recommended that the IAEA coordinate the development of an International Database on Reactor Pressure Vessel Material (IDRPVM) as the first step in generating an International Database on Aging Management. The purpose of this study was to provide special technical assistance to the NRC in monitoring and evaluating the IAEA activities in developing the IAEA IDRPVM, and to compare the IDRPVM with the Nuclear Regulatory Commission (NRC) - Oak Ridge National Laboratory (ORNL) Power Reactor Embrittlement Data Base (PR-EDB) and provide recommendations for improving the PR-EDB. A first test version of the IDRPVM was distributed at the First Meeting of Liaison Officers to the IAEA IDRPVM, in November 1996. No power reactor surveillance data were included in this version; the testing data were mainly from CRP Phase III data. Therefore, because of insufficient data and a lack of power reactor surveillance data received from the IAEA IDRPVM, the comparison is made based only on the structure of the IDRPVM. In general, the IDRPVM and the EDB have very similar data structure and data format. One anticipates that because the IDRPVM data will be collected from so many different sources, quality assurance of the data will be a difficult task. The consistency of experimental test results will be an important issue. A very wide spectrum of material characteristics of RPV steels and irradiation environments exists among the various countries. Hence the development of embrittlement prediction models will be a formidable task. 4 refs., 2 figs., 4 tabs.

  20. Ageing investigation and upgrading of components/systems of Kartini research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Syarip; Widi Setiawan [Yogyakarta Nuclear Research Centre, Yogyakarta (Indonesia)

    1998-10-01

    Kartini research reactor has been operated in good condition and has demonstrated successful operation for the past 18 years, utilized for: reactor kinetic and control studies, instrumentation tests, neutronic and thermohydraulic studies, routine neutron activation analysis, reactor safety studies, training for research reactor operators and supervisors, and reactor physics experiments. Several components of Kartini reactor use components from the abandoned IRT-2000 Project at Serpong and from Bandung Reactor Centre such as: reactor tank, reactor core, heat exchanger, motor blower for ventilation system, fuel elements, etc. To maintain a good operating performance and also for aging investigation purposes, the component failure data collection has been done. The method used is based on the Manual on Reliability Data Collection For Research Reactor PSAs, IAEA TECDOC 636, and analyzed by using Data Entry System (DES) computer code. Analysis result shows that the components/systems failure rate of Kartini reactor is around 1,5.10{sup -4} up to 2,8.10{sup -4} per hour, these values are within the ranges of the values indicated in IAEA TECDOC 478. Whereas from the analysis of irradiation history shows that the neutron fluence of fuel element with highest burn-up (2,05 gram U-235 in average) is around 1.04.10{sup 16} n Cm{sup -2} and this value is still far below its limiting value. Some reactor components/systems have been replaced and upgraded such as heat exchanger, instrumentation and control system (ICS), etc. The new reactor ICS was installed in 1994 which is designed as a distributed structure by using microprocessor based systems and bus system technology. The characteristic and operating performance of the new reactor ICS, as well as the operation history and improvement of the Kartini research reactor is presented. (J.P.N.)

  1. Influence of external circulation on sludge characteristics during start-up of internal circulation reactor

    Institute of Scientific and Technical Information of China (English)

    DING Jian-nan; WANG Dian-zuo

    2005-01-01

    The start-up of external circulation-added internal circulation(IC) reactor was finished in 26 d, 32 d fewer than that of IC reactor. To evaluate the influence of the added external circulation on the development of granular sludge, the characteristics of the granular sludge taken from the two tested laboratory-scale reactors during start-up were studied. The results show that the added external circulation can enhance biomass granulation, accelerate granule development and improve sludge characteristics. At the end of start-up, the granular size of sludge in external circulation-added IC reactor greatly increases with a size distribution much better than that of sludge in IC reactor. The granular sludge originated from external circulation-added IC reactor contains more extracellular polymers and has a greater settling velocity than that from IC reactor. Methanogenic activity of the granular sludge from the external circulation-added IC reactor started 26 d ago reaches 358.23 mL·g-1·d-1, 1.66 and 1.20 times as great as that of the sludge from the IC reactor started 26 d and 58 d ago respectively.

  2. Ageing management and refurbishment of Ghana Research Reactor-1 (GHARR-1)

    Energy Technology Data Exchange (ETDEWEB)

    Amponsahabu, Edward Oscar; Gbadago, Joseph Korbla; Addo, Moses Ankamah; Sogbadji, Robert Bright Mawuko; Odoi, Henry Cecil; Gyamfi, Kwame; Ampong, Atta Gyekye; Opate, Nicholas Sackitey [Ghana Atomic Energy Commission, Accra (Ghana)

    2013-07-01

    Ageing management is an essential component of the routine practices at the Ghana Research Reactor-1 (GHARR-1) Facility. The reactor is Miniature Neutron Source Reactor with a rated power of 30 kW. GHARR-1 was installed and attained criticality on December 17, 1994 and commissioned on 8th March, 1995. It has since been in operation. The routine practices and operational procedures have been set out with clear emphasis on ageing policy at the facility. Some electronic components are changed regularly during maintenance sessions and keeping to regular purification of the reactor and pool water to mitigate against corrosion. This paper outlines the ageing management programme, mitigation practices, strategies for ageing management, periodic safety reviews, consideration of ageing during designing, design features for components and unit replacement, top beryllium shim addition, and succession planning. Information sharing with other operating organization is one of the means considered by GHARR-1 to attain excellence.

  3. Exposure conditions of reactor internals of Rovno VVER-440 NPP units 1 and 2

    Energy Technology Data Exchange (ETDEWEB)

    Grytsenko, O.V.; Pugach, S.M.; Diemokhin, V.L.; Bukanov, V.N. [Inst. for Nuclear Research, Kyiv, 03680 (Ukraine); Marek, M.; Vandlik, S. [Nuclear Research Inst. Rez Plc., Rez, 25068 (Czech Republic)

    2011-07-01

    Results of determination of irradiation conditions for vessel internals of VVER-440 reactor No. 1 and 2 at Rovno Nuclear Power Plant, obtained by specialists at Inst. for Nuclear Research Kyiv (Ukraine)), and Nuclear Research Inst. Rez (Czech Republic)), are presented. To calculate neutron transport, detailed calculation models of these reactors were prepared. Distribution of neutron flux functionals on the surface of reactor VVER-440 baffle and core barrel for different core loads was studied. Agreement between results obtained by specialists at Inst. for Nuclear Research and at Nuclear Research Inst. is shown. (authors)

  4. U.S. Department of Energy Program of International Technical Cooperation for Research Reactor Utilization

    Energy Technology Data Exchange (ETDEWEB)

    Chong, D.; Manning, M.; Ellis, R.; Apt, K.; Flaim, S.; Sylvester, K.

    2004-10-03

    The U.S. Department of Energy, National Nuclear Security Administration (DOE/NNSA) has initiated collaborations with the national nuclear authorities of Egypt, Peru, and Romania for the purpose of advancing the commercial potential and utilization of their respective research reactors. Under its Office of International Safeguards ''Sister Laboratory'' program, DOE/NNSA has undertaken numerous technical collaborations over the past decade intended to promote peaceful applications of nuclear technology. Among these has been technical assistance in research reactor applications, such as neutron activation analysis, nuclear analysis, reactor physics, and medical radioisotope production. The current collaborations are intended to provide the subject countries with a methodology for greater commercialization of research reactor products and services. Our primary goal is the transfer of knowledge, both in administrative and technical issues, needed for the establishment of an effective business plan and utilization strategy for the continued operation of the countries' research reactors. Technical consultation, cooperation, and the information transfer provided are related to: identification, evaluation, and assessment of current research reactor capabilities for products and services; identification of opportunities for technical upgrades for new or expanded products and services; advice and consultation on research reactor upgrades and technical modifications; characterization of markets for reactor products and services; identification of competition and estimation of potential for market penetration; integration of technical constraints; estimation of cash flow streams; and case studies.

  5. The Design of Reactor Internals Hold-Down Spring

    Directory of Open Access Journals (Sweden)

    Guohong Xue

    2016-01-01

    Full Text Available Hold down spring(HDS, clamped between the upper support plate flange and the core barrel flange inside a pressurized reactor vessel, is used to provide a downward force to keep the core barrel stable and from being lifted off from the vessel ledge during reactor normal operation. Spring designer used to think under certain extreme operating conditions, the spring could be lifted off from the ledge because the spring is not stiff enough to prevent the “lift-off”. Therefore, the spring was designed as stiff as it practically can be. However, finite element study indicated that the magnitude of preload is a strong function of friction coefficient. To find the magnitude of the friction coefficient, a series of tests were conducted on 1/10 scale hold down spring samples of three different spring designs. From the test results, it was found that the friction coefficient increases rapidly as the number of loading cycles increases. This implies that the spring preload would increase rapidly during plant operation. Therefore, it is concluded that, because of the strong frictional effect, not only the current design is more than adequate, but also should be made even softer.

  6. Design of a management information system for the Shielding Experimental Reactor ageing management

    Energy Technology Data Exchange (ETDEWEB)

    He Jie, E-mail: hejiejoe@163.co [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China); Xu Xianhong [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China)

    2010-01-15

    The problem of nuclear reactor ageing is a topic of increasing importance in nuclear safety recent years. Ageing management is usually implemented for reactors maintenance. In the practice, a large number of data and records need to be processed. However, there are few professional software applications that aid reactor ageing management, especially for research reactors. This paper introduces the design of a new web-based management information system (MIS), named the Shielding Experimental Reactor Ageing Management Information System (SERAMIS). It is an auxiliary means that helps to collect data, keep records, and retrieve information for a research reactor ageing management. The Java2 Enterprise Edition (J2EE) and network database techniques, such as three-tiered model, Model-View-Controller architecture, transaction-oriented operations, and JavaScript techniques, are used in the development of this system. The functionalities of the application cover periodic safety review (PSR), regulatory references, data inspection, and SSCs classification according to ageing management methodology. Data and examples are presented to demonstrate the functionalities. For future work, techniques of data mining will be employed to support decision-making.

  7. Proceedings of the 1988 International Meeting on Reduced Enrichment for Research and Test Reactors

    Energy Technology Data Exchange (ETDEWEB)

    1993-07-01

    The international effort to develop and implement new research reactor fuels utilizing low-enriched uranium, instead of highly- enriched uranium, continues to make solid progress. This effort is the cornerstone of a widely shared policy aimed at reducing, and possibly eliminating, international traffic in highly-enriched uranium and the nuclear weapon proliferation concerns associated with this traffic. To foster direct communication and exchange of ideas among the specialists in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at Argonne National Laboratory, sponsored this meeting as the eleventh of a series which began 1978. Individual papers presented at the meeting have been cataloged separately.

  8. An easily regenerable enzyme reactor prepared from polymerized high internal phase emulsions.

    Science.gov (United States)

    Ruan, Guihua; Wu, Zhenwei; Huang, Yipeng; Wei, Meiping; Su, Rihui; Du, Fuyou

    2016-04-22

    A large-scale high-efficient enzyme reactor based on polymerized high internal phase emulsion monolith (polyHIPE) was prepared. First, a porous cross-linked polyHIPE monolith was prepared by in-situ thermal polymerization of a high internal phase emulsion containing styrene, divinylbenzene and polyglutaraldehyde. The enzyme of TPCK-Trypsin was then immobilized on the monolithic polyHIPE. The performance of the resultant enzyme reactor was assessed according to the conversion ability of Nα-benzoyl-l-arginine ethyl ester to Nα-benzoyl-l-arginine, and the protein digestibility of bovine serum albumin (BSA) and cytochrome (Cyt-C). The results showed that the prepared enzyme reactor exhibited high enzyme immobilization efficiency and fast and easy-control protein digestibility. BSA and Cyt-C could be digested in 10 min with sequence coverage of 59% and 78%, respectively. The peptides and residual protein could be easily rinsed out from reactor and the reactor could be regenerated easily with 4 M HCl without any structure destruction. Properties of multiple interconnected chambers with good permeability, fast digestion facility and easily reproducibility indicated that the polyHIPE enzyme reactor was a good selector potentially applied in proteomics and catalysis areas.

  9. Gas reactor international cooperative program. HTR-synfuel application assessment

    Energy Technology Data Exchange (ETDEWEB)

    1979-09-01

    This study assesses the technical, environmental and economic factors affecting the application of the High Temperature Gas-Cooled Thermal Reactor (HTR) to: synthetic fuel production; and displacement of fossil fuels in other industrial and chemical processes. Synthetic fuel application considered include coal gasification, direct coal liquefaction, oil shale processing, and the upgrading of syncrude to motor fuel. A wide range of other industrial heat applications was also considered, with emphasis on the use of the closed-loop thermochemical energy pipeline to supply heat to dispersed industrial users. In this application syngas (H/sub 2/ +CO/sub 2/) is produced at the central station HTR by steam reforming and the gas is piped to individual methanators where typically 1000/sup 0/F steam is generated at the industrial user sites. The products of methanation (CH/sub 4/ + H/sub 2/O) are piped back to the reformer at the central station HTR.

  10. An easily regenerable enzyme reactor prepared from polymerized high internal phase emulsions

    Energy Technology Data Exchange (ETDEWEB)

    Ruan, Guihua, E-mail: guihuaruan@hotmail.com [Guangxi Key Laboratory of Electrochemical and Magnetochemical Functional Materials, College of Chemistry and Bioengineering, Guilin University of Technology, Guangxi 541004 (China); Guangxi Collaborative Innovation Center for Water Pollution Control and Water Safety in Karst Area, Guilin University of Technology, Guilin 541004 (China); Wu, Zhenwei; Huang, Yipeng; Wei, Meiping; Su, Rihui [Guangxi Key Laboratory of Electrochemical and Magnetochemical Functional Materials, College of Chemistry and Bioengineering, Guilin University of Technology, Guangxi 541004 (China); Du, Fuyou, E-mail: dufu2005@126.com [Guangxi Key Laboratory of Electrochemical and Magnetochemical Functional Materials, College of Chemistry and Bioengineering, Guilin University of Technology, Guangxi 541004 (China); Guangxi Collaborative Innovation Center for Water Pollution Control and Water Safety in Karst Area, Guilin University of Technology, Guilin 541004 (China)

    2016-04-22

    A large-scale high-efficient enzyme reactor based on polymerized high internal phase emulsion monolith (polyHIPE) was prepared. First, a porous cross-linked polyHIPE monolith was prepared by in-situ thermal polymerization of a high internal phase emulsion containing styrene, divinylbenzene and polyglutaraldehyde. The enzyme of TPCK-Trypsin was then immobilized on the monolithic polyHIPE. The performance of the resultant enzyme reactor was assessed according to the conversion ability of N{sub α}-benzoyl-L-arginine ethyl ester to N{sub α}-benzoyl-L-arginine, and the protein digestibility of bovine serum albumin (BSA) and cytochrome (Cyt-C). The results showed that the prepared enzyme reactor exhibited high enzyme immobilization efficiency and fast and easy-control protein digestibility. BSA and Cyt-C could be digested in 10 min with sequence coverage of 59% and 78%, respectively. The peptides and residual protein could be easily rinsed out from reactor and the reactor could be regenerated easily with 4 M HCl without any structure destruction. Properties of multiple interconnected chambers with good permeability, fast digestion facility and easily reproducibility indicated that the polyHIPE enzyme reactor was a good selector potentially applied in proteomics and catalysis areas. - Graphical abstract: Schematic illustration of preparation of hypercrosslinking polyHIPE immobilized enzyme reactor for on-column protein digestion. - Highlights: • A reactor was prepared and used for enzyme immobilization and continuous on-column protein digestion. • The new polyHIPE IMER was quite suit for protein digestion with good properties. • On-column digestion revealed that the IMER was easy regenerated by HCl without any structure destruction.

  11. Development of an Internally Circulating Fluidized Bed Membrane Reactor for Hydrogen Production from Natural Gas

    Institute of Scientific and Technical Information of China (English)

    XIE Dong-lai; GRACE John R; LIM C Jim

    2006-01-01

    An innovative Internally Circulating Fluidized Bed Membrane Reactor (ICFBMR) was designed and operated for ultra-pure hydrogen production from natural gas. The reactor includes internal catalyst solids circulation for conveying heat between a reforming zone and an oxidation zone. In the reforming zone, catalyst particles are transported upwards by reactant gas where steam reforming reactions are taking place and hydrogen is permeating through the membrane surfaces. Air is injected into the oxidation zone to generate heat which is carried by catalyst particles to the reforming zone supporting the endothermic steam reforming reaction. The technology development process is introduced: cold model test,pilot plant and industrial demonstration unit. The process flow diagram and key components of each unit are described.The ICFBMR process has the potential to provide improved performance relative to conventional SMR fixed-bed tubular reactors.

  12. Antenna design for fast ion collective Thomson scattering diagnostic for the international thermonuclear experimental reactor

    DEFF Research Database (Denmark)

    Leipold, Frank; Furtula, Vedran; Salewski, Mirko

    2009-01-01

    Fast ion physics will play an important role for the international thermonuclear experimental reactor (ITER), where confined alpha particles will affect and be affected by plasma dynamics and thereby have impacts on the overall confinement. A fast ion collective Thomson scattering (CTS) diagnostic...

  13. Optimization of Internal Cooling Fins for Metal Hydride Reactors

    Directory of Open Access Journals (Sweden)

    Vamsi Krishna Kukkapalli

    2016-06-01

    Full Text Available Metal hydride alloys are considered as a promising alternative to conventional hydrogen storage cylinders and mechanical hydrogen compressors. Compared to storing in a classic gas tank, metal hydride alloys can store hydrogen at nearly room pressure and use less volume to store the same amount of hydrogen. However, this hydrogen storage method necessitates an effective way to reject the heat released from the exothermic hydriding reaction. In this paper, a finned conductive insert is adopted to improve the heat transfer in the cylindrical reactor. The fins collect the heat that is volumetrically generated in LaNi5 metal hydride alloys and deliver it to the channel located in the center, through which a refrigerant flows. A multiple-physics modeling is performed to analyze the transient heat and mass transfer during the hydrogen absorption process. Fin design is made to identify the optimum shape of the finned insert for the best heat rejection. For the shape optimization, use of a predefined transient heat generation function is proposed. Simulations show that there exists an optimal length for the fin geometry.

  14. Ageing problems and renovation programme of ET-RR-1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Khattab, M.S. [Reactors Dept., Nuclear Research Center, Cairo (Egypt); Sultan, M.A. [Reactors Dept., Nuclear Research Center, Cairo (Egypt)

    1995-12-31

    Based on Practical Experience gained from interfacing ageing systems in addition to operating new systems, current problems could be deduced whenever in-service inspection are carried out. This paper summarizes the in-service inspection made, and the proposed programme of rehabilitation of mechanical system in the ET-RR-1 research reactor at Inshass. Exchangeable experience in solving common problems in similar reactors play an important role in the effectiveness of such rehabilitation programme. The paper summarizes also the modernization of control, measuring and radiation monitoring system already carried out at the reactor. (orig.)

  15. Study on the flow characteristics and the wastewater treatment performance in modified internal circulation reactor.

    Science.gov (United States)

    Wang, Jiade; Xu, Weijun; Yan, Jingjia; Yu, Jianming

    2014-12-01

    A modified internal circulation (MIC) reactor with an external circulation system was proposed and the performance of treating dyeing wastewater using both MIC and typical IC reactor were compared. Utilization of the external circulation system in the MIC reactor could dramatically improve the mixing intensity of the biomass with the wastewater and resulted in better performance. The COD removal efficiency, biogas production, volatile fatty acids and effluent color were approximately 87%, 98 L d−1, 180 mg L−1 and 100 times, respectively, in the MIC reactor with a hydraulic retention time of 5 h and organic loading rate of 15 kg COD m−3 d−1. The hydrodynamics of the MIC reactor under different flows rate of external circulation were also analyzed using computational fluid dynamics (CFD) method. The optimal flow rate of external circulation was 12 L min−1, which resulted in a corresponding up-flow velocity of 40 m h−1. The consistency of the result between experiment and simulation validated the scientificity of CFD technique applied to numerical simulation of the MIC reactor.

  16. Simulation of a Neutron Noise Analysis Method for the Detection of Reactor Internals Vibration

    Energy Technology Data Exchange (ETDEWEB)

    Christian, Robby [Korea Advanced Institue of Science and Technology, Daejeon (Korea, Republic of); Song, Seon Ho [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2013-10-15

    The results were compared against expected hypothesis. This simulation technique developed in C++ programming environment can successfully illustrate the principle of the neutron noise analysis as a vibration monitoring method. The addition of a white noise signal spectrum into the neutron flux data may result in a better coherence analysis. Examination of the phase data on adjacent and opposite flux pairs may be used to determine the vibration mode 3 session. Safety aspect is always highly demanded in any nuclear power plants operation. To achieve a high level of safety, it is desirable to perform preventive measures instead of corrective ones. One of these measures is the monitoring of reactor internals vibration characteristics. Any changes in the vibration signatures indicates an anomaly in the reactor internals. One proven method for this purpose is by analyzing the neutron flux sensed by ex-core detectors around the reactor core. Standards and guides have been written on the proper conduct of this method. The American Society of Mechanical Engineers (ASME) published two similar guides in the ASME OM-S/G-2007 document. Part 5 focuses on specifically monitoring the core support barrel axial preload. Part 23 elaborates on monitoring of reactor internals vibrations. U. S. Nuclear Regulatory Commission (NRC) issued a Regulatory Guide 1.20 on Comprehensive Vibration Assessment Program (CVAP). To understand the principle of neutron noise analysis on vibration monitoring, a simple neutron-transport model was simulated.

  17. INTEGRAL BENCHMARKS AVAILABLE THROUGH THE INTERNATIONAL REACTOR PHYSICS EXPERIMENT EVALUATION PROJECT AND THE INTERNATIONAL CRITICALITY SAFETY BENCHMARK EVALUATION PROJECT

    Energy Technology Data Exchange (ETDEWEB)

    J. Blair Briggs; Lori Scott; Enrico Sartori; Yolanda Rugama

    2008-09-01

    Interest in high-quality integral benchmark data is increasing as efforts to quantify and reduce calculational uncertainties accelerate to meet the demands of next generation reactor and advanced fuel cycle concepts. The International Reactor Physics Experiment Evaluation Project (IRPhEP) and the International Criticality Safety Benchmark Evaluation Project (ICSBEP) continue to expand their efforts and broaden their scope to identify, evaluate, and provide integral benchmark data for method and data validation. Benchmark model specifications provided by these two projects are used heavily by the international reactor physics, nuclear data, and criticality safety communities. Thus far, 14 countries have contributed to the IRPhEP, and 20 have contributed to the ICSBEP. The status of the IRPhEP and ICSBEP is discussed in this paper, and the future of the two projects is outlined and discussed. Selected benchmarks that have been added to the IRPhEP and ICSBEP handbooks since PHYSOR’06 are highlighted, and the future of the two projects is discussed.

  18. Simultaneous nitrification and denitrification based on internal circulation baffled reactor

    Directory of Open Access Journals (Sweden)

    LU Xiaoya

    2014-06-01

    Full Text Available Nitrogen removal experiments were carried out by using an internal circulation baffled bioreactor (ICBBR. Nitrate, nitrite and ammonia were used as N source for nitrogen removal experiments. The ICBBR has high nitrogen removal capacity. The removal rates of total nitrogen, nitrate, and nitrite are almost the same. When nitrate and nitrite were used as N sources their kinetic orders were 0.88. When ammonia was used as N source simultaneous nitrification and denitrification (SND was realized in ICBBR and ammonia removal fitted also 0.88 order kinetics, but total nitrogen removal fitted third-order kinetics. Nitrate and nitrite removal rates were faster than ammonia removal rate under the same C/N ratio, and total nitrogen removal rate increased with increasing C/N ratio.

  19. Control Rod Drive Mechanism Installed in the Internal of Reactor Pressure Vessel

    Energy Technology Data Exchange (ETDEWEB)

    Choi, M. H.; Choi, S.; Park, J. S.; Lee, J. S.; Kim, D. O.; Hur, N. S.; Hur, H.; Yu, J. Y

    2008-09-15

    This report describes the review results and important technologies related to the in-vessel type control rod drive mechanism. Generally, most of the CRDMs used in the PWR are attached outside of the reactor pressure vessel, and the pernetration of the vessel head can not avoid. However, in-vessel type CRDMs, which are installed inside the reactor vessel, can eliminate the possibility of rod ejection accidents and the penetration of the vessel head, and provide a compact design of the reactor vessel and containment. There are two kinds of in-vessel type CRDM concerning the driving force-driven by a driving motor and by a hydraulic force. Motor driven CRDMs have been mainly investigated in Japan(MRX, IMR, DRX, next generation BWR etc.), and developed the key components such as a canned motor, an integrated rod position indicator, a separating ball-nut and a ball bearing that can operate under the water conditions of a high temperature and pressure. The concept of hydraulically driven CRDMs have been first reported by KWU and Siemens for KWU 200 reactor, and Argentina(CAREM) and China(NHR-5, NHR-200) have been developed the internal CRDM with the piston and cylinder of slightly different geometries. These systems are driven by the hydraulic force which is produced by pumps outside of the reactor vessel and transmitted through a pipe penetrating the reactor vessel, and needs complicated control and piping systems including pumps, valves and pipes etc.. IRIS has been recently decided the internal CRDMs as the reference design, and an analytical and experimental investigations of the hydraulic drive concept are performed by POLIMI in Italy. Also, a small French company, MP98 has been developed a new type of control rods, called 'liquid control rods', where reactivity is controlled by the movement of a liquid absorber in a manometer type device.

  20. Proceedings of the 1990 International Meeting on Reduced Enrichment for Research and Test Reactors

    Energy Technology Data Exchange (ETDEWEB)

    1993-07-01

    The global effort to reduce, and possibly, eliminate the international traffic in highly-enriched uranium caused by its use in research reactors requires extensive cooperation and free exchange of information among all participants. To foster this free exchange of information, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at Argonne National Laboratory, sponsored this meeting as the thirteenth of a series which began in 1978. The common effort brought together, past, a large number of specialists from many countries. On hundred twenty-three participants from 26 countries, including scientists, reactor operators, and personnel from commercial fuel suppliers, research centers, and government organizations, convened in Newport, Rhode Island to discuss their results, their activities, and their plans relative to converting research reactors to low-enriched fuels. As more and more reactors convert to the use of low-enriched uranium, the emphasis of our effort has begun to shift from research and development to tasks more directly related to implementation of the new fuels and technologies that have been developed, and to refinements of those fuels and technologies. It is appropriate, for this reason, that the emphasis of this meeting was placed on safety and on conversion experiences. This individual papers in this report have been cataloged separately.

  1. High rate nitrogen removal by ANAMMOX internal circulation reactor (IC) for old landfill leachate treatment.

    Science.gov (United States)

    Phan, The Nhat; Van Truong, Thi Thanh; Ha, Nhu Biec; Nguyen, Phuoc Dan; Bui, Xuan Thanh; Dang, Bao Trong; Doan, Van Tuan; Park, Joonhong; Guo, Wenshan; Ngo, Huu Hao

    2017-06-01

    This study aimed to evaluate the performance of a high rate nitrogen removal lab-scale ANAMMOX reactor, namely Internal Circulation (IC) reactor, for old landfill leachate treatment. The reactor was operated with pre-treated leachate from a pilot Partial Nitritation Reactor (PNR) using a high nitrogen loading rate ranging from 2 to 10kgNm(-3)d(-1). High rate removal of nitrogen (9.52±1.11kgNm(-3)d(-1)) was observed at an influent nitrogen concentration of 1500mgNL(-1). The specific ANAMMOX activity was found to be 0.598±0.026gN2-NgVSS(-1)d(-1). Analysis of ANAMMOX granules suggested that 0.5-1.0mm size granular sludge was the dominant group. The results of DNA analysis revealed that Candidatus Kueneniastuttgartiensis was the dominant species (37.45%) in the IC reactor, whereas other species like uncultured Bacteroidetes bacterium only constituted 5.37% in the system, but they were still responsible for removing recalcitrant organic matter.

  2. GROWTH OF THE INTERNATIONAL CRITICALITY SAFETY AND REACTOR PHYSICS EXPERIMENT EVALUATION PROJECTS

    Energy Technology Data Exchange (ETDEWEB)

    J. Blair Briggs; John D. Bess; Jim Gulliford

    2011-09-01

    Since the International Conference on Nuclear Criticality Safety (ICNC) 2007, the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and the International Reactor Physics Experiment Evaluation Project (IRPhEP) have continued to expand their efforts and broaden their scope. Eighteen countries participated on the ICSBEP in 2007. Now, there are 20, with recent contributions from Sweden and Argentina. The IRPhEP has also expanded from eight contributing countries in 2007 to 16 in 2011. Since ICNC 2007, the contents of the 'International Handbook of Evaluated Criticality Safety Benchmark Experiments1' have increased from 442 evaluations (38000 pages), containing benchmark specifications for 3955 critical or subcritical configurations to 516 evaluations (nearly 55000 pages), containing benchmark specifications for 4405 critical or subcritical configurations in the 2010 Edition of the ICSBEP Handbook. The contents of the Handbook have also increased from 21 to 24 criticality-alarm-placement/shielding configurations with multiple dose points for each, and from 20 to 200 configurations categorized as fundamental physics measurements relevant to criticality safety applications. Approximately 25 new evaluations and 150 additional configurations are expected to be added to the 2011 edition of the Handbook. Since ICNC 2007, the contents of the 'International Handbook of Evaluated Reactor Physics Benchmark Experiments2' have increased from 16 different experimental series that were performed at 12 different reactor facilities to 53 experimental series that were performed at 30 different reactor facilities in the 2011 edition of the Handbook. Considerable effort has also been made to improve the functionality of the searchable database, DICE (Database for the International Criticality Benchmark Evaluation Project) and verify the accuracy of the data contained therein. DICE will be discussed in separate papers at ICNC 2011. The status of the

  3. Neutron Age Determination in Fast Reactor Materials using the Group Method

    Directory of Open Access Journals (Sweden)

    Kabanova Marina F.

    2016-01-01

    Full Text Available The article deals with the methods of identifying fast neutron age in sodium (Na and uranium-238 (238U; describes the model of advanced and effective fast neutron nuclear reactors (FN, where Na is a coolant while 238U is involved in the fuel cycle in large quantities; justifies the choice of the group method for calculating the neutron age value in the substances mentioned above that can show the accuracy of the used constants for Na and estimate various versions of multilevel description of neutron moderation in 238U – the most powerful resonance absorber of the neutron reactor active zone.

  4. Naval Reactors Prime Contractor Team (NRPCT) Experiences and Considerations With Irradiation Test Performance in an International Environment

    Energy Technology Data Exchange (ETDEWEB)

    MH Lane

    2006-02-15

    This letter forwards a compilation of knowledge gained regarding international interactions and issues associated with Project Prometheus. The following topics are discussed herein: (1) Assessment of international fast reactor capability and availability; (2) Japanese fast reactor (JOYO) contracting strategy; (3) NRPCT/Program Office international contract follow; (4) Completion of the Japan Atomic Energy Agency (JAEA)/Pacific Northwest National Laboratory (PNNL) contract for manufacture of reactor test components; (5) US/Japanese Departmental interactions and required Treaties and Agreements; and (6) Non-technical details--interactions and considerations.

  5. IAEA international studies on irradiation embrittlement of reactor pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Brumovsky, M. [Nuclear Research Institute Rez plc (Czech Republic); Steele, L.E. [Chief Scientific Investigator of the Programme, Springfield, VA (United States)

    1997-02-01

    In last 25 years, three phases a Co-operative Research Programme on Irradiation Embrittlement of Reactor Pressure Vessel Steels has been organized by the International Atomic Energy Agency. This programme started with eight countries in 1971 and finally 16 countries took part in phase III of the Programme in 1983. Several main efforts were put into preparation of the programme, but the principal task was concentrated on an international comparison of radiation damage characterization by different laboratories for steels of {open_quotes}old{close_quotes} (with high impurity contents) and {open_quotes}advanced{close_quotes} (with low impurity contents) types as well as on development of small scale fracture mechanics procedures applicable to reactor pressure vessel surveillance programmes. This year, a new programme has been opened, concentrated mostly on small scale fracture mechanics testing.

  6. [Radiation transformation mechanism in a photocatalytic reactor of three-phase internal circulating fluidized bed].

    Science.gov (United States)

    You, Hong; Luo, Wei-nan; Yao, Jie; Chen, Ping; Cai, Wei-min

    2005-01-01

    A novel three-phase internal circulating fluidized bed photocatalytic reactor was established and the radiation transformation in which was investigated. The experimental results indicate that with the interaction of gas and solid (gas flux > 0.3m3/h), the radiation transformation in the reactor along radial direction conforms to a definite exponential function, which agrees to formula Rose about the rules of light intensity distribution through evenly suspended particles. The value of radiation energy is affected by the initial light intensity, the concentration of photocatalyst and the thickness of liquid layer. The aerated gas amount only influence the state of the fluidized bed and has little effect on the distribution of light intensity along radical direction. Photocatalytic degradation of Rhodamine B indicate that the efficiency of three-phase internal circulating fluidized bed is much higher than slurry bed. The optimal catalyst concentration of this system is 10 - 12g/L.

  7. Thermal ageing mechanisms of VVER-1000 reactor pressure vessel steels

    Science.gov (United States)

    Shtrombakh, Yaroslav I.; Gurovich, Boris A.; Kuleshova, Evgenia A.; Maltsev, Dmitry A.; Fedotova, Svetlana V.; Chernobaeva, Anna A.

    2014-09-01

    In this paper a complex of microstructural studies (TEM and SEM) and a comparative analysis of the results of these studies with the data of mechanical tests of temperature sets of VVER-1000 RPV surveillance specimens with exposure times up to ∼200,000 h were conducted. Special annealing of control and temperature sets of SS which provides the dissolution of grain boundary segregation was performed to clarify the mechanisms of thermal ageing. It was demonstrated that during long-term exposures up to 200,000 h at the operating temperature of about 310-320 °C thermal ageing effects reveal themselves only for the weld metal (Ni content ⩾ 1.35%) and are the result of grain boundary segregation accumulation (development of reversible temper brittleness). The obtained results improve the accuracy of prediction of the thermal ageing rate of VVER-1000 materials in case of RPV service life extension up to 60 years.

  8. Reactor Design and Decommissioning - An Overview of International Activities in Post Fukushima Era1 - 12396

    Energy Technology Data Exchange (ETDEWEB)

    Devgun, Jas S. [Nuclear Power Technologies, Sargent and Lundy LLC, Chicago, IL (United States); Laraia, Michele [private consultant, formerly from IAEA, Kolonitzgasse 10/2, 1030, Vienna (Austria); Pescatore, Claudio [OECD, Nuclear Energy Agency, Issy-les-Moulineaux, Paris (France); Dinner, Paul [International Atomic Energy Agency, Wagramerstrasse 5, A-1400 Vienna (Austria)

    2012-07-01

    Accidents at the Fukushima Dai-ichi reactors as a result of the devastating earthquake and tsunami of March 11, 2011 have not only dampened the nuclear renaissance but have also initiated a re-examination of the design and safety features for the existing and planned nuclear reactors. Even though failures of some of the key site features at Fukushima can be attributed to events that in the past would have been considered as beyond the design basis, the industry as well as the regulatory authorities are analyzing what features, especially passive features, should be designed into the new reactor designs to minimize the potential for catastrophic failures. It is also recognized that since the design of the Fukushima BWR reactors which were commissioned in 1971, many advanced safety features are now a part of the newer reactor designs. As the recovery efforts at the Fukushima site are still underway, decisions with respect to the dismantlement and decommissioning of the damaged reactors and structures have not yet been finalized. As it was with Three Mile Island, it could take several decades for dismantlement, decommissioning and clean up, and the project poses especially tough challenges. Near-term assessments have been issued by several organizations, including the IAEA, the USNRC and others. Results of such investigations will lead to additional improvements in system and site design measures including strengthening of the anti-tsunami defenses, more defense-in-depth features in reactor design, and better response planning and preparation involving reactor sites. The question also arises what would the effect be on the decommissioning scene worldwide, and what would the effect be on the new reactors when they are eventually retired and dismantled. This paper provides an overview of the US and international activities related to recovery and decommissioning including the decommissioning features in the reactor design process and examines these from a new

  9. International Comparison of Age Discrimination Laws.

    Science.gov (United States)

    Lahey, Joanna N

    2010-11-01

    European age discrimination legislation is discussed in the context of the US Age Discrimination in Employment Act (ADEA) and related state laws. US law was originally introduced to protect productive older workers from age stereotypes, but more recently preventing age discrimination has become important as a means of keeping costs down on entitlement programs as the population ages. Changes in enforcement, penalties, exemptions, length of time to file, and burden of proof have changed the effects of the laws over time. The ADEA has had both positive effects on currently employed older workers and negative effects on the hiring of older workers. Enforcement and publicity are offered as possible explanations for the strength of these positive and negative effects. Age discrimination legislation in Europe, indicated in the Framework Directive 2000/78, is driven by economic and political considerations. European legislation calls for less enforcement and more exemptions than the corresponding US cases which could lead to smaller effects on employment. However, pensions, disability, unemployment, and social security potentially have a stronger effect on social norms for retirement age than does anti-discrimination legislation.

  10. Mixing Characteristics and Bubble Behavior in an Airlift Internal Loop Reactor with Low Aspect Ratio

    Institute of Scientific and Technical Information of China (English)

    张伟鹏; 雍玉梅; 张广积; 杨超; 毛在砂

    2014-01-01

    The present study summarizes the results of macro-and micro-mixing characteristics in an airlift inter-nal loop reactor with low aspect ratio (H/D≤5) using the electrolytic tracer response technique and the method of parallel competing reactions respectively. The micro-mixing has never been investigated in airlift loop reactors. The dual-tip electrical conductivity probe technique is used for measurement of local bubble behavior in the reactor. The effects of several operating parameters and geometric variables are investigated. It is found that the increase in su-perficial gas velocity corresponds to the increase in energy input, liquid circulation velocity and shear rate, decreas-ing the macro-mixing time and segregation index. Moreover, it is shown that top clearance and draft diameter affect flow resistance. However, the bubble redistribution with a screen mesh on the perforated plate distributor for macro-mixing is insignificant. The top region with a high energy dissipation rate is a suitable location for feeding reactants. The analysis of present experimental data provides a valuable insight into the interaction between gas and liquid phases for mixing and improves the understanding of intrinsic roles of hydrodynamics upon the reactor de-sign and operating parameter selection.

  11. Thermal-hydraulics of internally heated molten salts and application to the Molten Salt Fast Reactor

    Science.gov (United States)

    Fiorina, Carlo; Cammi, Antonio; Luzzi, Lelio; Mikityuk, Konstantin; Ninokata, Hisashi; Ricotti, Marco E.

    2014-04-01

    The Molten Salt Reactors (MSR) are an innovative kind of nuclear reactors and are presently considered in the framework of the Generation IV International Forum (GIF-IV) for their promising performances in terms of low resource utilization, waste minimization and enhanced safety. A unique feature of MSRs is that molten fluoride salts play the distinctive role of both fuel (heat source) and coolant. The presence of an internal heat generation perturbs the temperature field and consequences are to be expected on the heat transfer characteristics of the molten salts. In this paper, the problem of heat transfer for internally heated fluids in a straight circular channel is first faced on a theoretical ground. The effect of internal heat generation is demonstrated to be described by a corrective factor applied to traditional correlations for the Nusselt number. It is shown that the corrective factor can be fully characterized by making explicit the dependency on Reynolds and Prandtl numbers. On this basis, a preliminary correlation is proposed for the case of molten fluoride salts by interpolating the results provided by an analytic approach previously developed at the Politecnico di Milano. The experimental facility and the related measuring procedure for testing the proposed correlation are then presented. Finally, the developed correlation is used to carry out a parametric investigation on the effect of internal heat generation on the main out-of-core components of the Molten Salt Fast Reactor (MSFR), the reference circulating-fuel MSR design in the GIF-IV. The volumetric power determines higher temperatures at the channel wall, but the effect is significant only in case of large diameters and/or low velocities.

  12. Technical report on implementation of reactor internal 3D modeling and visual database system

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yeun Seung; Eom, Young Sam; Lee, Suk Hee; Ryu, Seung Hyun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-06-01

    In this report was described a prototype of reactor internal 3D modeling and VDB system for NSSS design quality improvement. For improving NSSS design quality several cases of the nuclear developed nation`s integrated computer aided engineering system, such as Mitsubishi`s NUWINGS (Japan), AECL`s CANDID (Canada) and Duke Power`s PASCE (USA) were studied. On the basis of these studies the strategy for NSSS design improvement system was extracted and detail work scope was implemented as follows : 3D modelling of the reactor internals were implemented by using the parametric solid modeler, a prototype system of design document computerization and database was suggested, and walk-through simulation integrated with 3D modeling and VDB was accomplished. Major effects of NSSS design quality improvement system by using 3D modeling and VDB are the plant design optimization by simulation, improving the reliability through the single design database system and engineering cost reduction by improving productivity and efficiency. For applying the VDB to full scope of NSSS system design, 3D modelings of reactor coolant system and nuclear fuel assembly and fuel rod were attached as appendix. 2 tabs., 31 figs., 7 refs. (Author) .new.

  13. IGORR-1: Proceedings of the first meeting of the international group on research reactors

    Energy Technology Data Exchange (ETDEWEB)

    West, C.D. (comp.)

    1990-05-01

    Many organizations, in several countries, are planning or implementing new or upgraded research reactor projects, but there has been no organized forum devoted entirely to discussion and exchange of information in this field. Over the past year or so, informal discussions resulted in widespread agreement that such a forum would serve a useful purpose. Accordingly, a proposal to form a group was submitted to the leading organizations known to be involved in projects to build or upgrade reactor facilities. Essentially all agreed to join in the formation of the International Group on Research Reactors (IGORR) and nominated a senior staff member to serve on its international organizing committee. The first IGORR meeting took place on February 28--March 2, 1990. It was very successful and well attended; some 52 scientists and engineers from 25 organizations in 10 countries participated in 2-1/2 days of open and informative presentations and discussions. Two workshop sessions offered opportunities for more detailed interaction among participants and resulted in identification of common R D needs, sources of data, and planned new facilities. Individual papers have been cataloged separately.

  14. Overview of Experiments for Physics of Fast Reactors from the International Handbooks of Evaluated Criticality Safety Benchmark Experiments and Evaluated Reactor Physics Benchmark Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Bess, J. D.; Briggs, J. B.; Gulliford, J.; Ivanova, T.; Rozhikhin, E. V.; Semenov, M. Yu.; Tsibulya, A. M.; Koscheev, V. N.

    2017-07-01

    Overview of Experiments to Study the Physics of Fast Reactors Represented in the International Directories of Critical and Reactor Experiments John D. Bess Idaho National Laboratory Jim Gulliford, Tatiana Ivanova Nuclear Energy Agency of the Organisation for Economic Cooperation and Development E.V.Rozhikhin, M.Yu.Sem?nov, A.M.Tsibulya Institute of Physics and Power Engineering The study the physics of fast reactors traditionally used the experiments presented in the manual labor of the Working Group on Evaluation of sections CSEWG (ENDF-202) issued by the Brookhaven National Laboratory in 1974. This handbook presents simplified homogeneous model experiments with relevant experimental data, as amended. The Nuclear Energy Agency of the Organization for Economic Cooperation and Development coordinates the activities of two international projects on the collection, evaluation and documentation of experimental data - the International Project on the assessment of critical experiments (1994) and the International Project on the assessment of reactor experiments (since 2005). The result of the activities of these projects are replenished every year, an international directory of critical (ICSBEP Handbook) and reactor (IRPhEP Handbook) experiments. The handbooks present detailed models of experiments with minimal amendments. Such models are of particular interest in terms of the settlements modern programs. The directories contain a large number of experiments which are suitable for the study of physics of fast reactors. Many of these experiments were performed at specialized critical stands, such as BFS (Russia), ZPR and ZPPR (USA), the ZEBRA (UK) and the experimental reactor JOYO (Japan), FFTF (USA). Other experiments, such as compact metal assembly, is also of interest in terms of the physics of fast reactors, they have been carried out on the universal critical stands in Russian institutes (VNIITF and VNIIEF) and the US (LANL, LLNL, and others.). Also worth mentioning

  15. Overview of the 2014 Edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook)

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess; J. Blair Briggs; Jim Gulliford; Ian Hill

    2014-10-01

    The International Reactor Physics Experiment Evaluation Project (IRPhEP) is a widely recognized world class program. The work of the IRPhEP is documented in the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook). Integral data from the IRPhEP Handbook is used by reactor safety and design, nuclear data, criticality safety, and analytical methods development specialists, worldwide, to perform necessary validations of their calculational techniques. The IRPhEP Handbook is among the most frequently quoted reference in the nuclear industry and is expected to be a valuable resource for future decades.

  16. Health perspectives: international epidemiology of ageing.

    Science.gov (United States)

    Ward, Stephanie Alison; Parikh, Seema; Workman, Barbara

    2011-09-01

    Populations globally are ageing, in part due to dramatic increases in life expectancies, forcing a reconsideration of what constitutes being "elderly" and "old." The proportion of older adults living with disability may be decreasing, yet older individuals are living with a significant burden of chronic disease, geriatric impairments in cognition, vision and hearing and reduced physiological reserve (frailty). Caring for a growing number of medically complex individuals has implications for medical workforce size and composition, health programmes and expenditure. Future responses to an ageing population will require further innovation in health-care delivery models, and increasing representation of older adults in clinical trials. 2011 Elsevier Ltd. All rights reserved.

  17. Fast reactor safety: proceedings of the international topical meeting. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    1985-07-01

    The emphasis of this meeting was on the safety-related aspects of fast reactor design, analysis, licensing, construction, and operation. Relative to past meetings, there was less emphasis on the scientific and technological basis for accident assessment. Because of its broad scope, the meeting attracted 217 attendees from a wide cross section of the design, safety analysis, and safety technology communities. Eight countries and two international organizations were represented. A total of 126 papers were presented, with contributions from the United States, France, Japan, the United Kingdom, Germany, and Italy. Sessions covered in Volume 1 include: impact of safety and licensing considerations on fast reactor design; safety aspects of innovative designs; intra-subassembly behavior; operational safety; design accommodation of seismic and other external events; natural circulation; safety design concepts; safety implications derived from operational plant data; decay heat removal; and assessment of HCDA consequences.

  18. International academic program in technologies of light-water nuclear reactors. Phases of development and implementation

    Science.gov (United States)

    Geraskin, N. I.; Glebov, V. B.

    2017-01-01

    The results of implementation of European educational projects CORONA and CORONA II dedicated to preserving and further developing nuclear knowledge and competencies in the area of technologies of light-water nuclear reactors are analyzed. Present article addresses issues of design and implementation of the program for specialized training in the branch of technologies of light-water nuclear reactors. The systematic approach has been used to construct the program for students of nuclear specialties, which corresponding to IAEA standards and commonly accepted nuclear principles recognized in the European Union. Possibilities of further development of the international cooperation between countries and educational institutions are analyzed. Special attention is paid to e-learning/distance training, nuclear knowledge preservation and interaction with European Nuclear Education Network.

  19. RADIATION DOSIMETRY OF THE PRESSURE VESSEL INTERNALS OF THE HIGH FLUX BEAM REACTOR.

    Energy Technology Data Exchange (ETDEWEB)

    HOLDEN,N.E.; RECINIELLO,R.N.; HU,J.P.; RORER,D.C.

    2002-08-18

    In preparation for the eventual decommissioning of the High Flux Beam Reactor after the permanent removal of its fuel elements from the Brookhaven National Laboratory, both measurements and calculations of the decay gamma-ray dose rate have been performed for the reactor pressure vessel and vessel internal structures which included the upper and lower thermal shields, the transition plate, and the control rod blades. The measurements were made using Red Perspex{trademark} polymethyl methacrylate high-level film dosimeters, a Radcal ''peanut'' ion chamber, and Eberline's high-range ion chamber. To compare with measured gamma-ray dose rate, the Monte Carlo MCNP code and geometric progressive Microshield code were used to model the gamma transport and dose buildup.

  20. Benchmark Data Through The International Reactor Physics Experiment Evaluation Project (IRPHEP)

    Energy Technology Data Exchange (ETDEWEB)

    J. Blair Briggs; Dr. Enrico Sartori

    2005-09-01

    The International Reactor Physics Experiments Evaluation Project (IRPhEP) was initiated by the Organization for Economic Cooperation and Development (OECD) Nuclear Energy Agency’s (NEA) Nuclear Science Committee (NSC) in June of 2002. The IRPhEP focus is on the derivation of internationally peer reviewed benchmark models for several types of integral measurements, in addition to the critical configuration. While the benchmarks produced by the IRPhEP are of primary interest to the Reactor Physics Community, many of the benchmarks can be of significant value to the Criticality Safety and Nuclear Data Communities. Benchmarks that support the Next Generation Nuclear Plant (NGNP), for example, also support fuel manufacture, handling, transportation, and storage activities and could challenge current analytical methods. The IRPhEP is patterned after the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and is closely coordinated with the ICSBEP. This paper highlights the benchmarks that are currently being prepared by the IRPhEP that are also of interest to the Criticality Safety Community. The different types of measurements and associated benchmarks that can be expected in the first publication and beyond are described. The protocol for inclusion of IRPhEP benchmarks as ICSBEP benchmarks and for inclusion of ICSBEP benchmarks as IRPhEP benchmarks is detailed. The format for IRPhEP benchmark evaluations is described as an extension of the ICSBEP format. Benchmarks produced by the IRPhEP add new dimension to criticality safety benchmarking efforts and expand the collection of available integral benchmarks for nuclear data testing. The first publication of the "International Handbook of Evaluated Reactor Physics Benchmark Experiments" is scheduled for January of 2006.

  1. Reactor

    Science.gov (United States)

    Evans, Robert M.

    1976-10-05

    1. A neutronic reactor having a moderator, coolant tubes traversing the moderator from an inlet end to an outlet end, bodies of material fissionable by neutrons of thermal energy disposed within the coolant tubes, and means for circulating water through said coolant tubes characterized by the improved construction wherein the coolant tubes are constructed of aluminum having an outer diameter of 1.729 inches and a wall thickness of 0.059 inch, and the means for circulating a liquid coolant through the tubes includes a source of water at a pressure of approximately 350 pounds per square inch connected to the inlet end of the tubes, and said construction including a pressure reducing orifice disposed at the inlet ends of the tubes reducing the pressure of the water by approximately 150 pounds per square inch.

  2. The second international conference "genetics of aging and longevity".

    NARCIS (Netherlands)

    Anisimov, V.N.; Bartke, A.; Barzilai, N.; Batin, M.A.; Blagosklonny, M.V.; Brown-Borg, H.; Budovskaya, Y.; Campisi, J.; Friguet, B.; Fraifeld, V.; Franceschi, C.; Gems, D.; Gladyshev, V.; Gorbunova, V.; Gudkov, A.V.; Kennedy, B.; Konovalenko, M.; Kraemer, B.; Moskalev, A.; Petropoulos, I.; Pasyukova, E.; Rattan, S.; Rogina, B.; Seluanov, A.; Shaposhnikov, M.; Shmookler Reis, R.; Tavernarakis, N.; Vijg, J.; Yashin, A.; Zimniak, P.

    2012-01-01

    The ongoing revolution in aging research was manifested by the Second International Conference "Genetics of Aging and Longevity" (Moscow, April 22-25, 2012). The Conference goal was to identify the most promising areas of genetics, life expectancy, and aging, including: the search for longevity gene

  3. Determination of thermal-hydraulic loads on reactor internals in a DBA-situation

    Energy Technology Data Exchange (ETDEWEB)

    Ville Lestinen; Timo Toppila [POB 10, 00048 FORTUM (Finland)

    2005-07-01

    Full text of publication follows: According to Finnish regulatory requirements, reactor internals have to stay intact in a design basis accident (DBA) situation, so that control rods can still penetrate into the core. To fulfill this demand some criteria must be followed in periodical in-service inspections. This is the motivation for studying and developing more detailed methods for analysis of thermal-hydraulic loads on reactor internals during the DBA-situation for the Loviisa NPP in Finland. The objective of this research program is to connect thermal-hydraulic and mechanical analysis methods with the goal to produce a reliable method for determination of thermal-hydraulic and mechanical loads on reactor internals in the accident situation. The tools studied are thermal-hydraulic system codes, computational fluid dynamics (CFD) codes and finite element analysis (FEA) codes. This paper concentrates mainly on thermal-hydraulic part of the research, but also the mechanical aspects are discussed. Firstly, the paper includes a short literary review of the available methods to analyse the described problem including both thermal-hydraulic and structural analysis parts. Secondly, different possibilities to carry out thermal-hydraulic analyses have been studied. The DBA-case includes complex physical phenomena and therefore modelling is difficult. The accident situation can be for example LLOCA. When the pipe has broken, the pressure decreases and water starts to evaporate, which consumes energy and that way limits the pressure decrease. After some period of time, the system reaches a new equilibrium state. To perform exact thermal-hydraulic analysis also two phase phenomena must be included. Therefore CFD codes are not capable of modelling the DBA situation very well, but the use of CFD codes requires that the effect of two phase flow must be added somehow. One method to calculate two phase phenomena with CFD codes is to use thermal-hydraulic system codes to calculate

  4. Mitigating IASCC of Reactor Core Internals by Post-Irradiation Annealing

    Energy Technology Data Exchange (ETDEWEB)

    Was, Gary [Univ. of Michigan, Ann Arbor, MI (United States)

    2017-06-02

    This final report summarizes research performed during the period between September 2012 and December 2016, with the objective of establishing the effectiveness of post-irradiation annealing (PIA) as an advanced mitigation strategy for irradiation-assisted stress corrosion cracking (IASCC). This was completed by using irradiated 304SS control blade material to conduct crack initiation and crack growth rate (CGR) experiments in simulated BWR environment. The mechanism by which PIA affects IASCC susceptibility will also be verified. The success of this project will provide a foundation for the use of PIA as a mitigation strategy for core internal components in commercial reactors.

  5. Dynamic Analysis of Coolant Channel and Its Internals of Indian 540 MWe PHWR Reactor

    Directory of Open Access Journals (Sweden)

    N. Dharmaraju

    2008-01-01

    Full Text Available The horizontal coolant channel is one of the important parts of primary heat transport system in PHWR type of reactors. There are in all 392 channels in the core of Indian 540 MWe reactor. Each channel houses 13 natural uranium fuel bundles and shielding and sealing plugs one each on either side of the channel. The heavy water coolant flows through the coolant channel and carries the nuclear heat to outside the core for steam generation and power production in the turbo-generator. India has commissioned one 540 MWe PHWR reactor in September 2005 and another similar unit will be going into operation very shortly. For a complete dynamic study of the channel and its internals under the influence of high coolant flow, experimental and modeling studies have been carried out. A good correlation has been achieved between the results of experimental and analytical models. The operating life of a typical coolant channel typically ranges from 10 to 15 full-power years. Towards the end of its operating life, its health monitoring becomes an important activity. Vibration diagnosis plays an important role as a tool for life management of coolant. Through the study of dynamic characteristics of the coolant channel under simulated loading condition, an attempt has been made to develop a diagnostics to monitor the health of the coolant channel over its operating life. A study has been also carried out to characterize the fuel vibration under different flow condition.

  6. Modeling and temperature regulation of a thermally coupled reactor system via internal model control strategy

    Energy Technology Data Exchange (ETDEWEB)

    Lee, S.Y.; Coronella, C.J.; Bhadkamkar, A.S.; Seader, J.D. [Univ. of Utah, Salt Lake City, UT (United States). Dept. of Chemical and Fuels Engineering

    1993-12-01

    A two-stage, thermally coupled fluidized-bed reactor system has been developed for energy-efficient conversion of tar-sand bitumen to synthetic crude oil. Modeling and temperature control of a system are addressed in this study. A process model and transfer function are determined by a transient response technique and the reactor temperature are controlled by PI controllers with tuning settings determined by an internal model control (IMC) strategy. Using the IMC tuning method, sufficiently good control performance was experimentally observed without lengthy on-line tuning. It is shown that IMC strategy provides a means to directly use process knowledge to make a control decision. Although this control method allows for fine tuning by adjusting a single tuning parameter, it is not easy to determine the optimal value of this tuning parameter, which must be specified by the user. A novel method is presented to evaluate that parameter, which must be specified by the user. A novel method is presented to evaluate that parameter in this study. It was selected based on the magnitude of elements on the off-diagonal of the relative gain array to account for the effect of thermal coupling on control performance. It is shown that this method provides stable and fast control of reactor temperatures. By successfully decoupling the system, a simple method of extending the IMC tuning technique to multiinput/multioutput systems is obtained.

  7. Homogenization of the internal structures of a reactor with the cooling fluid

    Energy Technology Data Exchange (ETDEWEB)

    Robbe, M.F. [CEA Saclay, SEMT, 91 - Gif sur Yvette (France); Bliard, F. [Socotec Industrie, Service AME, 78 - Montigny le Bretonneux (France)

    2001-07-01

    To take into account the influence of a structure net among a fluid flow, without modelling exactly the structure shape, a concept of ''equivalent porosity method'' was developed. The structures are considered as solid pores inside the fluid. The structure presence is represented by three parameters: a porosity, a shape coefficient and a pressure loss coefficient. The method was studied for an Hypothetical Core Disruptive Accident in a Liquid Metal Fast Breeder Reactor, but it can be applied to any problem involving fluid flow getting through a solid net. The model was implemented in the computer code CASTEM-PLEXUS and validated on an analytical shock tube test, simulating an horizontal slice of a schematic LMFBR in case of a HCDA (bubble at high pressure, liquid sodium and internal structures of the reactor). A short parametric study shows the influence of the porosity and the structure shape on the pressure wave impacting the shock tube bottom. These results were used to simulate numerically the HCDA mechanical effects in a small scale reactor mock-up. (author)

  8. DETERMINING THE EFFECTS OF RADIATION ON AGING CONCRETE STRUCTURES OF NUCLEAR REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    Serrato, M.

    2010-01-29

    The U.S. Department of Energy Office of Environmental Management (DOE-EM) is responsible for the Decontamination and Decommissioning (D&D) of nuclear facilities throughout the DOE Complex. Some of these facilities will be completely dismantled, while others will be partially dismantled and the remaining structure will be stabilized with cementitious fill materials. The latter is a process known as In-Situ Decommissioning (ISD). The ISD decision process requires a detailed understanding of the existing facility conditions, and operational history. System information and material properties are need for aged nuclear facilities. This literature review investigated the properties of aged concrete structures affected by radiation. In particular, this review addresses the Savannah River Site (SRS) isotope production nuclear reactors. The concrete in the reactors at SRS was not seriously damaged by the levels of radiation exposure. Loss of composite compressive strength was the most common effect of radiation induced damage documented at nuclear power plants.

  9. Creep-Fatigue Damage Evaluation of a Model Reactor Vessel and Reactor Internals of Sodium Test Facility according to ASME-NH and RCC-MRx Codes

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Dong-Won; Lee, Hyeong-Yeon; Eoh, Jae-Hyuk; Son, Seok-Kwon; Kim, Jong-Bum; Jeong, Ji-Young [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    The objective of the STELLA-2 is to support the specific design approval for PGSFR by synthetic reviews of key safety issues and code validations through the integral effect tests. Due to its high temperature operation in SFRs (and in a testing facility) up to 550 °C, thermally induced creep-fatigue damage is very likely in components including a reactor vessel, reactor internals (interior structures), heat exchangers, pipelines, etc. In this study, structural integrity of the components such as reactor vessel and internals in STELLA-2 has been evaluated against creep-fatigue failures at a concept-design step. As 2D analysis yields far conservative results, a realistic 3D simulation is performed by a commercial software. A design integrity guarding against a creep-fatigue damage failure operating at high temperature was evaluated for the reactor vessel with its internal structure of the STELLA-2. Both the high temperature design codes were used for the evaluation, and results were compared. All the results showed the vessel as a whole is safely designed at the given operating conditions, while the ASME-NH gives a conservative evaluation.

  10. The Halden reactor, a facility open to the international nuclear community; Halden, un reacteur ouvert a la communaute internationale

    Energy Technology Data Exchange (ETDEWEB)

    Vitanza, C. [Organisation for Economic Co-Operation and Development, Nuclear Energy Agency (OECD/NEA), 75 - Paris (France)

    2005-07-01

    The Halden test reactor is a boiling-type reactor moderated and cooled by heavy water, it yields a thermal power of 20 MW. The reactor operates for 2 periods of about 100 days each year. The Halden reactor has been in operation for more than 45 years and is the largest OECD-NEA project, it carries out the OECD joint program and bilateral contract work. Its experimental programs are supported by about 100 organisations in 20 countries. The fuel and materials programs for the years to come focus on the following main areas: -) fuel high burn-up capabilities in normal operating conditions, -) fuel response to transients aiming at generating experimental data on the behaviour of high burn-up fuels in short duration transients and on phenomena occurring during loss of coolant accident and coolant flow oscillations, -) cladding corrosion and water chemistry issues, and -) pressure vessel embrittlement and irradiation assisted stress corrosion cracking of reactor internals. (A.C.)

  11. ITER (International Thermonuclear Experimental Reactor) shield and blanket work package report

    Energy Technology Data Exchange (ETDEWEB)

    1988-06-01

    This report summarizes nuclear-related work in support of the US effort for the International Thermonuclear Experimental Reactor (ITER) Study. The purpose of this work was to prepare for the first international ITER workshop devoted to defining a basic ITER concept that will serve as a basis for an indepth conceptual design activity over the next 2-1/2 years. Primary tasks carried out during the past year included: design improvements of the inboard shield developed for the TIBER concept, scoping studies of a variety of tritium breeding blanket options, development of necessary design guidelines and evaluation criteria for the blanket options, further safety considerations related to nuclear components and issues regarding structural materials for an ITER device. 44 refs., 31 figs., 29 tabs.

  12. International Nuclear Safety Center database on thermophysical properties of reactor materials

    Energy Technology Data Exchange (ETDEWEB)

    Fink, J.K.; Sofu, T.; Ley, H.

    1997-08-01

    The International Nuclear Safety Center (INSC) database has been established at Argonne National Laboratory to provide easily accessible data and information necessary to perform nuclear safety analyses and to promote international collaboration through the exchange of nuclear safety information. The INSC database, located on the World Wide Web at http://www.insc.anl.gov, contains critically assessed recommendations for reactor material properties for normal operating conditions, transients, and severe accidents. The initial focus of the database is on thermodynamic and transport properties of materials for water reactors. Materials that are being included in the database are fuel, absorbers, cladding, structural materials, coolant, and liquid mixtures of combinations of UO{sub 2}, ZrO{sub 2}, Zr, stainless steel, absorber materials, and concrete. For each property, the database includes: (1) a summary of recommended equations with uncertainties; (2) a detailed data assessment giving the basis for the recommendations, comparisons with experimental data and previous recommendations, and uncertainties; (3) graphs showing recommendations, uncertainties, and comparisons with data and other equations; and (4) property values tabulated as a function of temperature.

  13. The effects of low dose rate irradiation and thermal aging on reactor structural alloys

    Science.gov (United States)

    Allen, T. R.; Trybus, C. L.; Cole, J. I.

    As part of the EBR-II reactor materials surveillance program, test samples of fifteen different alloys were placed into EBR-II in 1965. The surveillance (SURV) program was intended to determine property changes in reactor structural materials caused by irradiation and thermal aging. In this work, the effect of low dose rate (approximately 2 × 10 -8 dpa/s) irradiation at 380-410°C and long term thermal aging at 371°C on the properties of 20% cold worked 304 stainless steel, 420 stainless steel, Inconel X750, 304/308 stainless weld material, and 17-4 PH steel are evaluated. Doses of up to 6.8 dpa and thermal aging to 2994 days did not significantly affect the density of these alloys. The strength of 304 SS, X750, 17-4 PH, and 304/308 weld material increased with irradiation. In contrast, the strength of 420 stainless steel decreased with irradiation. Irradiation decreased the impact energy in both Inconel X750 and 17-4 PH steel. Thermal aging decreased the impact energy in 17-4 PH steel and increased the impact energy in Inconel X750. Tensile property comparisons of 304 SURV samples with 304 samples irradiated in EBR-II at a higher dose rate show that the higher dose rate samples had greater increases in strength and greater losses in ductility.

  14. Requirements for US regulatory approval of the International Thermonuclear Experimental Reactor (ITER)

    Energy Technology Data Exchange (ETDEWEB)

    Petti, D.A.; Haire, J.C.

    1993-12-01

    The International Thermonuclear Experimental Reactor (ITER) is the first fusion machine that will have sufficient decay heat and activation product inventory to pose potential nuclear safety concerns. As a result, nuclear safety and environmental issues will be much more important in the approval process for the design, siting, construction, and operation of ITER in the United States than previous fusion devices, such as the Tokamak Fusion Test Reactor. The purpose of this report is (a) to provide an overview of the regulatory approval process for a Department of Energy (DOE) nuclear facility; (b) to present the dose limits used by DOE to protect workers, the public, and the environment from the risks of exposure to radiation and hazardous materials; (c) to discuss some key nuclear safety-related issues that must be addressed early in the Engineering Design Activities (EDA) to obtain regulatory approval; and (d) to provide general guidelines to the ITER Joint Central Team (JCT) concerning the development of a regulatory framework for the ITER project.

  15. Aging assessment of reactor instrumentation and protection system components. Aging-related operating experiences

    Energy Technology Data Exchange (ETDEWEB)

    Gehl, A.C.; Hagen, E.W. [Oak Ridge National Lab., TN (United States)

    1992-07-01

    A study of the aging-related operating experiences throughout a five-year period (1984--1988) of six generic instrumentation modules (indicators, sensors, controllers, transmitters, annunciators, and recorders) was performed as a part of the Nuclear Plant Aging Research Program. The effects of aging from operational and environmental stressors were characterized from results depicted in Licensee Event Reports (LERs). The data are graphically displayed as frequency of events per plant year for operating plant ages from 1 to 28 years to determine aging-related failure trend patterns. Three main conclusions were drawn from this study: (1) Instrumentation and control (I&C) modules make a modest contribution to safety-significant events: 17% of LERs issued during 1984--1988 dealt with malfunctions of the six I&C modules studied, and 28% of the LERs dealing with these I&C module malfunctions were aging related (other studies show a range 25--50%); (2) Of the six modules studied, indicators, sensors, and controllers account for the bulk (83%) of aging-related failures; and (3) Infant mortality appears to be the dominant aging-related failure mode for most I&C module categories (with the exception of annunciators and recorders, which appear to fail randomly).

  16. The studies of irradiation hardening of stainless steel reactor internals under proton and xenon irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Xu, Chaoliang; Zhang, Lu; Qian, Wangjie; Mei, Jinna; Liu, Xiang Bing [Suzhou Nuclear Power Research Institute, Suzuhou (China)

    2016-06-15

    Specimens of stainless steel reactor internals were irradiated with 240 keV protons and 6 MeV Xe ions at room temperature. Nanoindentation constant stiffness measurement tests were carried out to study the hardness variations. An irradiation hardening effect was observed in proton- and Xe-irradiated specimens and more irradiation damage causes a larger hardness increment. The Nix-Gao model was used to extract the bulk-equivalent hardness of irradiation-damaged region and critical indentation depth. A different hardening level under H and Xe irradiation was obtained and the discrepancies of displacement damage rate and ion species may be the probable reasons. It was observed that the hardness of Xe-irradiated specimens saturate at about 2 displacement/atom (dpa), whereas in the case of proton irradiation, the saturation hardness may be more than 7 dpa. This discrepancy may be due to the different damage distributions.

  17. Fast reactor safety: proceedings of the international topical meeting. Volume 2. [R

    Energy Technology Data Exchange (ETDEWEB)

    1985-07-01

    The emphasis of this meeting was on the safety-related aspects of fast reactor design, analysis, licensing, construction, and operation. Relative to past meetings, there was less emphasis on the scientific and technological basis for accident assessment. Because of its broad scope, the meeting attracted 217 attendees from a wide cross section of the design, safety analysis, and safety technology communities. Eight countries and two international organizations were represented. A total of 126 papers were presented, with contributions from the United States, France, Japan, the United Kingdom, Germany, and Italy. Sessions covered in Volume 2 include: safety design concepts; operational transient experiments; analysis of seismic and external events; HCDA-related codes, analysis, and experiments; sodium fires; instrumentation and control/PPS design; whole-core accident analysis codes; and impact of safety design considerations on future LMFBR developments.

  18. Study on the safety and on international developments of small modular reactors (SMR). Final report; Studie zur Sicherheit und zu internationalen Entwicklungen von Small Modular Reactors (SMR). Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Buchholz, Sebastian; Kruessenberg, Anne; Schaffrath, Andreas; Zipper, Reinhard

    2015-05-15

    This report documents the work and results of the project RS1521 Study of Safety and International Development of Small Modular Reactors (SMR). The aims of this study can be summarized as - setting-up of a sound overview on SMR, - identification of essential issues of reactor safety research and future R and D projects, - identification of needs for adaption of system codes of GRS used in reactor safety research. The sound overview consists of the descriptions of in total 69 SMR (Small and Medium Sized Rector) concepts (32 light water reactors (LWR), 22 liquid metal cooled reactors (LMR), 2 heavy water reactors, 9 gas cooled reactors (GCR) and 4 molten salt reactors (MSR)). It provides information about the core, the cooling circuits and the safety systems. The quality of the given specifications depends on their availability and public accessibility. Using the safety requirements for nuclear power plants and the fundamental safety functions, the safety relevant issues of the described SMR concepts were identified. The systems and measures used in the safety requirements were summarized in table form. Finally it was evaluated whether these systems and measures can be already simulated with the nuclear simulation chain of GRS and where further code development and validation is necessary. The results of this study can be summarized as follows: Many of the current SMR concepts are based on integral design. Here the main components like steam generators, intermediate heat exchangers or - in case of forced convection core cooling - main cooling pumps are located within the reactor pressure vessel. Most of the SMR fulfil highest safety standards and their safety concepts are mainly based on passive safety systems. The safety of theses reactors is achieved indefinitely without energy supply or additional measures of the operators. Since SMR's aim is not only to produce electricity but also couple them with chemical or physical process plants, the safety aspects of

  19. International Law in a Global Age. A Teacher Handbook.

    Science.gov (United States)

    Croddy, Marshall; Maxey, Phyllis

    This global approach to teaching high school students about international law uses existing curriculum materials from a variety of social studies disciplines to present five major perspectives. Perspective I, "Global Links," focuses on the meaning of citizenship in a global age and the interconnectedness between individuals and the…

  20. International Law in a Global Age. Student Materials.

    Science.gov (United States)

    Croddy, Marshall; Maxey, Phyllis

    This global approach to teaching high school students about international law uses existing curriculum materials from a variety of social studies disciplines to present five major perspectives. Perspective I "Global Links," focuses on the meaning of citizenship in a global age and the interconnectedness between individuals and the…

  1. Gas Reactor International Cooperative program. Pebble bed reactor plant: screening evaluation. Volume 3. Appendix A. Equipment list

    Energy Technology Data Exchange (ETDEWEB)

    1979-11-01

    This report consists of three volumes which describe the design concepts and screening evaluation for a 3000 MW(t) Pebble Bed Reactor Multiplex Plant (PBR-MX). The Multiplex plant produces both electricity and transportable chemical energy via the thermochemical pipeline (TCP). The evaluation was limited to a direct cycle plant which has the steam generators and steam reformers in the primary circuit. Volume 1 reports the overall plant and reactor system and was prepared by the General Electric Company. Core scoping studies were performed which evaluated the effects of annular and cylindrical core configurations, radial blanket zones, burnup, and ball heavy metal loadings. The reactor system, including the PCRV, was investigated for both the annular and cylindrical core configurations. Volume 3 is an Appendix containing the equipment list for the plant and was also prepared by United Engineers and Constructors, Inc. It tabulates the major components of the plant and describes each in terms of quantity, type, orientation, etc., to provide a basis for cost estimation.

  2. Aging biofilm from a full-scale moving bed biofilm reactor: characterization and enzymatic treatment study.

    Science.gov (United States)

    Huang, Hui; Ren, Hongqiang; Ding, Lili; Geng, Jinju; Xu, Ke; Zhang, Yan

    2014-02-01

    Effective removal of aging biofilm deserves to receive more attention. This study aimed to characterized aging biofilm from a full-scale moving bed biofilm reactor treating pharmaceutical wastewater and evaluate the hydrolysis effects of biofilm by different enzymatic treatments. Results from FTIR and biochemical composition analyses showed that it was a predominately organic-based biofilm with the ratio of total protein (PN) to polysaccharide (PS) of 20.17. A reticular structure of extracellular polymeric matrix (EPM) with filamentous bacteria as the skeleton was observed on the basal layer through SEM-EDS test. Among the four commercial proteases and amylases from Genencor®, proteases were shown to have better performances than amylases either on the removal of MLSS and PN/MLSS or on DOC (i.e., dissolved organic carbon)/MLSS raising of biofilm pellets. Difference of dynamic fluorescence characteristics of dissolved organic matters after treated by the two proteases indicated distinguishing mechanisms of the treating process.

  3. A New Solar Chemical Reactor with an Internally Circulating Fluidized bed for Direct Irradiation of Reacting Particles

    Energy Technology Data Exchange (ETDEWEB)

    Kodama, T.; Enomoto, S.; Hatamachi, T.; Gokon, N.

    2006-07-01

    Solar thermochemical processes require the development of a high temperature solar reactor operating at 1000-1500 degree celsius, such as solar gasification of coal and the thermal reduction of metal oxides as part of a two-step water splitting cycle. Direct solar energy absorption by reacting particles of coal or metal oxides provides efficient heat transfer directly to the reaction site. The present paper describes a new type of the windowed solar chemical reactor directly illuminating reacting particles in an internally circulating fluidized bed. The reactor body is made of stainless steel having a quartz window on the top as aperture. A draft tube is centrally inserted in the fluidized bed region. Gases such as steam, CO2, or N2 are introduced into the draft tube and annulus regions in the bed separately. The concentrated solar radiation passes downwards TROUGH the window and directly heats the internally circulating fluidized bed of reacting particles. The prototype reactor was constructed in a laboratory scale and demonstrated on CO2 gasification of coal coke using solar-simulated, concentrated visible light from sun-simulator as the energy source. About 12% of the maximum chemical storage efficiency was obtained by the solar-simulated gasification of the coke. This new reactor will be also applied for a two-step water splitting cycle using redox metal-oxide particles. (Author)

  4. Reactor Pressure Vessel Integrity Assessments with the Grizzly Aging Simulation Code

    Energy Technology Data Exchange (ETDEWEB)

    Spencer, Benjamin; Backman, Marie; Hoffman, William; Chakraborty, Pritam

    2015-08-01

    Grizzly is a simulation tool being developed at Idaho National Laboratory (INL) as part of the US Department of Energy’s Light Water Reactor Sustainability program to provide improved safety assessments of systems, components, and structures in nuclear power plants subjected to age-related degradation. Its goal is to provide an improved scientific basis for decisions surrounding license renewal, which would permit operation of commercial nuclear power plants beyond 60 years. Grizzly is based on INL’s MOOSE framework, which enables multiphysics simulations in a parallel computing environment. It will address a wide variety of aging issues in nuclear power plant systems, components, and structures, modelling both the aging processes and the ability of age-degraded components to perform safely. The reactor pressure vessel (RPV) was chosen as the initial application for Grizzly. Grizzly solves tightly coupled equations of heat conduction and solid mechanics to simulate the global response of the RPV to accident conditions, and uses submodels to represent regions with pre-existing flaws. Domain integrals are used to calculate stress intensity factors on those flaws. A physically based empirical model is used to evaluate material embrittlement, and is used to evaluate whether crack growth would occur. Grizzly can represent the RPV in 2D or 3D, allowing it to evaluate effects that require higher dimensionality models to capture. Work is underway to use lower length scale models of material evolution to inform engineering models of embrittlement. This paper demonstrates an application of Grizzly to RPV failure assessment, and summarizes on-going work.

  5. Gas Reactor International Cooperative program. Pebble bed reactor plant: screening evaluation. Volume 2. Conceptual balance of plant design

    Energy Technology Data Exchange (ETDEWEB)

    1979-11-01

    This report consists of three volumes which describe the design concepts and screening evaluation for a 3000 MW(t) Pebble Bed Reactor Multiplex Plant (PBR-MX). The Multiplex plant produces both electricity and transportable chemical energy via the thermochemical pipeline (TCP). The evaluation was limited to a direct cycle plant which has the steam generators and steam reformers in the primary circuit. This volume describes the conceptual balance-of-plant (BOP) design and was prepared by United Engineers and Constructors, Inc. of Philadelphia, Pennsylvania. The major emphasis of the BOP study was a preliminary design of an overall plant to provide a basis for future studies.

  6. Simulation model and methodology for calculating the damage by internal radiation in a PWR reactor; Modelo de simulacion y metodologia para el calculo del dano por irradiacion en los internos de un reactor PWR

    Energy Technology Data Exchange (ETDEWEB)

    Cadenas Mendicoa, A. M.; Benito Hernandez, M.; Barreira Pereira, P.

    2012-07-01

    This study involves the development of the methodology and three-dimensional models to estimate the damage to the vessel internals of a commercial PWR reactor from irradiation history of operating cycles.

  7. Ageing implementation and refurbishment development at the IEA-R1 nuclear research reactor: a 15 years experience

    Energy Technology Data Exchange (ETDEWEB)

    Cardenas, Jose Patricio N.; Ricci Filho, Walter; Carvalho, Marcos R. de; Berretta, Jose Roberto; Marra Neto, Adolfo, E-mail: ahiru@ipen.b, E-mail: wricci@ipen.b, E-mail: carvalho@ipen.b, E-mail: jrretta@ipen.b, E-mail: amneto@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    IPEN (Instituto de Pesquisas Energeticas e Nucleares) is a nuclear research center established into the Secretary of Science and Technology from the government of the state of Sao Paulo, and administered both technically and financially by Comissao Nacional de Energia Nuclear (CNEN), a federal government organization under the Ministry of Science and Technology. The institute is located inside the campus of the University of Sao Paulo, Sao Paulo city, Brazil. One of major nuclear facilities at IPEN is the IEA-R1 nuclear research reactor. It is the unique Brazilian research reactor with substantial power level suitable for application with research in physics, chemistry, biology and engineering, as well as radioisotope production for medical and other applications. Designed and built by Babcok-Wilcox, in accordance with technical specifications established by the Brazilian Nuclear Energy Commission, and financed by the US Atoms for Peace Program, it is a swimming pool type reactor, moderated and cooled by light water and uses graphite and beryllium as reflector elements. The first criticality was achieved on September 16, 1957 and the reactor is currently operating at 4.0 MW on a 64h per week cycle. Since 1996, an IEA-R1 reactor ageing study was established at the Research Reactor Center (CRPq) related with general deterioration of components belonging to some operational systems, as cooling towers from secondary cooling system, piping and pumps, sample irradiation devices, radiation monitoring system, fuel elements, rod drive mechanisms, nuclear and process instrumentation and safety operational system. Although basic structures are almost the same as the original design, several improvements and modifications in components, systems and structures had been made along reactor life. This work aims to show the development of the ageing program in the IEA-R1 reactor and the upgrading (modernization) that was carried out, concerning several equipment and system in the

  8. Status of advanced fuel candidates for Sodium Fast Reactor within the Generation IV International Forum

    Science.gov (United States)

    Delage, F.; Carmack, J.; Lee, C. B.; Mizuno, T.; Pelletier, M.; Somers, J.

    2013-10-01

    The main challenge for fuels for future Sodium Fast Reactor systems is the development and qualification of a nuclear fuel sub-assembly which meets the Generation IV International Forum goals. The Advanced Fuel project investigates high burn-up minor actinide bearing fuels as well as claddings and wrappers to withstand high neutron doses and temperatures. The R&D outcome of national and collaborative programs has been collected and shared between the AF project members in order to review the capability of sub-assembly material and fuel candidates, to identify the issues and select the viable options. Based on historical experience and knowledge, both oxide and metal fuels emerge as primary options to meet the performance and the reliability goals of Generation IV SFR systems. There is a significant positive experience on carbide fuels but major issues remain to be overcome: strong in-pile swelling, atmosphere required for fabrication as well as Pu and Am losses. The irradiation performance database for nitride fuels is limited with longer term R&D activities still required. The promising core material candidates are Ferritic/Martensitic (F/M) and Oxide Dispersed Strengthened (ODS) steels.

  9. Bulk-bronzied graphites for plasma-facing components in ITER (International Thermonuclear Experimental Reactor)

    Energy Technology Data Exchange (ETDEWEB)

    Hirooka, Y.; Conn, R.W.; Doerner, R.; Khandagle, M. (California Univ., Los Angeles, CA (USA). Inst. of Plasma and Fusion Research); Causey, R.; Wilson, K. (Sandia National Labs., Livermore, CA (USA)); Croessmann, D.; Whitley, J. (Sandia National Labs., Albuquerque, NM (USA)); Holland, D.; Smolik, G. (Idaho National Engineering Lab., Idaho Falls, ID (USA)); Matsuda, T.; Sogabe, T. (Toyo Tanso Co. Ltd., O

    1990-06-01

    Newly developed bulk-boronized graphites and boronized C-C composites with a total boron concentration ranging from 1 wt % to 30 wt % have been evaluated as plasma-facing component materials for the International Thermonuclear Experimental Reactor (ITER). Bulk-boronized graphites have been bombarded with high-flux deuterium plasmas at temperatures between 200 and 1600{degree}C. Plasma interaction induced erosion of bulk-boronized graphites is observed to be a factor of 2--3 smaller than that of pyrolytic graphite, in regimes of physical sputtering, chemical sputtering and radiation enhanced sublimation. Postbombardment thermal desorption spectroscopy indicates that bulk-boronized graphites enhance recombinative desorption of deuterium, which leads to a suppression of the formation of deuterocarbon due to chemical sputtering. The tritium inventory in graphite has been found to decrease by an order of magnitude due to 10 wt % bulk-boronization at temperatures above 1000{degree}C. The critical heat flux to induce cracking for bulk-boronized graphites has been found to be essentially the same as that for non-boronized graphites. Also, 10 wt % bulk-boronization of graphite hinders air oxidation nearly completely at 800{degree}C and reduces the steam oxidation rate by a factor of 2--3 at around 1100 and 1350{degree}C. 38 refs., 5 figs.

  10. Economic impacts on the United States of siting decisions for the international thermonuclear experimental reactor

    Energy Technology Data Exchange (ETDEWEB)

    Peerenboom, J.P.; Hanson, M.E.; Huddleston, J.R. [and others

    1996-08-01

    This report presents the results of a study that examines and compares the probable short-term economic impacts of the International Thermonuclear Experimental Reactor (ITER) on the United States (U.S.) if (1) ITER were to be sited in the U.S., or (2) ITER were to be sited in one of the other countries that, along with the U.S., is currently participating in the ITER program. Life-cycle costs associated with ITER construction, operation, and decommissioning are analyzed to assess their economic impact. A number of possible U.S. host and U.S. non-host technology and cost-sharing arrangements with the other ITER Parties are examined, although cost-sharing arrangements and the process by which the Parties will select a host country and an ITER site remain open issues. Both national and local/regional economic impacts, as measured by gross domestic product, regional output, employment, net exports, and income, are considered. These impacts represent a portion of the complex, interrelated set of economic considerations that characterize U.S. host and U.S. non-host participation in ITER. A number of other potentially important economic and noneconomic considerations are discussed qualitatively.

  11. Antenna design for fast ion collective Thomson scattering diagnostic for the international thermonuclear experimental reactor.

    Science.gov (United States)

    Leipold, F; Furtula, V; Salewski, M; Bindslev, H; Korsholm, S B; Meo, F; Michelsen, P K; Moseev, D; Nielsen, S K; Stejner, M

    2009-09-01

    Fast ion physics will play an important role for the international thermonuclear experimental reactor (ITER), where confined alpha particles will affect and be affected by plasma dynamics and thereby have impacts on the overall confinement. A fast ion collective Thomson scattering (CTS) diagnostic using gyrotrons operated at 60 GHz will meet the requirements for spatially and temporally resolved measurements of the velocity distributions of confined fast alphas in ITER by evaluating the scattered radiation (CTS signal). While a receiver antenna on the low field side of the tokamak, resolving near perpendicular (to the magnetic field) velocity components, has been enabled, an additional antenna on the high field side (HFS) would enable measurements of near parallel (to the magnetic field) velocity components. A compact design solution for the proposed mirror system on the HFS is presented. The HFS CTS antenna is located behind the blankets and views the plasma through the gap between two blanket modules. The viewing gap has been modified to dimensions 30x500 mm(2) to optimize the CTS signal. A 1:1 mock-up of the HFS mirror system was built. Measurements of the beam characteristics for millimeter-waves at 60 GHz used in the mock-up agree well with the modeling.

  12. Microstructure evolution and degradation mechanisms of reactor internal steel irradiated with heavy ions

    Science.gov (United States)

    Borodin, O. V.; Bryk, V. V.; Kalchenko, A. S.; Parkhomenko, A. A.; Shilyaev, B. A.; Tolstolutskaya, G. D.; Voyevodin, V. N.

    2009-03-01

    Structure evolution and degradation mechanisms during irradiation of 18Cr-10Ni-Ti steel (material of VVER-1000 reactor internals are investigated). Using accelerator irradiations with Cr3+ and Ar+ ions allowed studying effects of dose rate, different initial structure state and implanted ions on features of structure evolution and main mechanisms of degradation including low temperature swelling and embrittlement of the 18Cr-10Ni-Ti steel. It is shown that differences in dose rate at most irradiation temperatures mainly exert their influence on the duration of the swelling transient regime. Calculations of possible transmutation products during irradiation of this steel in a VVER-1000 spectrum were performed. It is shown that gaseous atoms (He and H), which are generated simultaneously with radiation defects, stabilize the elements of radiation microstructure and influence the swelling. The nature of deformation under different temperatures of irradiation and of mechanical testing is investigated. It is shown that the temperature sensitivity of swelling behaviour in the investigated steel, with different initial structures can be connected with the dynamic behaviour of point defect sinks.

  13. Insights for aging management of light water reactor components: Metal containments. Volume 5

    Energy Technology Data Exchange (ETDEWEB)

    Shah, V.N.; Sinha, U.P. [EG and G Idaho, Inc., Idaho Falls, ID (United States); Smith, S.K. [Ogden Environmental and Energy Services, Southfield, MI (United States)

    1994-03-01

    This report evaluates the available technical information and field experience related to management of aging damage to light water reactor metal containments. A generic aging management approach is suggested for the effective and comprehensive aging management of metal containments to ensure their safe operation. The major concern is corrosion of the embedded portion of the containment vessel and detection of this damage. The electromagnetic acoustic transducer and half-cell potential measurement are potential techniques to detect corrosion damage in the embedded portion of the containment vessel. Other corrosion-related concerns include inspection of corrosion damage on the inaccessible side of BWR Mark I and Mark II containment vessels and corrosion of the BWR Mark I torus and emergency core cooling system piping that penetrates the torus, and transgranular stress corrosion cracking of the penetration bellows. Fatigue-related concerns include reduction in the fatigue life (a) of a vessel caused by roughness of the corroded vessel surface and (b) of bellows because of any physical damage. Maintenance of surface coatings and sealant at the metal-concrete interface is the best protection against corrosion of the vessel.

  14. On the lifetime of the first mirrors in the diagnostic systems of the international thermonuclear experimental reactor

    OpenAIRE

    De Temmerman, Gregory

    2006-01-01

    Plasma diagnostic systems will be necessary tools for the future success of the International Thermonuclear Experimental Reactor (ITER) both to better understand the physics involved in magnetically confined burning plasma and for the protection of the device in case of disruptions etc. In contrast to conditions in today’s tokamaks, a high level of radiation and neutrons is expected in ITER. To reduce the extent of the possible neutron leakage and to protect the optical compone...

  15. ITER: The International Thermonuclear Experimental Reactor and the nuclear weapons proliferation implications of thermonuclear-fusion energy systems

    OpenAIRE

    Gsponer, Andre; Hurni, Jean-Pierre

    2004-01-01

    This report contains two parts: (1) A list of "points" highlighting the strategic-political and military-technical reasons and implications of the very probable siting of ITER (the International Thermonuclear Experimental Reactor) in Japan, which should be confirmed sometimes in early 2004. (2) A technical analysis of the nuclear weapons proliferation implications of inertial- and magnetic-confinement fusion systems substantiating the technical points highlighted in the first part, and showin...

  16. Proceedings of the 4th International Symposium on Material Testing Reactors; December 5-9, 2011, Oarai, Japan

    OpenAIRE

    石原 正博; 鈴木 雅秀

    2012-01-01

    This report is the Proceedings of the 4th International Symposium on Materials Testing Reactors hosted by Japan Atomic Energy Agency (JAEA). The 4th symposium was originally scheduled to be held INVAP in Argentina. However, the aftermath of volcanic explosion at Chili forced the symposium to change place. Total 111 participants attended from Argentina, Belgium, France, Germany, Indonesia, Malaysia, Korea, South Africa, Switzerland, the United State and Japan. This symposium addressed the gene...

  17. Cable aging and condition monitoring of radiation resistant nano-dielectrics in advanced reactor applications

    Energy Technology Data Exchange (ETDEWEB)

    Duckworth, Robert C [ORNL; Aytug, Tolga [ORNL; Paranthaman, Mariappan Parans [ORNL; Kidder, Michelle [ORNL; Polyzos, Georgios [ORNL; Leonard, Keith J [ORNL

    2015-01-01

    Cross-linked polyethylene (XLPE) nanocomposites have been developed in an effort to improve cable insulation lifetime to serve in both instrument cables and auxiliary power systems in advanced reactor applications as well as to provide an alternative for new or retro-fit cable insulation installations. Nano-dielectrics composed of different weight percentages of MgO & SiO2 have been subjected to radiation at accumulated doses approaching 20 MRad and thermal aging temperatures exceeding 100 C. Depending on the composition, the performance of the nanodielectric insulation was influenced, both positively and negatively, when quantified with respect to its electrical and mechanical properties. For virgin unradiated or thermally aged samples, XLPE nanocomposites with 1wt.% SiO2 showed improvement in breakdown strength and reduction in its dissipation factor when compared to pure undoped XLPE, while XLPE 3wt.% SiO2 resulted in lower breakdown strength. When aged in air at 120 C, retention of electrical breakdown strength and dissipation factor was observed for XLPE 3wt.% MgO nanocomposites. Irrespective of the nanoparticle species, XLPE nanocomposites that were gamma irradiated up to the accumulated dose of 18 MRad showed a significant drop in breakdown strength especially for particle concentrations greater than 3 wt.%. Additional attenuated total reflectance Fourier transform infrared (ATR-FTIR) spectroscopy measurements suggest changes in the structure of the XLPE SiO2 nanocomposites associated with the interaction of silicon and oxygen. Discussion on the relevance of property changes with respect to cable aging and condition monitoring is presented.

  18. Drinking water treatment using a submerged internal-circulation membrane coagulation reactor coupled with permanganate oxidation.

    Science.gov (United States)

    Zhang, Zhongguo; Liu, Dan; Qian, Yu; Wu, Yue; He, Peiran; Liang, Shuang; Fu, Xiaozheng; Li, Jiding; Ye, Changqing

    2017-06-01

    A submerged internal circulating membrane coagulation reactor (MCR) was used to treat surface water to produce drinking water. Polyaluminum chloride (PACl) was used as coagulant, and a hydrophilic polyvinylidene fluoride (PVDF) submerged hollow fiber microfiltration membrane was employed. The influences of trans-membrane pressure (TMP), zeta potential (ZP) of the suspended particles in raw water, and KMnO4 dosing on water flux and the removal of turbidity and organic matter were systematically investigated. Continuous bench-scale experiments showed that the permeate quality of the MCR satisfied the requirement for a centralized water supply, according to the Standards for Drinking Water Quality of China (GB 5749-2006), as evaluated by turbidity (water flux, the removal of turbidity, TOC and dissolved organic carbon (DOC) in the raw water also increased with increasing TMP in the range of 0.01-0.05MPa. High ZP induced by PACl, such as 5-9mV, led to an increase in the number of fine and total particles in the MCR, and consequently caused serious membrane fouling and high permeate turbidity. However, the removal of TOC and DOC increased with increasing ZP. A slightly positive ZP, such as 1-2mV, corresponding to charge neutralization coagulation, was favorable for membrane fouling control. Moreover, dosing with KMnO4 could further improve the removal of turbidity and DOC, thereby mitigating membrane fouling. The results are helpful for the application of the MCR in producing drinking water and also beneficial to the research and application of other coagulation and membrane separation hybrid processes. Copyright © 2016. Published by Elsevier B.V.

  19. Sulfide-oxidizing bacteria establishment in an innovative microaerobic reactor with an internal silicone membrane for sulfur recovery from wastewater.

    Science.gov (United States)

    Valdés, F; Camiloti, P R; Rodriguez, R P; Delforno, T P; Carrillo-Reyes, J; Zaiat, M; Jeison, D

    2016-06-01

    A novel bioreactor, employing a silicone membrane for microaeration, was studied for partial sulfide oxidation to elemental sulfur. The objective of this study was to assess the feasibility of using an internal silicone membrane reactor (ISMR) to treat dissolved sulfide and to characterize its microbial community. The ISMR is an effective system to eliminate sulfide produced in anaerobic reactors. Sulfide removal efficiencies reached 96 % in a combined anaerobic/microaerobic reactor and significant sulfate production did not occur. The oxygen transfer was strongly influenced by air pressure and flow. Pyrosequencing analysis indicated various sulfide-oxidizing bacteria (SOB) affiliated to the species Acidithiobacillus thiooxidans, Sulfuricurvum kujiense and Pseudomonas stutzeri attached to the membrane and also indicated similarity between the biomass deposited on the membrane wall and the biomass drawn from the material support, supported the establishment of SOB in an anaerobic sludge under microaerobic conditions. Furthermore, these results showed that the reactor configuration can develop SOB under microaerobic conditions and can improve and reestablish the sulfide conversion to elemental sulfur.

  20. Effect of particle pinch on the fusion performance and profile features of an international thermonuclear experimental reactor-like fusion reactor

    Science.gov (United States)

    Wang, Shijia; Wang, Shaojie

    2015-04-01

    The evolution of the plasma temperature and density in an international thermonuclear experimental reactor (ITER)-like fusion device has been studied by numerically solving the energy transport equation coupled with the particle transport equation. The effect of particle pinch, which depends on the magnetic curvature and the safety factor, has been taken into account. The plasma is primarily heated by the alpha particles which are produced by the deuterium-tritium fusion reactions. A semi-empirical method, which adopts the ITERH-98P(y,2) scaling law, has been used to evaluate the transport coefficients. The fusion performances (the fusion energy gain factor, Q) similar to the ITER inductive scenario and non-inductive scenario (with reversed magnetic shear) are obtained. It is shown that the particle pinch has significant effects on the fusion performance and profiles of a fusion reactor. When the volume-averaged density is fixed, particle pinch can lower the pedestal density by ˜30 % , with the Q value and the central pressure almost unchanged. When the particle source or the pedestal density is fixed, the particle pinch can significantly enhance the Q value by 60 % , with the central pressure also significantly raised.

  1. A longitudinal investigation of children internationally adopted at school age.

    Science.gov (United States)

    Helder, Emily J; Mulder, Elizabeth; Gunnoe, Marjorie Linder

    2016-01-01

    Most existing research on children adopted internationally has focused on those adopted as infants and toddlers. The current study longitudinally tracked several outcomes, including cognitive, behavioral, emotional, attachment, and family functioning, in 25 children who had been internationally adopted at school age (M = 7.7 years old at adoption, SD = 3.4, range = 4–15 years). We examined the incidence of clinically significant impairments, significant change in outcomes over the three study points, and variables that predicted outcomes over time. Clinically significant impairments in sustained attention, full-scale intelligence, reading, language, executive functioning, externalizing problems, and parenting stress were common, with language and executive functioning impairments present at higher levels in the current study compared with past research focusing on children adopted as infants and toddlers. Over the three study points, significant improvements across most cognitive areas and attachment functioning were observed, though significant worsening in executive functioning and internalizing problems was present. Adoptive family-specific variables, such as greater maternal education, smaller family size, a parenting approach that encouraged age-expected behaviors, home schooling, and being the sole adopted child in the family were associated with greater improvement across several cognitive outcomes. In contrast, decreased parenting stress was predicted by having multiple adopted children and smaller family sizes were associated with greater difficulties with executive functioning. Child-specific variables were also linked to outcomes, with girls displaying worse attachment and poorer cognitive performance and with less time in orphanage care resulting in greater adoption success. Implications for future research and clinical applications are discussed.

  2. Growth and Expansion of the International Criticality Safety Benchmark Evaluation Project and the Newly Organized International Reactor Physics Experiment Evaluation Project

    Energy Technology Data Exchange (ETDEWEB)

    J. Blair Briggs; Lori Scott; Yolanda Rugama; Enrico Satori

    2007-05-01

    Since ICNC 2003, the International Criticality Safety Benchmark Evaluation Project (ICSBEP) has continued to expand its efforts and broaden its scope. Criticality-alarm / shielding type benchmarks and fundamental physics measurements that are relevant to criticality safety applications are not only included in the scope of the project, but benchmark data are also included in the latest version of the handbook. A considerable number of improvements have been made to the searchable database, DICE and the criticality-alarm / shielding benchmarks and fundamental physics measurements have been included in the database. There were 12 countries participating on the ICSBEP in 2003. That number has increased to 18 with recent contributions of data and/or resources from Brazil, Czech Republic, Poland, India, Canada, and China. South Africa, Germany, Argentina, and Australia have been invited to participate. Since ICNC 2003, the contents of the “International Handbook of Evaluated Criticality Safety Benchmark Experiments” have increased from 350 evaluations (28,000 pages) containing benchmark specifications for 3070 critical or subcritical configurations to 442 evaluations (over 38,000 pages) containing benchmark specifications for 3957 critical or subcritical configurations, 23 criticality-alarm-placement / shielding configurations with multiple dose points for each, and 20 configurations that have been categorized as fundamental physics measurements that are relevant to criticality safety applications in the 2006 Edition of the ICSBEP Handbook. Approximately 30 new evaluations and 250 additional configurations are expected to be added to the 2007 Edition of the Handbook. Since ICNC 2003, a reactor physics counterpart to the ICSBEP, The International Reactor Physics Experiment Evaluation Project (IRPhEP) was initiated. Beginning in 1999, the IRPhEP was conducted as a pilot activity by the by the Organization of Economic Cooperation and Development (OECD) Nuclear Energy

  3. Wear plates control rod guide tubes top internal reactor vessel C. N. VANDELLOS II; Desgaste placas tubos guia barras de control interno superior vasija del reactor C.N. Vandellos II

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2010-07-01

    The guide tubes for control rods forming part of the upper internals of the reactor vessel, its function is to guide the control rod to permit its insertion in the reactor core. These guide tubes are suspended from the upper support plate which are fixed by bolts and extending to the upper core plate which is fastened by clamping bolts (split pin) to prevent lateral displacement of the guide tubes, while allowing axial expansion.

  4. Impact of beryllium reflector ageing on Safari–1 reactor core parameters / L.E. Moloko

    OpenAIRE

    Moloko, Lesego Ernest

    2011-01-01

    The build–up of 6Li and 3He, that is, the strong thermal neutron absorbers or the so called "neutron poisons", in the beryllium reflector changes the physical characteristics of the reactor, such as reactivity, neutron spectra, neutron flux level, power distribution, etc.; furthermore,gaseous isotopes such as 3H and 4He induce swelling and embrittlement of the reflector. The SAFARI–1 research reactor, operated by Necsa at Pelindaba in South Africa, uses a beryllium reflector on...

  5. Coupling CFAST fire modeling and SAPHIRE probabilistic assessment software for internal fire safety evaluation of a typical TRIGA research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Safaei Arshi, Saiedeh [School of Engineering, Shiraz University, 71348-51154 Shiraz (Iran, Islamic Republic of); Nematollahi, Mohammadreza, E-mail: nema@shirazu.ac.i [School of Engineering, Shiraz University, 71348-51154 Shiraz (Iran, Islamic Republic of); Safety Research Center of Shiraz University, 71348-51154 Shiraz (Iran, Islamic Republic of); Sepanloo, Kamran [Safety Research Center of Shiraz University, 71348-51154 Shiraz (Iran, Islamic Republic of)

    2010-03-15

    Due to the significant threat of internal fires for the safety operation of nuclear reactors, presumed fire scenarios with potential hazards for loss of typical research reactor safety functions are analyzed by coupling CFAST fire modeling and SAPHIRE probabilistic assessment software. The investigations show that fire hazards associated with electrical cable insulation, lubricating oils, diesel, electrical equipment and carbon filters may lead to unsafe situations called core damage states. Using system-specific event trees, the occurrence frequency of core damage states after the occurrence of each possible fire scenario in critical fire compartments is evaluated. Probability that the fire ignited in the given fire compartment will burn long enough to cause the extent of damage defined by each fire scenario is calculated by means of detection-suppression event tree. As a part of detection-suppression event trees quantification, and also for generating the necessary input data for evaluating the frequency of core damage states by SAPHIRE 7.0 software, CFAST fire modeling software is applied. The results provide a probabilistic measure of the quality of existing fire protection systems in order to maintain the reactor at a reasonable safety level.

  6. Aging of the containment pressure boundary in light-water reactor plants

    Energy Technology Data Exchange (ETDEWEB)

    Naus, D.J.; Oland, C.B. [Oak Ridge National Lab., TN (United States); Ellingwood, B.R. [Johns Hopkins Univ., Baltimore, MD (United States)] [and others

    1997-01-01

    Research is being conducted by the Oak Ridge National Laboratory to address aging of the containment pressure boundary in light-water reactor plants. The objectives of this work are to (1) identify the significant factors related to occurrence of corrosion, efficacy of inspection, and structural capacity reduction of steel containments and liners of concrete containments, and to make recommendations on use of risk models in regulatory decisions; (2) provide NRC reviewers a means of establishing current structural capacity margins for steel containments, and concrete containments as limited by liner integrity; and (3) provide recommendations, as appropriate, on information to be requested of licensees for guidance that could be utilized by NRC reviewers in assessing the seriousness of reported incidences of containment degradation. In meeting these objectives research is being conducted in two primary task areas - pressure boundary condition assessment and root-cause resolution practices, and reliability-based condition assessments. Under the first task area a degradation assessment methodology was developed for use in characterizing the in-service condition of metal and concrete containment pressure boundary components and quantifying the amount of damage that is present. An assessment of available destructive and nondestructive techniques for examining steel containments and liners is ongoing. Under the second task area quantitative structural reliability analysis methods are being developed for application to degraded metallic pressure boundaries to provide assurances that they will be able to withstand future extreme loads during the desired service period with a level of reliability that is sufficient for public safety. To date, mathematical models that describe time-dependent changes in steel due to aggressive environmental factors have been identified, and statistical data supporting their use in time-dependent reliability analysis have been summarized.

  7. 78 FR 70076 - Aging Management of Internal Surfaces, Fire Water Systems, Atmospheric Storage Tanks, and...

    Science.gov (United States)

    2013-11-22

    ... COMMISSION Aging Management of Internal Surfaces, Fire Water Systems, Atmospheric Storage Tanks, and... Guidance (LR-ISG), LR-ISG-2012-02, ``Aging Management of Internal Surfaces, Fire Water Systems, Atmospheric... aging management programs (AMPs), aging management review (AMR) items, and definitions in NUREG-...

  8. 78 FR 21980 - Aging Management of Internal Surfaces, Service Level III and Other Coatings, Atmospheric Storage...

    Science.gov (United States)

    2013-04-12

    ... COMMISSION Aging Management of Internal Surfaces, Service Level III and Other Coatings, Atmospheric Storage...-2012-02, ``Aging Management of Internal Surfaces, Service Level III and Other Coatings, Atmospheric... aging management programs (AMP) and aging management review (AMR) items in NUREG-1801, Revision 2...

  9. Method for producing components with internal architectures, such as micro-channel reactors, via diffusion bonding sheets

    Science.gov (United States)

    Alman, David E.; Wilson, Rick D.; Davis, Daniel L.

    2011-03-08

    This invention relates to a method for producing components with internal architectures, and more particularly, this invention relates to a method for producing structures with microchannels via the use of diffusion bonding of stacked laminates. Specifically, the method involves weakly bonding a stack of laminates forming internal voids and channels with a first generally low uniaxial pressure and first temperature such that bonding at least between the asperites of opposing laminates occurs and pores are isolated in interfacial contact areas, followed by a second generally higher isostatic pressure and second temperature for final bonding. The method thereby allows fabrication of micro-channel devices such as heat exchangers, recuperators, heat-pumps, chemical separators, chemical reactors, fuel processing units, and combustors without limitation on the fin aspect ratio.

  10. 10th international conference on gas-liquid and gas-liquid-solid reactor engineering preface

    OpenAIRE

    J.A. Teixeira; Vicente, A. A.; Middelberg, A.P.J

    2011-01-01

    Following the success of the nine previous conferences on Gas–Liquid and Gas–Liquid–Solid reactor Engineering which were held at Columbus, OH, USA (1992), Cambridge, UK (1995), Kanagawa, Japan (1997), Delft, The Netherlands (1999), Melbourne, Australia (2001), Vancouver, Canada (2003), Strasbourg, France (2005), New Delhi, India (2007) and Montreal, Canada (2009) the tenth conference with the same theme is being held in Braga, Portugal, from 26 to 29 June 2011. This conference will cover all ...

  11. Swelling in light water reactor internal components: Insights from computational modeling

    Energy Technology Data Exchange (ETDEWEB)

    Stoller, Roger E. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Barashev, Alexander V. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Univ. of Tennessee, Knoxville, TN (United States); Golubov, Stanislav I. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-08-01

    A modern cluster dynamics model has been used to investigate the materials and irradiation parameters that control microstructural evolution under the relatively low-temperature exposure conditions that are representative of the operating environment for in-core light water reactor components. The focus is on components fabricated from austenitic stainless steel. The model accounts for the synergistic interaction between radiation-produced vacancies and the helium that is produced by nuclear transmutation reactions. Cavity nucleation rates are shown to be relatively high in this temperature regime (275 to 325°C), but are sensitive to assumptions about the fine scale microstructure produced under low-temperature irradiation. The cavity nucleation rates observed run counter to the expectation that void swelling would not occur under these conditions. This expectation was based on previous research on void swelling in austenitic steels in fast reactors. This misleading impression arose primarily from an absence of relevant data. The results of the computational modeling are generally consistent with recent data obtained by examining ex-service components. However, it has been shown that the sensitivity of the model s predictions of low-temperature swelling behavior to assumptions about the primary damage source term and specification of the mean-field sink strengths is somewhat greater that that observed at higher temperatures. Further assessment of the mathematical model is underway to meet the long-term objective of this research, which is to provide a predictive model of void swelling at relevant lifetime exposures to support extended reactor operations.

  12. Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Volume 1, Sessions 1-5

    Energy Technology Data Exchange (ETDEWEB)

    Block, R.C.; Feiner, F. [comps.] [American Nuclear Society, La Grange Park, IL (United States)

    1995-09-01

    This document, Volume 1, includes papers presented at the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-7) September 10--15, 1995 at Saratoga Springs, N.Y. The following subjects are discussed: Progress in analytical and experimental work on the fundamentals of nuclear thermal-hydraulics, the development of advanced mathematical and numerical methods, and the application of advancements in the field in the development of novel reactor concepts. Also combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

  13. International benchmark study of advanced thermal hydraulic safety analysis codes against measurements on IEA-R1 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hainoun, A., E-mail: pscientific2@aec.org.sy [Atomic Energy Commission of Syria (AECS), Nuclear Engineering Department, P.O. Box 6091, Damascus (Syrian Arab Republic); Doval, A. [Nuclear Engineering Department, Av. Cmdt. Luis Piedrabuena 4950, C.P. 8400 S.C de Bariloche, Rio Negro (Argentina); Umbehaun, P. [Centro de Engenharia Nuclear – CEN, IPEN-CNEN/SP, Av. Lineu Prestes 2242-Cidade Universitaria, CEP-05508-000 São Paulo, SP (Brazil); Chatzidakis, S. [School of Nuclear Engineering, Purdue University, West Lafayette, IN 47907 (United States); Ghazi, N. [Atomic Energy Commission of Syria (AECS), Nuclear Engineering Department, P.O. Box 6091, Damascus (Syrian Arab Republic); Park, S. [Research Reactor Design and Engineering Division, Basic Science Project Operation Dept., Korea Atomic Energy Research Institute (Korea, Republic of); Mladin, M. [Institute for Nuclear Research, Campului Street No. 1, P.O. Box 78, 115400 Mioveni, Arges (Romania); Shokr, A. [Division of Nuclear Installation Safety, Research Reactor Safety Section, International Atomic Energy Agency, A-1400 Vienna (Austria)

    2014-12-15

    Highlights: • A set of advanced system thermal hydraulic codes are benchmarked against IFA of IEA-R1. • Comparative safety analysis of IEA-R1 reactor during LOFA by 7 working teams. • This work covers both experimental and calculation effort and presents new out findings on TH of RR that have not been reported before. • LOFA results discrepancies from 7% to 20% for coolant and peak clad temperatures are predicted conservatively. - Abstract: In the framework of the IAEA Coordination Research Project on “Innovative methods in research reactor analysis: Benchmark against experimental data on neutronics and thermal hydraulic computational methods and tools for operation and safety analysis of research reactors” the Brazilian research reactor IEA-R1 has been selected as reference facility to perform benchmark calculations for a set of thermal hydraulic codes being widely used by international teams in the field of research reactor (RR) deterministic safety analysis. The goal of the conducted benchmark is to demonstrate the application of innovative reactor analysis tools in the research reactor community, validation of the applied codes and application of the validated codes to perform comprehensive safety analysis of RR. The IEA-R1 is equipped with an Instrumented Fuel Assembly (IFA) which provided measurements for normal operation and loss of flow transient. The measurements comprised coolant and cladding temperatures, reactor power and flow rate. Temperatures are measured at three different radial and axial positions of IFA summing up to 12 measuring points in addition to the coolant inlet and outlet temperatures. The considered benchmark deals with the loss of reactor flow and the subsequent flow reversal from downward forced to upward natural circulation and presents therefore relevant phenomena for the RR safety analysis. The benchmark calculations were performed independently by the participating teams using different thermal hydraulic and safety

  14. CFD analysis of PWR core top and reactor vessel upper plenum internal subdomain models

    Energy Technology Data Exchange (ETDEWEB)

    Kao, Min-Tsung; Wu, Chung-Yun [National Tsing Hua University, Hsinchu 30043, Taiwan (China); Chieng, Ching-Chang, E-mail: cchieng@ess.nthu.edu.tw [National Tsing Hua University, Hsinchu 30043, Taiwan (China); Xu Yiban; Yuan Kun; Dzodzo, Milorad; Conner, Michael; Beltz, Steven; Ray, Sumit; Bissett, Teresa [Westinghouse Electric Company, Cranberry Township, PA 16066 (United States)

    2011-10-15

    Highlights: > The paper develops a CFD flow model for upper portion of AP1000 and determines how lateral flow in the top core and upper plenum. > Mesh sensitivities and geometrical modification strategies give the guidelines to reduce the size of overall computation mesh. > Pressure drop measurement data act as a guideline for the mesh selection. > Lateral flows are mainly exiting through upper and lower windows of guide tubes ({approx}81%) and 18% flow through small side gaps. > The interactions between guide tubes and neighboring support column as well as flow characteristic are revealed. - Abstract: One aspect of the Westinghouse AP1000{sup TM} reactor design is the reduction in the number of major components and simplification in manufacturing. One design change relative to current Westinghouse reactors of similar size is that AP1000 reactor vessel has two nozzles/hot legs instead of three. With regard to fuel performance, this design difference creates a different flow field in the reactor vessel upper plenum. The flow exiting from the core and entering the upper plenum must turn toward one of the two outlet nozzles and flow laterally around numerous control rod guide tubes and support columns. Also, below the upper plenum are the upper core plate and the top core region of the 157 fuel assemblies and 69 guidetube assemblies. To determine how the lateral flow in the top of the core and upper plenum compares to the current reactors a CFD model of the flow in the upper portion of the AP1000 reactor vessel was created. Before detailed CFD simulations of the flow in the entire upper plenum and top core regions were performed, conducting local simulations for smaller sections of the domain provided crucial and detailed physical aspects of the flow. These sub-domain models were used to perform mesh sensitivities and to assess what geometrical details may be eliminated from the larger model in order to reduce mesh size and computational requirements. In this paper

  15. Regulatory instrument review: Aging management of LWR cables, containment and basemat, reactor coolant pumps, and motor-operated valves

    Energy Technology Data Exchange (ETDEWEB)

    Werry, E.V.; Somasundaram, S.

    1995-09-01

    The results of Stage 2 of the Regulatory Instrument Review are presented in this volume. Selected regulatory instruments, such as the Code of Federal Regulations (CFR), US Nuclear Regulatory Commission (NRC), Regulatory Guides, and ASME Codes, were investigated to determine the extent to which these regulations apply aging management to selected safety-related components in nuclear power plants. The Regulatory Instrument Review was funded by the NRC under the Nuclear Plant Aging Research (NPAR) program. Stage 2 of the review focused on four safety-related structures and components; namely, cables, containment and basemat, reactor coolant pumps, and motor-operated valves. The review suggests that the primary-emphasis of the regulatory instruments was on the design, construction, start-up, and operation of a nuclear power plant, and that aging issues were primarily addressed after an aging-related problem was recognized. This Stage 2 review confirms the results of the prior review; (see Regulatory Instrument Review: Management of Aging of LWR Major Safety-Related Components NUREG/CR-5490. The observations indicate that the regulations generally address management of age-related degradation indirectly. Specific age-related degradation phenomena frequently are dealt with in bulletins and notices or through generic issues, letters, etc. The major recommendation of this report, therefore, is that the regulatory instruments should more directly and explicitly address the aging phenomenon and the management of the age-related degradation process.

  16. ITER: The International Thermonuclear Experimental Reactor and the nuclear weapons proliferation implications of thermonuclear-fusion energy

    CERN Document Server

    Gsponer, A; Gsponer, Andre; Hurni, Jean-Pierre

    2004-01-01

    This paper contains two parts: (I) A list of "points" highlighting the strategic-political and military-technical reasons and implications of the very probable siting of ITER (the International Thermonuclear Experimental Reactor) in Japan, which should be confirmed sometimes in early 2004. (II) A technical analysis of the nuclear weapons proliferation implications of inertial- and magnetic-confinement fusion systems substantiating the technical points highlighted in the first part, and showing that while full access to the physics of thermonuclear weapons is the main implication of ICF, full access to large-scale tritium technology is the main proliferation impact of MCF. The conclusion of the paper is that siting ITER in a country such as Japan, which already has a large separated-plutonium stockpile, and an ambitious laser-driven ICF program (comparable in size and quality to those of the United States or France) will considerably increase its latent (or virtual) nuclear weapons proliferation status, and fo...

  17. Porous nuclear fuel element with internal skeleton for high-temperature gas-cooled nuclear reactors

    Science.gov (United States)

    Youchison, Dennis L.; Williams, Brian E.; Benander, Robert E.

    2013-09-03

    Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

  18. Preparation, characterization, and photocatalytic studies on anatase nano-TiO{sub 2} at internal air lift circulating photocatalytic reactor

    Energy Technology Data Exchange (ETDEWEB)

    Xu, Hang, E-mail: xhinbj@126.com; Li, Mei; Jun, Zhang

    2013-09-01

    Graphical abstract: The micro morphological structure of the nano-TiO{sub 2} particles was also observed with TEM, as shown in figure. The TEM images clearly exhibited the homogeneous microstructure of particles with a size of around 10–15 nm. - Highlights: • Nano-TiO{sub 2} was prepared by complex techniques of sol–gel, micro-emulsion and solvent thermal. • The size of TiO{sub 2} was nano level and uniformity. • Nano-TiO{sub 2} exhibited high photo-catalytic activity at internal air lift circulating reactor. • The best nano-TiO{sub 2} dosage was obtained. - Abstract: Anatase nano-titania (TiO{sub 2}) powder was prepared by using a sol–gel process mediated in reverse microemulsion combined with a solvent thermal technique. The structures of the obtained TiO{sub 2} were characterized by TG-DSC, XRD, TEM. The photocatalytic decomposition of methylene blue (MB) on nano-TiO{sub 2} was studied by using an internal air lift circulating photocatalytic reactor. The results show that the anatase structure appears in the calcination temperature range of 400–510 °C, while the transformation of anatase into rutile takes place above 510 °C. The homogeneous microstructure of nano-TiO{sub 2} particles was obtained with a size of around 10–15 nm. In the photocatalytic performance, degradation process follows pseudo first order kinetics with different dosages of photocatalyst and initial MB concentrations and optimal TiO{sub 2} dosage is 0.1 g/L with neutral medium.

  19. Studies of the Future Aged. An International Symposium.

    Science.gov (United States)

    Friis, Henning; Sheppard, Harold L., Ed.

    These six papers report on future-oriented studies of the situation of the elderly. "Changing Elderly in a Changing Society: Danish Elderly in the Next Century" (Henning Friis) reports on research dealing with preferences of the future elderly for their life when they grow older. "Aging Effectively: Meeting the Challenge of an Aging World" (J.…

  20. Smoothing internal migration age profiles for comparative research

    Directory of Open Access Journals (Sweden)

    Aude Bernard

    2015-05-01

    Full Text Available Background: Age patterns are a key dimension to compare migration between countries and over time. Comparative metrics can be reliably computed only if data capture the underlying age distribution of migration. Model schedules, the prevailing smoothing method, fit a composite exponential function, but are sensitive to function selection and initial parameter setting. Although non-parametric alternatives exist, their performance is yet to be established. Objective: We compare cubic splines and kernel regressions against model schedules by assessingwhich method provides an accurate representation of the age profile and best performs on metrics for comparing aggregate age patterns. Methods: We use full population microdata for Chile to perform 1,000 Monte-Carlo simulations for nine sample sizes and two spatial scales. We use residual and graphic analysis to assess model performance on the age and intensity at which migration peaks and the evolution of migration age patterns. Results: Model schedules generate a better fit when (1 the expected distribution of the age profile is known a priori, (2 the pre-determined shape of the model schedule adequately describes the true age distribution, and (3 the component curves and initial parameter values can be correctly set. When any of these conditions is not met, kernel regressions and cubic splines offer more reliable alternatives. Conclusions: Smoothing models should be selected according to research aims, age profile characteristics, and sample size. Kernel regressions and cubic splines enable a precise representation of aggregate migration age profiles for most sample sizes, without requiring parameter setting or imposing a pre-determined distribution, and therefore facilitate objective comparison.

  1. Effect of internal diffusional restrictions on the hydrolysis of penicillin G: reactor performance and specific productivity of 6-APA with immobilized penicillin acylase.

    Science.gov (United States)

    Valencia, Pedro; Flores, Sebastián; Wilson, Lorena; Illanes, Andrés

    2011-09-01

    A mathematical model that describes the heterogeneous reaction-diffusion process involved in penicillin G hydrolysis in a batch reactor with immobilized penicillin G acylase is presented. The reaction system includes the bulk liquid phase containing the dissolved substrate (and products) and the solid biocatalyst phase represented by glyoxyl-agarose spherical porous particles carrying the enzyme. The equations consider reaction and diffusion components that are presented in dimensionless form. This is a complex reaction system in which both products of reaction and the substrate itself are inhibitors. The simulation of a batch reactor performance with immobilized penicillin G acylase is presented and discussed for the internal diffusional restrictions impact on effectiveness and productivity. Increasing internal diffusional restrictions, through increasing catalyst particle size and enzyme loading, causes impaired catalyst efficiency expressed in a reduction of effectiveness factor and specific productivity. High penicillin G initial concentrations decrease the impact of internal diffusional restrictions by increasing the mass transfer towards porous catalyst until product inhibition becomes significant over approximately 50 mM of initial penicillin G, where a drop in conversion rate and a maximum in specific productivity are then obtained. Results highlight the relevance of considering internal diffusional restrictions, reactor performance, and productivity analysis for proper catalyst and reactor design.

  2. Varying gestational age patterns in cesarean delivery: An international comparison

    NARCIS (Netherlands)

    Delnord, M.; Blondel, B.; Drewniak, N.; Klungsøyr, K.; Bolumar, F.; Mohangoo, A.; Gissler, M.; Szamotulska, K.; Lack, N.; Nijhuis, J.; Velebil, P.; Sakkeus, L.; Chalmers, J.; Zeitlin, J.; Haidinger, G.; XMartens, G.; Misselwitz, B.; Wenzlaff, P.; Bonham, S.; Jaselioniene, J.; Gatt, M.; Klungsøyr, K.; Barros, H.; Novak, Z.; Gottvall, K.

    2014-01-01

    Background: While international variations in overall cesarean delivery rates are well documented, less information is available for clinical sub-groups. Cesarean data presented by subgroups can be used to evaluate uptake of cesarean reduction policies or to monitor delivery practices for high and

  3. ECO LOGIC INTERNATIONAL GAS-PHASE CHEMICAL REDUCTION PROCESS - THE REACTOR SYSTEM - APPLICATIONS ANALYSIS REPORT

    Science.gov (United States)

    The ELI Eco Logic International Inc. (Eco Logic) process thermally separates organics, then chemically reduces them in a hydrogen atmosphere, converting them to a reformed gas that consists of light hydrocarbons and water. A scrubber treats the reformed gas to remove hydrogen chl...

  4. Effect of pedestal height and internal transport barriers on International Thermonuclear Experimental Reactor target steady state simulations

    Energy Technology Data Exchange (ETDEWEB)

    Rafiq, T.; Kritz, A. H.; Bateman, G. [Department of Physics, Lehigh University, 16 Memorial Drive East, Bethlehem, Pennsylvania 18015 (United States); Kessel, C.; McCune, D. C.; Budny, R. V. [Princeton Plasma Physics Laboratory, Princeton University, Princeton, New Jersey 08540 (United States)

    2011-11-15

    The Tokamak simulation code (TSC) is used to provide initial conditions for predictive TRANSPort and integrated modeling code (PTRANSP) simulations of ITER target steady state scenarios. The PTRANSP simulations are carried out using the new multi-mode (MMM7.1) and the gyro-Landau-fluid (GLF23) transport models. It is found that there are circumstances under which the total fusion power decreases with increasing pedestal temperature height. When the total current (from magnetic axis to plasma edge) is fixed, an increased fraction of the current is concentrated in the pedestal region as the pedestal height is increased. As a consequence of the fixed total current, this results a smaller fraction of the current in the core plasma and, consequently, lower energy confinement. In previous simulations of ITER, in which the fusion power increased with increasing pedestal temperature height, the plasma current from the top of the pedestal to the magnetic axis was held fixed independent of the pedestal temperature. Simulations presented in this paper also indicate that improvement in fusion power production occurs when the lower hybrid current drive is replaced with electron cyclotron current drive. Again, the improvement results from the redistribution of plasma current since the lower hybrid power generally drives current closer to the plasma edge than does the electron cyclotron power. ITER simulation results obtained using the MMM7.1 transport model are compared with those using the GLF23 model. It is found that, in simulations of target steady state scenarios, momentum transport and flow-shear suppression features of the new MMM7.1 model can lead to predictions of internal transport barriers in temperature and rotation frequency.

  5. International Variation in Ageing and Economic Dependency: A Cohort Perspective

    Directory of Open Access Journals (Sweden)

    Elke Loichinger

    2016-08-01

    Full Text Available Within this analysis of demographic and economic dependency ratios for 45 countries around the world, we reiterate the importance of age- and gender-specific employment levels as well as their determinants when discussing the economic challenges associated with population ageing. Building upon existing research on economic dependency, we portray and discuss cohort variation in employment and its possible effect on the challenges of population ageing, focusing on the implications of high youth unemployment, the role of changes in female employment and the evolution of retirement patterns across cohorts. The insights from our analysis reaffirm findings elsewhere that younger populations may not be as well off in the light of demographic change as an analysis of their demographic structure alone would suggest and stress the importance of considering the cohort dimension of employment in this discussion.

  6. Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hogerton, John

    1964-01-01

    This pamphlet describes how reactors work; discusses reactor design; describes research, teaching, and materials testing reactors; production reactors; reactors for electric power generation; reactors for supply heat; reactors for propulsion; reactors for space; reactor safety; and reactors of tomorrow. The appendix discusses characteristics of U.S. civilian power reactor concepts and lists some of the U.S. reactor power projects, with location, type, capacity, owner, and startup date.

  7. Modernization of Cross Section Library for VVER-1000 Type Reactors Internals and Pressure Vessel Dosimetry

    Directory of Open Access Journals (Sweden)

    Voloschenko Andrey

    2016-01-01

    Full Text Available The broad-group library BGL1000_B7 for neutron and gamma transport calculations in VVER-1000 internals, RPV and shielding was carried out on a base of fine-group library v7-200n47g from SCALE-6 system. The comparison of the library BGL1000_B7 with the library v7-200n47g and the library BGL1000 (the latter is using for VVER-1000 calculations is demonstrated on several calculation and experimental tests.

  8. Modernization of Cross Section Library for VVER-1000 Type Reactors Internals and Pressure Vessel Dosimetry

    Science.gov (United States)

    Voloschenko, Andrey; Zaritskiy, Sergey; Egorov, Aleksander; Boyarinov, Viktor

    2016-02-01

    The broad-group library BGL1000_B7 for neutron and gamma transport calculations in VVER-1000 internals, RPV and shielding was carried out on a base of fine-group library v7-200n47g from SCALE-6 system. The comparison of the library BGL1000_B7 with the library v7-200n47g and the library BGL1000 (the latter is using for VVER-1000 calculations) is demonstrated on several calculation and experimental tests.

  9. Reinventing US Internal Migration Studies in the Age of International Migration.

    Science.gov (United States)

    Ellis, Mark

    2012-03-01

    I argue that researchers have sidelined attention to issues raised by US internal migration as they shifted focus to the questions posed by the post-1960s rise in US immigration. In this paper, I offer some reasons about why immigration has garnered more attention and why there needs to be greater consideration of US internal migration and its significant and myriad social, economic, political, and cultural impacts. I offer three ideas for motivating more research into US internal geographic mobility that would foreground its empirical and conceptual connections to international migration. First, there should be more work on linked migration systems investigating the connections between internal and international flows. Second, the questions asked about immigrant social, cultural, and economic impacts and adaptations in host societies should also be asked about internal migrants. Third, and more generally, migration researchers should jettison the assumption that the national scale is the pre-eminent delimiter of migration types and processes. Some groups can move easily across borders; others are constrained in their moves within countries. These subnational scales and constraints will become more visible if migration research decentres the national from its theory and empirics.

  10. Age-dependent recall bias for material of internal versus external origin.

    Science.gov (United States)

    Barrett, Anna M; Crucian, Gregory P; Wingard, Ellen M; Graybeal, Laura A; Heilman, Kenneth M

    2003-09-01

    To learn whether young and aged subjects exhibit different recall biases for internally derived (Internal) versus externally supplied (External) material. Internally derived knowledge, prized by educators and therapists, can bring about dramatic behavioral change. Such information, seldom assessed on formal memory testing, may be preferentially recalled compared with external-origin material. Under some circumstances, however, subjects may demonstrate a recall advantage for externally supplied over internally generated material. We compared Internal and External word recall in young and aged subjects with and without explicit intent to remember. Although overall the young and aged subjects recalled the same number of words, we did find a word-origin recall bias. This recall bias differed by age group (P = 0.005). When not instructed to remember words, the young subjects tended to remember more External words, while aged subjects remembered more Internal words. The differences in the brain mechanisms mediating Internal versus External recall bias are unknown. However, aging may modify an Internal-External memory bias.

  11. Characterisation of interfacial segregation to Cu-enriched precipitates in two thermally aged reactor pressure vessel steel welds

    Energy Technology Data Exchange (ETDEWEB)

    Styman, P.D., E-mail: paul.styman@materials.ox.ac.uk [Department of Materials, University of Oxford, Parks Road, Oxford OX1 3PH (United Kingdom); National Nuclear Laboratory, B168, Harwell Business Centre, Didcot, Oxon OX11 0QT (United Kingdom); Hyde, J.M. [Department of Materials, University of Oxford, Parks Road, Oxford OX1 3PH (United Kingdom); National Nuclear Laboratory, B168, Harwell Business Centre, Didcot, Oxon OX11 0QT (United Kingdom); Wilford, K.; Parfitt, D.; Riddle, N. [Rolls-Royce, PO Box 2000, Raynesway, Derby DE21 7XX (United Kingdom); Smith, G.D.W. [Department of Materials, University of Oxford, Parks Road, Oxford OX1 3PH (United Kingdom)

    2015-12-15

    To understand the contribution of long term thermal ageing to Reactor Pressure Vessel (RPV) embrittlement two high Cu steel welds with different Ni contents were thermally aged for times up to 100,000 h at 330 °C and 365 °C. Microstructural characterisation using Atom Probe Tomography was performed. Thermal ageing produced a high number density of nano-scale Cu-enriched precipitates. The precipitate–matrix interfaces were enriched in Ni, Mn and Si. The characterisation of these interfaces using a double cluster search approach is the subject of this work. The interface region around thermally-induced precipitates was found to be wider in steels with higher bulk Ni contents and where precipitates had larger core radii. The effect of ageing temperature on interface width was small when comparing precipitates of equal core radius. The narrower interface width in the lower Ni steels is reflected in the composition of the interface, which has a lower Ni content than in the higher Ni material. The reduction in interfacial energy due to the segregation of Ni, Mn and Si has been calculated and shows enhanced reductions in interfacial energy with increasing precipitate size, but no obvious effect of temperature. - Highlights: • Characterisation of interfacial segregation of Ni, Mn and Si to Cu-enriched clusters. • Analysis method gives information on interface composition and widths of large numbers of clusters. • Reduction in interface energy due to segregation of Ni, Mn and Si is calculated.

  12. Probabilistic fracture mechanics analysis of thermally aged nuclear piping in a pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Li, Shuxiao; Zhang, Hailong; Li, Shilei; Wang, Yanli [State Key Laboratory for Advanced Metals and Materials, University of Science and Technology Beijing, Beijing 100083 (China); Xue, Fei [Suzhou Nuclear Power Research Institute, Suzhou 215004 (China); Wang, Xitao, E-mail: xtwang@ustb.edu.cn [State Key Laboratory for Advanced Metals and Materials, University of Science and Technology Beijing, Beijing 100083 (China)

    2013-12-15

    Highlights: • Thermal aging embrittlement was considered in the PFM analysis of nuclear pipe. • Predicting program for pipe failure probability was developed based on thermal aging. • Cumulative failure probability is significantly affected by fracture toughness. • Cumulative failure probability is slightly affected by fatigue crack growth rate. • Tensile strength increase due to thermal aging slightly reduces pipe failure risk. - Abstract: A predicting program for pipe break probability based on thermal aging embrittlement was developed. In order for life prediction, evolutions of fracture toughness and tensile strength were estimated for a Z3CN20-09M piping steel using the Argonne National Laboratory (ANL) procedure. To understand the influence of thermal aging on failure probability, different evolutions of fracture toughness, tensile strength and fatigue crack growth rate were employed in the prediction of cumulative failure probability. The results show that the cumulative failure probability for 40-year thermal aging increases by almost four times compared to without consideration of fracture toughness degradation. The cumulative failure probability is slightly affected by fatigue crack growth rate. The increase of tensile strength due to thermal aging reduces the risk of pipe failure. This work demonstrates that the degradation of fracture toughness due to thermal aging should be fully considered in the probabilistic fracture mechanics analysis of nuclear pressure pipes.

  13. Modelling 3D crack propagation in ageing graphite bricks of Advanced Gas-cooled Reactor power plant

    Directory of Open Access Journals (Sweden)

    Thi-Tuyet-Giang Vo

    2015-10-01

    Full Text Available In this paper, crack propagation in Advanced Gas-cooled Reactor (AGR graphite bricks with ageing properties is studied using the eXtended Finite Element Method (X-FEM. A parametric study for crack propagation, including the influence of different initial crack shapes and propagation criteria, is conducted. The results obtained in the benchmark study show that the crack paths from X-FEM are similar to the experimental ones. The accuracy of the strain energy release rate computation in a heterogeneous material is also evaluated using a finite difference approach. Planar and non-planar 3D crack growth simulations are presented to demonstrate the robustness and the versatility of the method utilized. Finally, this work contributes to the better understanding of crack propagation behaviour in AGR graphite bricks and so contributes to the extension of the AGR plants’ lifetimes in the UK by reducing uncertainties.

  14. High power 1 MeV neutral beam system and its application plan for the international tokamak experimental reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hemsworth, R.S. [ITER Joint Central Team, Naka, Ibaraki (Japan)

    1997-03-01

    This paper describes the Neutral Beam Injection system which is presently being designed for the International Tokamak Experimental Reactor, ITER, in Europe Japan and Russia, with co-ordination by the Joint Central Team of ITER at Naka, Japan. The proposed system consists of three negative ion based neutral injectors, delivering a total of 50 MW of 1 MeV D{sup 0} to the ITER plasma for a pulse length of >1000 s. Each injectors uses a single caesiated volume arc discharge negative ion source, and a multi-grid, multi-aperture accelerator, to produce about 40 A of 1 MeV D{sup -}. This will be neutralized by collisions with D{sub 2} in a sub-divided gas neutralizer, which has a conversion efficiency of about 60%. The charged fraction of the beam emerging from the neutralizer is dumped in an electrostatic residual ion dump. A water cooled calorimeter can be moved into the beam path to intercept the neutral beam, allowing commissioning of the injector independent of ITER. ITER is scheduled to produce its first plasma at the beginning of 2008, and the planning of the R and D, construction and installation foresees the neutral injection system being available from the start of ITER operations. (author)

  15. International women's movement comes of age at FWCW.

    Science.gov (United States)

    O'haire, H

    1996-02-01

    This article focuses on what the women's conference achieved for women. The Fourth World Conference on Women (FWCW) recognized an international women's movement. Governments agreed that women were entitled to full human rights, reproductive rights, and equality in all aspects of life. Women were to become full and equal partners in the development process. Women are likely to demand action based on the paper declarations. The success of the conference is attributed to a powerful network of women's groups. A striking feature was the heading of most national delegations by women. It was recognized that women were full but neglected partners in families and societies. Governments acknowledged that women suffered discrimination within the family and society. Wording was rejected that would have given women the guaranteed right to determine the size and spacing of their families and the information and services to do so. The activity among women's groups in working together to restore this view strengthened the bonds between women's groups and created worldwide awareness and support for women's organizations. Networks that were established at Rio de Janeiro made their presence and position on reproductive health felt at the preparatory meetings to the UN Population Conference in Cairo. It was argued that a target-driven approach had the effect of treating women as reproduction machines. Women's groups also expressed strong positions on how women were to be treated in the formation of population policy and programs. 1) It was insisted that governments and population groups must stop dictating fertility regimes to women. 2) The concept of family planning must include reproductive health. 3) Women should have the freedom to exercise choice in planning their families. 4) Women should also have equal access to education and employment. Women came thus to Beijing with a clear vision of what they wanted. At Beijing domestic violence was for the first time condemned. Beijing

  16. Revisiting the Integrated Pressurized Thermal Shock Studies of an Aging Pressurized Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bryson, J.W.; Dickson, T.L.; Malik, S.N.M.; Simonen, F.A.

    1999-08-01

    The Integrated Pressurized Thermal Shock (IPTS) studies were a series of studies performed in the early-mid 1980s as part of an NRC-organized comprehensive research project to confirm the technical bases for the pressurized thermal shock (PTS) rule, and to aid in the development of guidance for licensee plant-specific analyses. The research project consisted of PTS pilot analyses for three PWRs: Oconee Unit 1, designed by Babcock and Wilcox; Calvert Cliffs Unit 1, designed by Combustion Engineering; and H.B. Robinson Unit 2, designed by Westinghouse. The primary objectives of the IPTS studies were (1) to provide for each of the three plants an estimate of the probability of a crack propagating through the wall of a reactor pressure vessel (RPV) due to PTS; (2) to determine the dominant overcooling sequences, plant features, and operator actions and the uncertainty in the plant risk due to PTS; and (3) to evaluate the effectiveness of potential corrective actions. The NRC is currently evaluating the possibility of revising current PTS regulatory guidance. Technical bases must be developed to support any revisions. In the years since the results of IPTS studies were published, the fracture mechanics model, the embrittlement database, embrittlement correlation, inputs for flaw distributions, and the probabilistic fracture mechanics (PFM) computer code have been refined. An ongoing effort is underway to determine the impact of these fracture-technology refinements on the conditional probabilities of vessel failure calculated in the IPTS Studies. This paper discusses the results of these analyses performed for one of these plants.

  17. Prioritization of reactor control components susceptible to fire damage as a consequence of aging

    Energy Technology Data Exchange (ETDEWEB)

    Lowry, W.; Vigil, R. [Science and Engineering Associates, Inc., Albuquerque, NM (United States); Nowlen, S. [Sandia National Labs., Albuquerque, NM (United States)

    1994-01-01

    The Fire Vulnerability of Aged Electrical Components Test Program is to identify and assess issues of plant aging that could lead to an increase in nuclear power plant risk because of fires. Historical component data and prior analyses are used to prioritize a list of components with respect to aging and fire vulnerability and the consequences of their failure on plant safety systems. The component list emphasizes safety system control components, but excludes cables, large equipment, and devices encompassed in the Equipment Qualification (EQ) program. The test program selected components identified in a utility survey and developed test and fire conditions necessary to maximize the effectiveness of the test program. Fire damage considerations were limited to purely thermal effects.

  18. Progress in preparing scenarios for operation of the International Thermonuclear Experimental Reactor

    Science.gov (United States)

    Sips, A. C. C.; Giruzzi, G.; Ide, S.; Kessel, C.; Luce, T. C.; Snipes, J. A.; Stober, J. K.

    2015-02-01

    The development of operating scenarios is one of the key issues in the research for ITER which aims to achieve a fusion gain (Q) of ˜10, while producing 500 MW of fusion power for ≥300 s. The ITER Research plan proposes a success oriented schedule starting in hydrogen and helium, to be followed by a nuclear operation phase with a rapid development towards Q ˜ 10 in deuterium/tritium. The Integrated Operation Scenarios Topical Group of the International Tokamak Physics Activity initiates joint activities among worldwide institutions and experiments to prepare ITER operation. Plasma formation studies report robust plasma breakdown in devices with metal walls over a wide range of conditions, while other experiments use an inclined EC launch angle at plasma formation to mimic the conditions in ITER. Simulations of the plasma burn-through predict that at least 4 MW of Electron Cyclotron heating (EC) assist would be required in ITER. For H-modes at q95 ˜ 3, many experiments have demonstrated operation with scaled parameters for the ITER baseline scenario at ne/nGW ˜ 0.85. Most experiments, however, obtain stable discharges at H98(y,2) ˜ 1.0 only for βN = 2.0-2.2. For the rampup in ITER, early X-point formation is recommended, allowing auxiliary heating to reduce the flux consumption. A range of plasma inductance (li(3)) can be obtained from 0.65 to 1.0, with the lowest values obtained in H-mode operation. For the rampdown, the plasma should stay diverted maintaining H-mode together with a reduction of the elongation from 1.85 to 1.4. Simulations show that the proposed rampup and rampdown schemes developed since 2007 are compatible with the present ITER design for the poloidal field coils. At 13-15 MA and densities down to ne/nGW ˜ 0.5, long pulse operation (>1000 s) in ITER is possible at Q ˜ 5, useful to provide neutron fluence for Test Blanket Module assessments. ITER scenario preparation in hydrogen and helium requires high input power (>50 MW). H

  19. Short-time plasma surface modification of HDPE powder in a Plasma Downer Reactor - process, wettability improvement and ageing effects

    Energy Technology Data Exchange (ETDEWEB)

    Arpagaus, C. [ETH Swiss Federal Institute of Technology Zurich, Institute of Process Engineering, Department of Mechanical and Process Engineering, ETH Zentrum, Sonneggstrasse 3, CH-8092 Zurich (Switzerland); Rossi, A. [ETH Swiss Federal Institute of Technology Zurich, Laboratory for Surface Science and Technology, Department of Materials, ETH Hoenggerberg, Wolfgang-Pauli-Strasse 10, CH-8093 Zurich (Switzerland); Universita degli Studi di Cagliari, Dipartimento di Chimica Inorganica ed Analitica, UdR INSTM I-09100 Cagliari (Italy); Rudolf von Rohr, Ph. [ETH Swiss Federal Institute of Technology Zurich, Institute of Process Engineering, Department of Mechanical and Process Engineering, ETH Zentrum, Sonneggstrasse 3, CH-8092 Zurich (Switzerland)]. E-mail: vonrohr@ipe.mavt.ethz.ch

    2005-12-15

    The effectiveness of improving the wettability of HDPE powders within less than 0.1 s by plasma surface modification in a Plasma Downer Reactor is investigated. A correlation is revealed between the XPS results (O/C-ratio) and the wettability (contact angle, polar surface tension by capillary rise method). The O{sub 2}-content in the plasma feed gas has been adjusted for best wettability properties. XPS results indicate the formation of C=O and COOH functional groups on the powder surface. The O/C-ratio increased from 0.0 (no oxygen on the non-treated powder) up to 0.15 for the plasma treated HDPE powder surface. With pure O{sub 2}-plasma treatment, a water contact angle reduction from >90{sup o} (no water penetration into the untreated PE powder) down to 65{sup o} was achieved. The total surface free energy increased from 31.2 to 45 mN/m. Ageing of treated powders occurs and proceeds mostly within the first 7 days of storage. Contact angle measurements and O1s/O2s intensity ratio data support that ageing is mainly a diffusion-controlled process. Nevertheless, XPS results show the presence of oxygen functional groups even after 40 days, which explains why the powder is still dispersible in water without any addition of surfactants.

  20. Microstructural changes of a thermally aged stainless steel submerged arc weld overlay cladding of nuclear reactor pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Takeuchi, T., E-mail: takeuchi.tomoaki@jaea.go.jp [Japan Atomic Energy Agency, Tokai, Ibaraki 319-1195 (Japan); Kameda, J. [National Institute for Materials Science, Sengen, Tsukuba 305-0047 (Japan); Nagai, Y.; Toyama, T.; Matsukawa, Y. [Oarai Center, Institute for Materials Research, Tohoku University, Oarai, Ibaraki 311-1313 (Japan); Nishiyama, Y.; Onizawa, K. [Japan Atomic Energy Agency, Tokai, Ibaraki 319-1195 (Japan)

    2012-06-15

    The effect of thermal aging on microstructural changes in stainless steel submerged arc weld-overlay cladding of reactor pressure vessels was investigated using atom probe tomography (APT). In as-received materials subjected to post-welding heat treatments (PWHTs), with a subsequent furnace cooling, a slight fluctuation of the Cr concentration was observed due to spinodal decomposition in the {delta}-ferrite phase but not in the austenitic phase. Thermal aging at 400 Degree-Sign C for 10,000 h caused not only an increase in the amplitude of spinodal decomposition but also the precipitation of G phases with composition ratios of Ni:Si:Mn = 16:7:6 in the {delta}-ferrite phase. The degree of the spinodal decomposition in the submerged arc weld sample was similar to that in the electroslag weld one reported previously. We also observed a carbide on the {gamma}-austenite and {delta}-ferrite interface. There were no Cr depleted zones around the carbide.

  1. Experimental reactor regulation: the nuclear safety authority's approach; Le controle des reacteurs experimentaux: la demarche de l'Autorite de surete nucleaire

    Energy Technology Data Exchange (ETDEWEB)

    Rieu, J.; Conte, D.; Chevalier, A. [Autorite de Surete Nucleaire, 75 - Paris (France)

    2007-07-15

    French research reactors can be classified into 6 categories: 1) critical scale models (Eole, Minerve and Masurca) whose purpose is the study of the neutron production through the fission reaction; 2) reactors that produce neutron beams (Orphee, and the high flux reactor in Grenoble); 3) reactors devoted to safety studies (Cabri, Phebus) whose purpose is to reproduce accidental configurations of power reactors in reduced scale; 4) experimental reactors (Osiris, Phenix) whose purpose is the carrying-out of irradiation experiments concerning nuclear fuels or structure materials; 5) teaching reactors (Ulysse, Isis); and 6) reactors involved in defense programs (Caliban, Prospero, Apareillage-B). We have to note that 3 research reactors are currently being dismantled: Strasbourg University's reactor, Siloe and Siloette. Research reactors in France are of different types and present different hazards. Even if methods of control become more and more similar to those of power reactors, the French Nuclear Safety Authority (ASN) works to allow the necessary flexibility in the ever changing research reactor field while ensuring a high level of safety. Adopting the internal authorizations for operations of minor safety significance, under certain conditions, is one example of this approach. Another challenge in the coming years for ASN is to monitor the ageing of the French research reactors. This includes periodic safety reviews for each facility every ten years. But ASN has also to regulate the new research reactor projects such as Jules Horowitz Reactor, International Thermonuclear Experimental Reactor, which are about to be built.

  2. Growing Old in a New Age: National and International Evaluation of a Gerontology Telecourse.

    Science.gov (United States)

    Dubanoski, Joan Pabst; Goodman, Rebecca J.; Braun, Kathryn L.; Roberts, Ellen; Lenzer, Anthony M.

    1999-01-01

    "Growing Old in a New Age," a Public Broadcasting Service telecourse, was evaluated by 40 faculty, 29 state aging administrators, 2 international public broadcasting users, and a training specialist. The course was generally considered effective, but improvements were recommended in content, integration of video and text, and updated…

  3. A Longitudinal Study of Language and Speech in Children Who Were Internationally Adopted at Different Ages

    Science.gov (United States)

    Glennen, Sharon

    2014-01-01

    Purpose: The author followed 56 internationally adopted children during the first 3 years after adoption to determine how and when they reached age-expected language proficiency in Standard American English. The influence of age of adoption was measured, along with the relationship between early and later language and speech outcomes. Method:…

  4. Aging Assessment of an Oak Ridge National Laboratory High Flux Isotope Reactor (HFIR) Service Cable.

    Energy Technology Data Exchange (ETDEWEB)

    Bernstein, Robert; Celina, Mathias Christopher; Redline, Erica Marie; White, II, Gregory Von

    2014-07-01

    Nuclear energy is one industry where aging of safety-related materials and components is of great concern. Many U.S. nuclear power plants are approaching, or have already exceeded, 40 years of age. Analysis comparing the cost of new plant construction versus long-term operation under extended plant licensing through 60 years strongly favors the latter option. To ensure the safe, reliable, and cost-effective long-term operation of nuclear power plants, many systems, structures, and components must be evaluated. Furthermore, as new analytical techniques and testing approaches are developed, it is imperative that we also validate, and if necessary, improve upon the previously employed Institute of Electrical and Electronic Engineers (IEEE) qualification standards originally written in 1974. Fortunately, this daunting task has global support, particularly in light of the new social and political climate surrounding nuclear energy in a post-Fukushima era.

  5. Real-time measurements of secondary organic aerosol formation and aging from ambient air in an oxidation flow reactor in the Los Angeles area

    Science.gov (United States)

    Ortega, Amber M.; Hayes, Patrick L.; Peng, Zhe; Palm, Brett B.; Hu, Weiwei; Day, Douglas A.; Li, Rui; Cubison, Michael J.; Brune, William H.; Graus, Martin; Warneke, Carsten; Gilman, Jessica B.; Kuster, William C.; de Gouw, Joost; Gutiérrez-Montes, Cándido; Jimenez, Jose L.

    2016-06-01

    Field studies in polluted areas over the last decade have observed large formation of secondary organic aerosol (SOA) that is often poorly captured by models. The study of SOA formation using ambient data is often confounded by the effects of advection, vertical mixing, emissions, and variable degrees of photochemical aging. An oxidation flow reactor (OFR) was deployed to study SOA formation in real-time during the California Research at the Nexus of Air Quality and Climate Change (CalNex) campaign in Pasadena, CA, in 2010. A high-resolution aerosol mass spectrometer (AMS) and a scanning mobility particle sizer (SMPS) alternated sampling ambient and reactor-aged air. The reactor produced OH concentrations up to 4 orders of magnitude higher than in ambient air. OH radical concentration was continuously stepped, achieving equivalent atmospheric aging of 0.8 days-6.4 weeks in 3 min of processing every 2 h. Enhancement of organic aerosol (OA) from aging showed a maximum net SOA production between 0.8-6 days of aging with net OA mass loss beyond 2 weeks. Reactor SOA mass peaked at night, in the absence of ambient photochemistry and correlated with trimethylbenzene concentrations. Reactor SOA formation was inversely correlated with ambient SOA and Ox, which along with the short-lived volatile organic compound correlation, indicates the importance of very reactive (τOH ˜ 0.3 day) SOA precursors (most likely semivolatile and intermediate volatility species, S/IVOCs) in the Greater Los Angeles Area. Evolution of the elemental composition in the reactor was similar to trends observed in the atmosphere (O : C vs. H : C slope ˜ -0.65). Oxidation state of carbon (OSc) in reactor SOA increased steeply with age and remained elevated (OSC ˜ 2) at the highest photochemical ages probed. The ratio of OA in the reactor output to excess CO (ΔCO, ambient CO above regional background) vs. photochemical age is similar to previous studies at low to moderate ages and also extends to

  6. Sensitivity analysis of the spectra of the core neutronic source in the calculation of radiation damage in internal of PWR reactor vessel. Internal; Analisis de sensibilidad a los espectros de la fuente neutronica del nucleo en el calculo del dano por irradiacion en los internos de la vasija de un reactor PWR

    Energy Technology Data Exchange (ETDEWEB)

    Cadenas Mendicoa, A. M.; Benito Hernandez, M.; Barrerira Pereira, P.

    2012-07-01

    This study is to analyze the sensitivity to the expected differences in the energy spectra characterizing the neutron source that radiates the vessel internals of a commercial PWR reactor, in order to quantify their influence in the quantities that determine the damage in materials metal.

  7. Modeling of Late Blooming Phases and Precipitation Kinetics in Aging Reactor Pressure Vessel (RPV) Steels

    Energy Technology Data Exchange (ETDEWEB)

    Yongfeng Zhang; Pritam Chakraborty; S. Bulent Biner

    2013-09-01

    The principle work at the atomic scale is to develop a predictive quantitative model for the microstructure evolution of RPV steels under thermal aging and neutron radiation. We have developed an AKMC method for the precipitation kinetics in bcc-Fe, with Cu, Ni, Mn and Si being the alloying elements. In addition, we used MD simulations to provide input parameters (if not available in literature). MMC simulations were also carried out to explore the possible segregation/precipitation morphologies at the lattice defects. First we briefly describe each of the simulation algorithms, then will present our results.

  8. Dr Robert Aymar, Director of the International Thermonuclear Experimental Reactor (ITER), was nominated to succeed Professor Luciano Maiani as CERN's Director General, to take office on 1 January 2004.

    CERN Multimedia

    2002-01-01

    Dr Robert Aymar, Director of the International Thermonuclear Experimental Reactor (ITER), was nominated to succeed Professor Luciano Maiani as CERN's Director General, to take office on 1 January 2004.

  9. U.S. Perspectives: International Action on Aging. A Background Paper Prepared by the American Association for International Aging for the Select Committee on Aging. House of Representatives, Ninety-Eighth Congress, Second Session.

    Science.gov (United States)

    Congress of the U.S., Washington, DC. House Select Committee on Aging.

    In response to challenges and guidelines set forth in the 1982 International Plan of Action on Aging (IPAA) by the World Assembly on Aging, this background paper summarizes (1) immediate and long-range reasons for the World Assembly; (2) content and significance of the IPAA and the factual base on which action plan decisions were made; (3)…

  10. Development and evaluation of thermoelectric power measurements as a non destructive technique to evaluate ageing of reactor pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Acosta, B.; Debarberis, L. [European Commission, JRC Institute for Advanced Materials, Petten (Netherlands); Perlado, J.M. [Universidad Politecnica de Madrid, Instituto de Fusion Nuclear, DENIM, E.T.S.I.I., Madris (Spain)

    2001-07-01

    The STEAM (Seebeck and Thomson effects on aged materials) technique developed at the JRC-IAM (joint research centre - institute for advanced materials), is a new non-destructive method able to detect in a simple way degradation of materials, in particular to be applied on those steels that form the reactor pressure vessel of nuclear plants. The STEAM method is based on the measurement of the thermoelectric voltage generated by the Seebeck and Thomson effects taking place in the material under test. In order to evaluate the performance of the STEAM technique on irradiated material a set of 32 model alloys was selected. Measurements with the STEAM technique have been performed on the model alloys in both conditions, fresh and irradiated, with the aim of correlating the irradiation induced embrittlement and the change on the Seebeck coefficient due to irradiation. The results show that there is a relationship between transition temperature shifts and Seebeck coefficient value change between irradiated and fresh materials. In order to understand the response of the Seebeck coefficient to neutron irradiation damage a model based on multivariable correlation analysis is proposed. (A.C.)

  11. Biodegradation of 2,4,6-trichlorophenol in a packed-bed biofilm reactor equipped with an internal net draft tube riser for aeration and liquid circulation

    Energy Technology Data Exchange (ETDEWEB)

    Gomez-De Jesus, A.; Romano-Baez, F.J.; Leyva-Amezcua, L.; Juarez-Ramirez, C.; Ruiz-Ordaz, N. [Departamento de Ingenieria Bioquimica, Escuela Nacional de Ciencias Biologicas, IPN. Prol. Carpio y Plan de Ayala, Colonia Santo Tomas, s/n. CP 11340, Mexico, D.F. (Mexico); Galindez-Mayer, J. [Departamento de Ingenieria Bioquimica, Escuela Nacional de Ciencias Biologicas, IPN. Prol. Carpio y Plan de Ayala, Colonia Santo Tomas, s/n. CP 11340, Mexico, D.F. (Mexico)], E-mail: cmayer@encb.ipn.mx

    2009-01-30

    For the aerobic biodegradation of the fungicide and defoliant 2,4,6-trichlorophenol (2,4,6-TCP), a bench-scale packed-bed bioreactor equipped with a net draft tube riser for liquid circulation and oxygenation (PB-ALR) was constructed. To obtain a high packed-bed volume relative to the whole bioreactor volume, a high A{sub D}/A{sub R} ratio was used. Reactor's downcomer was packed with a porous support of volcanic stone fragments. PB-ALR hydrodynamics and oxygen mass transfer behavior was evaluated and compared to the observed behavior of the unpacked reactor operating as an internal airlift reactor (ALR). Overall gas holdup values {epsilon}{sub G}, and zonal oxygen mass transfer coefficients determined at various airflow rates in the PB-ALR, were higher than those obtained with the ALR. When comparing mixing time values obtained in both cases, a slight increment in mixing time was observed when reactor was operated as a PB-ALR. By using a mixed microbial community, the biofilm reactor was used to evaluate the aerobic biodegradation of 2,4,6-TCP. Three bacterial strains identified as Burkholderia sp., Burkholderia kururiensis and Stenotrophomonas sp. constituted the microbial consortium able to cometabolically degrade the 2,4,6-TCP, using phenol as primary substrate. This consortium removed 100% of phenol and near 99% of 2,4,6-TCP. Mineralization and dehalogenation of 2,4,6-TCP was evidenced by high COD removal efficiencies ({approx}95%), and by the stoichiometric release of chloride ions from the halogenated compound ({approx}80%). Finally, it was observed that the microbial consortium was also capable to metabolize 2,4,6-TCP without phenol as primary substrate, with high removal efficiencies (near 100% for 2,4,6-TCP, 92% for COD and 88% for chloride ions)

  12. 核电用堆内构件加工线工艺与设备方案%The process and design for the internal parts in reactor

    Institute of Scientific and Technical Information of China (English)

    严小林; 刘桂芝; 辛向东

    2012-01-01

    The demonstration power station with the high temperature gas cooled reactor (HTR-PM) is the latest generation nuclear power station. It has gained great attention owing to its high safety and reliability. The internal parts in reactor include the adiabatic layer made of the carbon, the material containing boron and the reflecting layer made of graphite. Both can not be changed within the life cycle of the reactor. Based on the analysis on the feature of structure of the internal parts in the certain demonstration power station with the high temperature gas cooled reactor, the technology requirement for the components and assembly requirements, the corresponding auto-machining line is designed.%高温气冷反应堆示范电站(HTR-PM)是最新一代核电站,因其安全性、高可靠性及经济性受到广泛关注.其反应堆堆芯一般用炭及含硼炭材料做绝热层,用石墨质材料做反射层,并且在反应堆使用寿命周期内不可拆换.本文通过对国内某高温气冷反应堆示范电站堆芯堆内构件的结构特点、零件的技术要求、装配要求等进行分析与综合,提出堆内构件的加工工艺方案,由此而设计出适合堆内构件加工的自动线.

  13. Contribution of the cornea and internal surfaces to the change of ocular aberrations with age

    Science.gov (United States)

    Artal, Pablo; Berrio, Esther; Guirao, Antonio; Piers, Patricia

    2002-01-01

    We studied the age dependence of the relative contributions of the aberrations of the cornea and the internal ocular surfaces to the total aberrations of the eye. We measured the wave-front aberration of the eye with a Hartmann-Shack sensor and the aberrations of the anterior corneal surface from the elevation data provided by a corneal topography system. The aberrations of the internal surfaces were obtained by direct subtraction of the ocular and corneal wave-front data. Measurements were obtained for normal healthy subjects with ages ranging from 20 to 70 years. The magnitude of the RMS wave-front aberration (excluding defocus and astigmatism) of the eye increases more than threefold within the age range considered. However, the aberrations of the anterior corneal surface increase only slightly with age. In most of the younger subjects, total ocular aberrations are lower than corneal aberrations, while in the older subjects the reverse condition occurs. Astigmatism, coma, and spherical aberration of the cornea are larger than in the complete eye in younger subjects, whereas the contrary is true for the older subjects. The internal ocular surfaces compensate, at least in part, for the aberrations associated with the cornea in most younger subjects, but this compensation is not present in the older subjects. These results suggest that the degradation of the ocular optics with age can be explained largely by the loss of the balance between the aberrations of the corneal and the internal surfaces.

  14. Improved measures for the cross-national comparison of age profiles of internal migration.

    Science.gov (United States)

    Bernard, Aude; Bell, Martin; Charles-Edwards, Elin

    2014-01-01

    We develop and demonstrate the application of a concise set of measures intended to encapsulate key features of the age profile of internal migration and highlight the significant differences that exist between nations in these profiles. Model schedules have been the most common method of comparing internal migration patterns but issues related to the estimation and interpretation of their parameters hinder their use for cross-national comparison. We demonstrate that the interpretation of exponential coefficients as rates of ascent and descent does not best reflect the slopes of migration age profiles, and we propose more consistent measures based on the rate of change in migration intensity. We demonstrate, through correlation and factor analysis, that most of the inter-country variance in migration age profiles is captured by the age at and intensity of peak migration. The application of these two indicators to 25 countries reveals significant differences between regions.

  15. Microstructural characterization and model of hardening for the irradiated austenitic stainless steels of the internals of pressurized water reactors; Caracterisation microstructurale et modelisation du durcissement des aciers austenitiques irradies des structures internes des reacteurs a eau pressurisee

    Energy Technology Data Exchange (ETDEWEB)

    Pokor, C

    2003-07-01

    The core internals of Pressurized Water Reactors (PWR) are composed of SA 304 stainless steel plates and CW 316 stainless steel bolts. These internals undergo a neutron flux at a temperature between 280 deg C and 380 deg C which modifies their mechanical properties. These modifications are due to the changes in the microstructure of these materials under irradiation which depend on flux, dose and irradiation temperature. We have studied, by Transmission Electron Microscopy, the microstructure of stainless steels SA 304, CW 316 and CW 316Ti irradiated in a mixed flux reactor (OSIRIS at 330 deg C between 0,8 dpa et 3,4 dpa) and in a fast breeder reactor at 330 deg C (BOR-60) up to doses of 40 dpa. Moreover, samples have been irradiated at 375 deg C in a fast breeder reactor (EBR-II) up to doses of 10 dpa. The microstructure of the irradiated stainless steels consists in faulted Frank dislocation loops in the [111] planes of austenitic, with a Burgers vector of [111]. It is possible to find some voids in the solution annealed samples irradiated at 375 deg C. The evolution of the dislocations loops and voids has been simulated with a 'cluster dynamic' model. The fit of the model parameters has allowed us to have a quantitative description of our experimental results. This description of the microstructure after irradiation was coupled together with a hardening model by Frank loops that has permitted us to make a quantitative description of the hardening of SA 304, CW 316 and CW 316Ti stainless steels after irradiation at a certain dose, flux and temperature. The irradiation doses studied grow up to 90 dpa, dose of the end of life of PWR internals. (author)

  16. Study of impact of the AP1000{sup Registered-Sign} reactor vessel upper internals design on fuel performance

    Energy Technology Data Exchange (ETDEWEB)

    Xu Yiban; Conner, Michael; Yuan Kun; Dzodzo, Milorad B.; Karoutas, Zeses; Beltz, Steven A.; Ray, Sumit; Bissett, Teresa A. [Westinghouse Electric Company, Cranberry Township, PA 16066 (United States); Chieng, Ching-Chang, E-mail: cchieng@ess.nthu.edu.tw [National Tsing Hua University, Hsinchu 30043, Taiwan (China); Kao, Min-Tsung; Wu, Chung-Yun [National Tsing Hua University, Hsinchu 30043, Taiwan (China)

    2012-11-15

    One aspect of the AP1000{sup Registered-Sign} reactor design is the reduction in the number of major components and simplification in manufacturing. One design change relative to current Westinghouse reactors of similar size is the reduction in the number of reactor vessel outlet nozzles/hot legs leaving the upper plenum from three to two. With regard to fuel performance, this design difference creates a different flow field in the AP1000 reactor vessel upper plenum (the region above the core). The flow exiting core and entering the upper plenum must turn 90 Degree-Sign , flow laterally through the upper plenum around support structures, and exit through one of the two outlet nozzles. While the flow in the top of the core is mostly axial, there is some lateral flow component as the core flow reacts to the flow field and pressure distribution in the upper plenum. The pressure distribution in the upper plenum varies laterally depending upon various factors including the proximity to the outlet nozzles. To determine how the lateral flow in the top of the AP1000 core compares to current Westinghouse reactors, a computational fluid dynamics (CFD) model of the flow in the upper portion of the AP1000 reactor vessel including the top region of the core, the upper plenum, the reactor vessel outlet nozzles, and a portion of the hot legs was created. Due to geometric symmetry, the computational domain was reduced to a quarter (from the top view) that includes Vulgar-Fraction-One-Quarter of the top of the core, Vulgar-Fraction-One-Quarter of the upper plenum, and Vulgar-Fraction-One-Half of an outlet nozzle. Results from this model include predicted velocity fields and pressure distributions throughout the model domain. The flow patterns inside and around guide tubes clearly demonstrate the influence of lateral flow due to the presence of the outlet nozzles. From these results, comparisons of AP1000 flow versus current Westinghouse plants were performed. Field performance

  17. Experimental Investigation on the Internal Resistance of Lithium Iron Phosphate Battery Cells during Calendar Ageing

    DEFF Research Database (Denmark)

    Stroe, Daniel Ioan; Swierczynski, Maciej Jozef; Stan, Ana-Irina

    2013-01-01

    is directly related to its internal resistance. This work aims to investigate the dependency of the internal resistance of lithium-ion batteries on the storage temperature and on the storage time. For this purpose, accelerated ageing calendar lifetime tests were carried out over a period of one year. Based......Lithium-ion batteries are increasingly considered for a wide area of applications because of their superior characteristics in comparisons to other energy storage technologies. However, at present, Lithium-ion batteries are expensive storage devices and consequently their ageing behavior must...... be known in order to estimate their economic viability in different application. The ageing behavior of Lithium-ion batteries is described by the fade of their discharge capacity and by the decrease of their power capability. The capability of a Lithium-ion battery to deliver or to absorb a certain power...

  18. Safety-related Innovative Nuclear Reactor Technology Elements R and D (SINTER) Network and Global HTGR R and D Network (GHTRN). Strategic benefits of international networking

    Energy Technology Data Exchange (ETDEWEB)

    Von Lensa, W. [Institut fuer Sicherheitsforschung und Reaktortechnik ISR, Forschungszentrum Juelich, Juelich (Germany)

    1998-09-01

    The nuclear industries and the nuclear research and development (R and D) programmes world-wide have undergone considerable changes over recent years which have resulted in the formation of international industrial consortiums on the one hand and the need for synergistic collaboration in the R and D area due to the reductions of national R and D activities in the nuclear field on the other hand. International networking starting from precompetitive medium- or long-term oriented R and D could be an efficient mean to overcome the problems nuclear energy is facing today with respect to the lack of public acceptance and economic attractivity in a joint effort. Additional motivation is provided by the fact that there is not only a globalisation of markets but also a `globalisation of problems` to be addressed internationally like reductions of environmental impacts and long-term availability of economic energy supply. The tools for telecommunication and telecollaboration are evolving in parallel and offer better conditions for closer collaboration of different R and D teams at distant locations than ever before. It is obvious that these trends and boundary conditions will drastically influence the structures of collaboration not only in the industries, but for R and D on an international level, too. The chances emerging from the creation of a European Union and from the globalisation trends have to be converted into strategic benefits by active response on these `historic changes`. New initiatives have been undertaken in Europe to push for innovations of nuclear reactor technologies via international R and D Networks under the European R and D Framework Programmes (FWP). Innovative approaches are already addressed with limited funding under the actual 4th FWP and should be extended for complementing the commercial efforts on evolutionary LWR concepts by medium- and long-term oriented innovations and R and D. The MICHELANGELO initiative as well as the EU-funded Concerted

  19. Treatment of mature landfill leachate by internal micro-electrolysis integrated with coagulation: A comparative study on a novel sequencing batch reactor based on zero valent iron

    Energy Technology Data Exchange (ETDEWEB)

    Ying, Diwen [School of Environmental Science and Engineering, Shanghai Jiao Tong University, 800 Dongchuan Road, Shanghai, 200240 (China); Peng, Juan [Department of Civil and Environmental Engineering, Carnegie Mellon University, 5000 Forbes Avenue, Pittsburgh, PA 15213 (United States); Xu, Xinyan; Li, Kan; Wang, Yalin [School of Environmental Science and Engineering, Shanghai Jiao Tong University, 800 Dongchuan Road, Shanghai, 200240 (China); Jia, Jinping, E-mail: jpjia@sjtu.edu.cn [School of Environmental Science and Engineering, Shanghai Jiao Tong University, 800 Dongchuan Road, Shanghai, 200240 (China)

    2012-08-30

    Highlights: Black-Right-Pointing-Pointer Specifically-designed SIME reactor for treatment of mature landfill leachate. Black-Right-Pointing-Pointer Excellent removal efficiencies of COD (86.1%), color (95.3%), and HA (81.8%). Black-Right-Pointing-Pointer Combination effect of IME without aeration and IME with aeration. Black-Right-Pointing-Pointer Optimal pH of 5, Fe/C of 1:1, gas flow rate of 80 L h{sup -1}, and H{sub 2}O{sub 2} of 100 mg L{sup -1}. - Abstract: A comparative study of treating mature landfill leachate with various treatment processes was conducted to investigate whether the method of combined processes of internal micro-electrolysis (IME) without aeration and IME with full aeration in one reactor was an efficient treatment for mature landfill leachate. A specifically designed novel sequencing batch internal micro-electrolysis reactor (SIME) with the latest automation technology was employed in the experiment. Experimental data showed that combined processes obtained a high COD removal efficiency of 73.7 {+-} 1.3%, which was 15.2% and 24.8% higher than that of the IME with and without aeration, respectively. The SIME reactor also exhibited a COD removal efficiency of 86.1 {+-} 3.8% to mature landfill leachate in the continuous operation, which is much higher (p < 0.05) than that of conventional treatments of electrolysis (22.8-47.0%), coagulation-sedimentation (18.5-22.2%), and the Fenton process (19.9-40.2%), respectively. The innovative concept behind this excellent performance is a combination effect of reductive and oxidative processes of the IME, and the integration electro-coagulation. Optimal operating parameters, including the initial pH, Fe/C mass ratio, air flow rate, and addition of H{sub 2}O{sub 2}, were optimized. All results show that the SIME reactor is a promising and efficient technology in treating mature landfill leachate.

  20. Effects of aging in containment spray injection system of PWR reactor containment; Efeitos do envelhecimento no sistema de injecao de borrifo da contencao de reatores a agua pressurizada

    Energy Technology Data Exchange (ETDEWEB)

    Borges, Diogo da S.; Lava, Deise D.; Affonso, Renato R.W.; Guimaraes, Antonio C.F.; Moreira, Maria de L., E-mail: diogosb@outlook.com, E-mail: deise_dy@hotmail.com, E-mail: raoniwa@yahoo.com.br, E-mail: tony@ien.gov.br, E-mail: malu@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2014-07-01

    This paper presents a contribution to the study of the components aging process in commercial plants of Pressurized Water Reactors (PWR). The analysis is done by applying the method of Fault trees, Monte Carlo Method and Fussell-Vesely Importance Measurement. The study on the aging of nuclear plants, is related to economic factors involved directly with the extent of their operational life, and also provides important data on issues of safety. The most recent case involving the process of extending the life of a PWR plant can be seen in Angra 1 Nuclear Power Plant by investing $ 27 million in the installation of a new reactor cover. The corrective action generated an extension of the useful life of Angra 1 estimated in twenty years, and a great savings compared to the cost of building a new plant and the decommissioning of the first, if it had reached the operation time out 40 years. The extension of the lifetime of a nuclear power plant must be accompanied by special attention from the most sensitive components of the systems to the aging process. After the application of the methodology (aging analysis of Containment Spray Injection System (CSIS)) proposed in this paper, it can be seen that increasing the probability of failure of each component, due to the aging process, generate an increased general unavailability of the system that contains these basic components. The final results obtained were as expected and can contribute to the maintenance policy, preventing premature aging in nuclear power systems.

  1. Het constitutionele perspectief en de coming of age van internationale organisaties = The constitutional perspective and the coming of age of international organizations

    NARCIS (Netherlands)

    Brölmann, C.M.

    2012-01-01

    This vignette traces the coming of age of international organizations as international legal actors. It argues that the political and legal appraisal of the international organization - and thus, its identity - since the rise of organizations in the mid-nineteenth century has passed through

  2. Enhanced biological nutrient removal in a simultaneous fermentation, denitrification and phosphate removal reactor using primary sludge as internal carbon source.

    Science.gov (United States)

    Zhang, Liang; Zhang, Shujun; Wang, Shuying; Wu, Chengcheng; Chen, Yinguang; Wang, Yayi; Peng, Yongzhen

    2013-04-01

    The production of volatile fatty acids (VFAs) from primary sludge and the subsequent application to improve biological nutrient removal has drawn much attention. In this study, a novel approach of using primary sludge as an additional carbon source was conducted in batch tests. The nitritation effluent was directly injected into the sludge fermentation reactor to achieve nitrogen removal. Complete denitrification could be realized in the combined reactor. Moreover, injecting nitrite not only promoted the sludge stabilization process, but also reduced the release of phosphate and ammonium during sludge stabilization. The novel process was further evaluated in a continuous system by treating sludge dewatering liquors. Under optimum conditions, 85% removal of ammonium and 75% of total nitrogen could be obtained using primary sludge, resulting in the suitable effluent for recycling into the inlet of the wastewater treatment plant.

  3. Stabilizing internal stress as the thermodynamic factor of martensite aging effects

    Energy Technology Data Exchange (ETDEWEB)

    Kosogor, Anna [Department of Radiophysics, Taras Shevchenko University, 03022 Kyiv (Ukraine); L' vov, Victor A., E-mail: victorlvov@univ.kiev.ua [Department of Radiophysics, Taras Shevchenko University, 03022 Kyiv (Ukraine); Institute of Magnetism, Vernadsky Str. 36-b, 03142 Kyiv (Ukraine); Soederberg, Outi; Hannula, Simo-Pekka [Department of Materials Science and Engineering, Aalto University School of Chemical Technology, PO Box 16200, Espoo FI-00076 Aalto (Finland)

    2011-05-15

    Highlights: > A martensite aging is accompanied by reconfiguration of the crystal defects system. > The reconfiguration causes an internal stressing of crystal lattice. > The internal stressing elevates the martensitic transformation temperature. > A theory that accounts for the internal stress adequately describes aging effects. > A quantitative description of aging effects observed in Au-Cd alloys is presented. - Abstract: The symmetry-conforming Landau theory has been developed for the description of interplay between the symmetry of the deformable crystal lattice and the configuration of crystallographic defects in martensitic alloys. For this purpose, the multi-component non-scalar order parameter describing the slow reconfiguration of defects after the deformation of crystal lattice has been introduced within the framework of the Landau theory of martensitic transformations (MTs). Due to the transformational equivalence of the configurational order parameter and strain/stress tensor components, the conception of stabilizing internal stress (SIS), which is linearly related to this parameter, has been formulated. The complete agreement between the developed theory and the symmetry-conforming short-range-order principle formulated by Ren and Otsuka has been established. The effect of stabilizing the product (martensitic) phase after aging has been described by considering the stress-temperature phase diagram, which was constructed by taking into account the time dependence of SIS. The applicability of the theory to the aging effects in the Au-Cd shape memory alloy has been demonstrated. The time dependencies of the experimentally observed slow changes in the MT temperatures, lattice parameters and yield stress values have been derived from the SIS conception.

  4. Preparation of calcium alginate microgel beads in an electrodispersion reactor using an internal source of calcium carbonate nanoparticles.

    Science.gov (United States)

    Zhao, Yinyan; Carvajal, M Teresa; Won, You-Yeon; Harris, Michael T

    2007-12-04

    An electrodispersion reactor has been used to prepare calcium alginate (Ca-alginate) microgel beads in this study. In the electrodispersion reactor, pulsed electric fields are utilized to atomize aqueous mixtures of sodium alginate and CaCO3 nanoparticles (dispersed phase) from a nozzle into an immiscible, insulating second liquid (continuous phase) containing a soluble organic acid. This technique combines the features of the electrohydrodynamic force driven emulsion processes and externally triggered gelations in microreactors (the droplets) ultimately to yield soft gel beads. The average particle size of the Ca-alginate gels generated by this method changed from 412 +/- 90 to 10 +/- 3 microm as the applied peak voltage was increased. A diagram depicting structural information for the Ca-alginate was constructed as a function of the concentrations of sodium alginate and CaCO3 nanoparticles. From this diagram, a critical concentration of sodium alginate required for sol-gel transformation was observed. The characteristic highly porous structure of Ca-alginate particles made by this technique appears suitable for microencapsulation applications. Finally, time scale analysis was performed for the electrodispersion processes that include reactions in the microreactor droplets to provide guidelines for the future employment of this technique. This electrodispersion reactor can be used potentially in the formation of many reaction-based microencapsulation systems.

  5. Short-sludge age EBPR process – Microbial and biochemical process characterisation during reactor start-up and operation

    DEFF Research Database (Denmark)

    Valverde Pérez, Borja; Wágner, Dorottya Sarolta; Lóránt, Bálint

    2016-01-01

    . In this paper, we report the start-up and operation of a short-SRT enhanced biological phosphorus removal (EBPR) system operated as a sequencing batch reactor (SBR) fed with preclarified municipal wastewater, which is supplemented with propionate. The microbial community was analysed via 16S rRNA amplicon...... of the Thiothrix taxon proliferated in the reactor, thereby leading to filamentous bulking (sludge volume index up to SVI = 1100 mL/g). Phosphorus removal deteriorated during this period, likely due to the out-competition of polyphosphate accumulating organisms (PAO) by sulphate reducing bacteria (SRB...

  6. Evaluation of CANDU NPP containment structure subjected to aging and internal pressure increase

    Energy Technology Data Exchange (ETDEWEB)

    Huang, Xu [Department of Civil Engineering, University of Toronto, Toronto M5S 1A4 (Canada); Kwon, Oh-Sung, E-mail: os.kwon@utoronto.ca [Department of Civil Engineering, University of Toronto, Toronto M5S 1A4 (Canada); Bentz, Evan [Department of Civil Engineering, University of Toronto, Toronto M5S 1A4 (Canada); Tcherner, Julia [Candu Energy Inc. a member of SNC-Lavalin Group, Mississauga L5K 1B1 (Canada)

    2017-04-01

    Highlights: • The aging effects on the performance of a nuclear containment structure is evaluated. • A numerical model of the structure is subjected to increasing internal pressure. • No through-thickness cracks are predicted under the design level internal pressure. • The structure is predicted to be ductile up to large internal pressure levels. - Abstract: The objective of this study is to investigate the long-term performance of a typical CANDU® containment structure. A three-dimensional nonlinear finite element model was built to realistically evaluate the performance of the structure under service load as well as a hypothetical beyond-design level internal pressure. Consideration is given to the time-dependent effects, such as shrinkage, creep, and relaxation of prestressing tendons, over a 60-year timeframe. In addition, the sensitivity of the response of the containment structure against support condition, internal temperature profile and temporary construction openings was also investigated. The accuracy of the numerical model was validated against structural measurements made during a routine leak rate test. The analysis results show that the containment structure would develop a ductile mechanism if the internal pressure significantly exceeded the design pressure. The pressure-deformation relationship of the structure is sensitive to the considered time-dependent parameters.

  7. H Reactor

    Data.gov (United States)

    Federal Laboratory Consortium — The H Reactor was the first reactor to be built at Hanford after World War II.It became operational in October of 1949, and represented the fourth nuclear reactor on...

  8. Age-related changes in predictive capacity versus internal model adaptability: electrophysiological evidence that individual differences outweigh effects of age

    Directory of Open Access Journals (Sweden)

    Ina eBornkessel-Schlesewsky

    2015-11-01

    Full Text Available Hierarchical predictive coding has been identified as a possible unifying principle of brain function, and recent work in cognitive neuroscience has examined how it may be affected by age–related changes. Using language comprehension as a test case, the present study aimed to dissociate age-related changes in prediction generation versus internal model adaptation following a prediction error. Event-related brain potentials (ERPs were measured in a group of older adults (60–81 years; n=40 as they read sentences of the form The opposite of black is white/yellow/nice. Replicating previous work in young adults, results showed a target-related P300 for the expected antonym (white; an effect assumed to reflect a prediction match, and a graded N400 effect for the two incongruous conditions (i.e. a larger N400 amplitude for the incongruous continuation not related to the expected antonym, nice, versus the incongruous associated condition, yellow. These effects were followed by a late positivity, again with a larger amplitude in the incongruous non-associated versus incongruous associated condition. Analyses using linear mixed-effects models showed that the target-related P300 effect and the N400 effect for the incongruous non-associated condition were both modulated by age, thus suggesting that age-related changes affect both prediction generation and model adaptation. However, effects of age were outweighed by the interindividual variability of ERP responses, as reflected in the high proportion of variance captured by the inclusion of by-condition random slopes for participants and items. We thus argue that – at both a neurophysiological and a functional level – the notion of general differences between language processing in young and older adults may only be of limited use, and that future research should seek to better understand the causes of interindividual variability in the ERP responses of older adults and its relation to cognitive

  9. A kinetic model for impact/sliding wear of pressurized water reactor internal components. Application to rod cluster control assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Zbinden, M.; Durbec, V.

    1996-12-01

    A new concept of industrial wear model adapted to components of nuclear plants is proposed. Its originality is to be supported, on one hand, by experimental results obtained via wear machines of relatively short operational times, and, on the other hand, by the information obtained from the operating feedback over real wear kinetics of the reactors components. The proposed model is illustrated by an example which corresponds to a specific real situation. The determination of the coefficients permitting to cover all assembly of configurations and the validation of the model in these configurations have been the object of the most recent work. (author). 34 refs.

  10. Aging phenomena in high-Si steels studied by internal friction

    Energy Technology Data Exchange (ETDEWEB)

    Ruiz, D. [Department of Metallurgy and Materials Science, Ghent University, Technologiepark 903, B-9052 Gent (Belgium) and Department of Subatomic and Radiation Physics, Ghent University, Proeftuinstraat 86, B-9000 Gent (Belgium)]. E-mail: Daniel.Ruiz@UGent.be; Rivera-Tovar, J.L. [Department of Metallurgy and Materials Science, Ghent University, Technologiepark 903, B-9052 Gent (Belgium); Facultad de Ingenieria Mecanica y Electrica, Universidad Autonoma de Nuevo Leon, A.P. 149-F, 66451 San Nicolas de los Garza, N.L. (Mexico); Segers, D. [Department of Subatomic and Radiation Physics, Ghent University, Proeftuinstraat 86, B-9000 Gent (Belgium); Vandenberghe, R.E. [Department of Subatomic and Radiation Physics, Ghent University, Proeftuinstraat 86, B-9000 Gent (Belgium); Houbaert, Y. [Department of Metallurgy and Materials Science, Ghent University, Technologiepark 903, B-9052 Gent (Belgium)

    2006-12-20

    Si-steels with various Si-contents (1.9-5.6 wt.%) have been analyzed by internal friction and compared with an ultra-low carbon steel. Measurements have been carried out immediately after different thermomechanical treatments to study a believed aging phenomenon. Adding Si lowers the Snoek peak of carbon and produces a new peak associated to the formation of Si-C pairs. For Si contents higher than 4.6 wt.%, another peak appears at very low frequencies, which can be attributed to a Zener relaxation of Si-atom pairs. A room-temperature aging effect has been detected in the Si-steels, but not in the ultra-low carbon steel. This aging is caused by the migration of C atoms to the structural defects and by formation of short-range order in the Fe-Si solution.

  11. Application of fuzzy neural networks for modeling of biodegradation and biogas production in a full-scale internal circulation anaerobic reactor.

    Science.gov (United States)

    Ruan, Jujun; Chen, Xiaohong; Huang, Mingzhi; Zhang, Tao

    2017-01-02

    This paper presents the development and evaluation of three fuzzy neural network (FNN) models for a full-scale anaerobic digestion system treating paper-mill wastewater. The aim was the investigation of feasibility of the approach-based control system for the prediction of effluent quality and biogas production from an internal circulation (IC) anaerobic reactor system. To improve FNN performance, fuzzy subtractive clustering was used to identify model's architecture and optimize fuzzy rule, and a total of 5 rules were extracted in the IF-THEN format. Findings of this study clearly indicated that, compared to NN models, FNN models had smaller RMSE and MAPE as well as bigger R for the testing datasets than NN models. The proposed FNN model produced smaller deviations and exhibited a superior predictive performance on forecasting of both effluent quality and biogas (methane) production rates with satisfactory determination coefficients greater than 0.90. From the results, it was concluded that FNN modeling could be applied in IC anaerobic reactor for predicting the biodegradation and biogas production using paper-mill wastewater.

  12. Study on Alignment Technology of CAP1400 Reactor Internals%CAP1400堆内构件对中检测技术研究

    Institute of Scientific and Technical Information of China (English)

    舒华安; 童庆军; 王盛; 郝磊; 郭健

    2016-01-01

    对中检测是CAP1400堆内构件制造装配的关键检测技术,针对传统光学对中方法的检测结果易受环境和人眼观测误差影响,且作业时间受限等缺点,开发了激光垂准—CCD光靶数据采集处理的新型对中检测方法。通过专用试验平台进行对中比对试验,结果表明该方法测量精度高,成功避免了传统方法的缺点,为堆内构件制造及核电站现场安装提供了新的对中检测方法。%The alignment measurement is the key technology of CAP1400 reactor interal manufacturing and assembly. Traditional method using micrometer collimating telescope has high requirements on the operating environment, operation time. The system of Laser Vertical-CCD optical target data acquisition and processing is developed. The comparison tests made through the special test platform show that the method has high accuracy of measurement. The disadvantages of the traditional optical measurement is avoided. A new method for manufacturing and installation of reactor internal is gained.

  13. Dosimetric impact evaluation of primary coolant chemistry of the internal tritium breeding cycle of a fusion reactor DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Velarde, M. [Instituto de Fusion Nuclear (DENIM), ETSII, Universidad Politecnica Madrid UPM, J. Gutierrez Abascal 2, Madrid 28006 (Spain); Sedano, L. A. [Asociacion Euratom-Ciematpara Fusion, Av. Complutense 22, 28040 Madrid (Spain); Perlado, J. M. [Instituto de Fusion Nuclear (DENIM), ETSII, Universidad Politecnica Madrid UPM, J. Gutierrez Abascal 2, Madrid 28006 (Spain)

    2008-07-15

    Tritium will be responsible for a large fraction of the environmental impact of the first generation of DT fusion reactors. Today, the efforts of conceptual development of the tritium cycle for DEMO are mainly centred in the so called Inner Breeding Tritium Cycle, conceived as guarantee of reactor fuel self-sufficiency. The EU Fusion Programme develops for the short term of fusion power technology two breeding blanket conceptual designs both helium cooled. One uses Li-ceramic material (HCPB, Helium-Cooled Pebble Bed) and the other a liquid metal eutectic alloy (Pb15.7Li) (HCLL, Helium-Cooled Lithium Lead). Both are Li-6 enriched materials. At a proper scale designs will be tested as Test Blanket Modules in ITER. The tritium cycles linked to both blanket concepts are similar, with some different characteristics. The tritium is recovered from the He purge gas in the case of HCPB, and directly from the breeding alloy through a carrier gas in HCLL. For a 3 GWth self-sufficient fusion reactor the tritium breeding need is few hundred grams of tritium per day. Safety and environmental impact are today the top priority design criteria. Dose impact limits should determine the key margins and parameters in its conception. Today, transfer from the cycle to the environment is conservatively assumed to be operating in a 1-enclosure scheme through the tritium plant power conversion system (intermediate heat exchangers and helium blowers). Tritium loss is caused by HT and T{sub 2} permeation and simultaneous primary coolant leakage through steam generators. Primary coolant chemistry appears to be the most natural way to control tritium permeation from the breeder into primary coolant and from primary coolant through SG by H{sub 2} tritium flux isotopic swamping or steel (EUROFER/INCOLOY) oxidation. A primary coolant chemistry optimization is proposed. Dynamic flow process diagrams of tritium fluxes are developed ad-hoc and coupled with tritiated effluents dose impact evaluations

  14. Internal derangements of the temporomandibular joint: findings in the pediatric age group

    Energy Technology Data Exchange (ETDEWEB)

    Katzberg, R.W.; Tallents, R.H.; Hayakawa, K.; Miller, T.L.; Goske, M.J.; Wood, B.P.

    1985-01-01

    Findings in 31 pediatric patients with pain and dysfunction of the temporomandibular joint (TMJ) are reported. The average age was 14 years and the average duration of symptoms was 21.4 months. Internal derangements were found in 29 patients (94%) and degenerative arthritis in 13 (42%). In 12 patients (39%), the problem could be traced to an injury to the jaw. Secondary condylar hypoplasia was associated with the meniscal abnormality in 3 patients (10%). Further awareness of internal derangements of the TMJ in the pediatric population should permit greater recognition of their etiology. It is important that threatment be initiated as soon as possible, not only to minimize the development of osseous disease in young adults but also to prevent facial growth deformities.

  15. Aarne Michaël Tallgren and the International Discussion on the Bronze Age of Russia

    Directory of Open Access Journals (Sweden)

    Timo Salminen

    2017-04-01

    Full Text Available This paper is on international scholarly discussion on the Bronze Age of Russia from 1908 until 1939, and in particular on the related role of the internationally renowned Finnish archaeologist Aarne Michaël Tallgren (1885–1945. How did a social network of researchers produce new interpretations and what were the key factors that distinguished the participants in the discussion? Was it a continuous process or a series of sudden changes? How did different ideological backgrounds influence the interpretations? In Western Europe, Tallgren’s most important interlocutors were Gero von Merhart, V Gordon Childe and Ellis H Minns, and in Russia V A Gorodcov and A A Spicyn. The paper is mainly based on correspondence between Tallgren and his colleagues.

  16. Bullying and symptoms among school-aged children: international comparative cross sectional study in 28 countries

    DEFF Research Database (Denmark)

    Due, Pernille; Holstein, Bjørn E; Lynch, John

    2005-01-01

    BACKGROUND: There have been no large-scale international comparisons on bullying and health among adolescents. This study examined the association between bullying and physical and psychological symptoms among adolescents in 28 countries. METHODS: This international cross-sectional survey included...... 123,227 students 11, 13 and 15 years of age from a nationally representative sample of schools in 28 countries in Europe and North America in 1997-98.The main outcome measures were physical and psychological symptoms. RESULTS: The proportion of students being bullied varied enormously across countries....... The lowest prevalence was observed among girls in Sweden (6.3%, 95% CI: 5.2-7.4), the highest among boys in Lithuania (41.4%, 95% CI 39.4-43.5). The risk of high symptom load increased with increasing exposure to bullying in all countries. In pooled analyses, with sex stratified multilevel logistic models...

  17. Research on friction coefficient of nuclear Reactor Vessel Internals Hold Down Spring: Stress coefficient test analysis method

    Energy Technology Data Exchange (ETDEWEB)

    Linjun, Xie, E-mail: linjunx@zjut.edu.cn [College of Mechanical Engineering, Zhejiang University of Technology, Hangzhou 310014 (China); Guohong, Xue; Ming, Zhang [Shanghai Nuclear Engineering Research & Design Institute, Shanghai 200233 (China)

    2016-08-01

    Graphical abstract: HDS stress coefficient test apparatus. - Highlights: • This paper performs mathematic deduction to the physical model of Hold Down Spring (HDS), establishes a mathematic model of axial load P and stress, stress coefficient and friction coefficient and designs a set of test apparatuses for simulating the pretightening process of the HDS for the first time according to a model similarity criterion. • The mathematical relation between the load and the strain is obtained about the HDS, and the mathematical model of the stress coefficient and the friction coefficient is established. So, a set of test apparatuses for obtaining the stress coefficient is designed according to the model scaling criterion and the friction coefficient of the K1000 HDS is calculated to be 0.336 through the obtained stress coefficient. • The relation curve between the theoretical load and the friction coefficient is obtained through analysis and indicates that the change of the friction coefficient f would influence the pretightening load under the condition of designed stress. The necessary pretightening load in the design process is calculated to be 5469 kN according to the obtained friction coefficient. Therefore, the friction coefficient and the pretightening load under the design conditions can provide accurate pretightening data for the analysis and design of the reactor HDS according to the operations. - Abstract: This paper performs mathematic deduction to the physical model of Hold Down Spring (HDS), establishes a mathematic model of axial load P and stress, stress coefficient and friction coefficient and designs a set of test apparatuses for simulating the pretightening process of the HDS for the first time according to a model similarity criterion. By carrying out tests and researches through a stress testing technique, P–σ curves in loading and unloading processes of the HDS are obtained and the stress coefficient k{sub f} of the HDS is obtained. So, the

  18. Reactor monitoring using antineutrino detectors

    Science.gov (United States)

    Bowden, N. S.

    2011-08-01

    Nuclear reactors have served as the antineutrino source for many fundamental physics experiments. The techniques developed by these experiments make it possible to use these weakly interacting particles for a practical purpose. The large flux of antineutrinos that leaves a reactor carries information about two quantities of interest for safeguards: the reactor power and fissile inventory. Measurements made with antineutrino detectors could therefore offer an alternative means for verifying the power history and fissile inventory of a reactor as part of International Atomic Energy Agency (IAEA) and/or other reactor safeguards regimes. Several efforts to develop this monitoring technique are underway worldwide.

  19. Aging Q3: an initiative to improve internal medicine residents' geriatrics knowledge, skills, and clinical performance.

    Science.gov (United States)

    Moran, William P; Zapka, Jane; Iverson, Patty J; Zhao, Yumin; Wiley, M Kathleen; Pride, Pamela; Davis, Kimberly S

    2012-05-01

    A growing number of older adults coupled with a limited number of physicians trained in geriatrics presents a major challenge to ensuring quality medical care for this population. Innovations to incorporate geriatrics education into internal medicine residency programs are needed. To meet this need, in 2009, faculty at the Medical University of South Carolina developed Aging Q(3)-Quality Education, Quality Care, and Quality of Life. This multicomponent initiative recognizes the need for improved geriatrics educational tools and faculty development as well as systems changes to improve the knowledge and clinical performance of residents. To achieve these goals, faculty employ multiple intervention strategies, including lectures, rounds, academic detailing, visual cues, and electronic medical record prompts and decision support. The authors present examples from specific projects, based on care areas including vision screening, fall prevention, and caring for patients with dementia, all of which are based on the Assessing Care of Vulnerable Elders quality indicators. The authors describe the principles driving the design, implementation, and evaluation of the Aging Q(3) program. They present data from multiple sources that illustrate the effectiveness of the interventions to meet the knowledge, skill level, and behavior goals. The authors also address major challenges, including the maintenance of the teaching and modeling interventions over time within the context of demanding primary care and inpatient settings. This organized, evidence-based approach to quality improvement in resident education, as well as faculty leadership development, holds promise for successfully incorporating geriatrics education into internal medicine residencies.

  20. Fast Breeder Reactor studies

    Energy Technology Data Exchange (ETDEWEB)

    Till, C.E.; Chang, Y.I.; Kittel, J.H.; Fauske, H.K.; Lineberry, M.J.; Stevenson, M.G.; Amundson, P.I.; Dance, K.D.

    1980-07-01

    This report is a compilation of Fast Breeder Reactor (FBR) resource documents prepared to provide the technical basis for the US contribution to the International Nuclear Fuel Cycle Evaluation. The eight separate parts deal with the alternative fast breeder reactor fuel cycles in terms of energy demand, resource base, technical potential and current status, safety, proliferation resistance, deployment, and nuclear safeguards. An Annex compares the cost of decommissioning light-water and fast breeder reactors. Separate abstracts are included for each of the parts.

  1. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2001-04-01

    The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised.

  2. Projected Population Size and Age Structure for Canada and Provinces: With and Without International Migration

    Directory of Open Access Journals (Sweden)

    Loh, Shirley

    2007-01-01

    Full Text Available EnglishThis paper examines the effect of net international migration on prospectivepopulation growth and age structure in Canada for the next 50 years. It alsoexamines the impact of international migration on provincial growth anddistribution. The procedure used in this study is by comparing two projectedpopulation scenarios, one with international migration and the other withoutinternational migration, based on the latest 2005-based population projections.The analysis of the scenarios shows that the assumed level of internationalmigration which is higher than the current level contributes to a continuousincrease in population over the next 50 years, but has limited effect to prevent oroffset the overall aging trend.FrenchCe document examine l’effet de la migration internationale sur la croissanceprospective de la population et la structure par âge au Canada pour les 50prochaines années. Il examine aussi l’impact de la migration internationale surla croissance et la distribution provinciales. La démarche utilisée dans cetteétude est de comparer les deux scénarios de population projetée l’un avecmigration internationale et l’autre sans migration internationale, selon lesdernières projections de population de 2005. L’analyse des scénarios démontreque le niveau présumé de migration internationale, plus élevé que le niveauactuel, contribue à une croissance continuelle de la population au cours des 50prochaines années, mais a peu d’effet quand à la prévention ou la tendance auvieillissement.

  3. Reactor safeguards

    CERN Document Server

    Russell, Charles R

    1962-01-01

    Reactor Safeguards provides information for all who are interested in the subject of reactor safeguards. Much of the material is descriptive although some sections are written for the engineer or physicist directly concerned with hazards analysis or site selection problems. The book opens with an introductory chapter on radiation hazards, the construction of nuclear reactors, safety issues, and the operation of nuclear reactors. This is followed by separate chapters that discuss radioactive materials, reactor kinetics, control and safety systems, containment, safety features for water reactor

  4. Reactor operation

    CERN Document Server

    Shaw, J

    2013-01-01

    Reactor Operation covers the theoretical aspects and design information of nuclear reactors. This book is composed of nine chapters that also consider their control, calibration, and experimentation.The opening chapters present the general problems of reactor operation and the principles of reactor control and operation. The succeeding chapters deal with the instrumentation, start-up, pre-commissioning, and physical experiments of nuclear reactors. The remaining chapters are devoted to the control rod calibrations and temperature coefficient measurements in the reactor. These chapters also exp

  5. Internal Consistency and Associated Characteristics of Informant Discrepancies in Clinic Referred Youths Age 11 to 17 Years

    Science.gov (United States)

    De Los Reyes, Andres; Youngstrom, Eric A.; Pabon, Shairy C.; Youngstrom, Jennifer K.; Feeny, Norah C.; Findling, Robert L.

    2011-01-01

    In this study, we examined the internal consistency of informant discrepancies in reports of youth behavior and emotional problems and their unique relations with youth, caregiver, and family characteristics. In a heterogeneous multisite clinic sample of 420 youths (ages 11-17 years), high internal consistency estimates were observed across…

  6. 14th International HLA and Immunogenetics Workshop: report on the immunogenetics of aging.

    Science.gov (United States)

    Naumova, E; Pawelec, G; Ivanova, M; Constantinescu, I; Bogunia-Kubik, K; Lange, A; Qguz, F; Carin, M

    2007-04-01

    The 'Immunogenetics of Aging' is a newly included component within the 14th International HLA and Immunogenetics Workshop. The aim of this component was to determine the contribution of human leukocyte antigen (HLA), cytokine genes and other major histocompatibility complex-encoded loci to successful aging and to determine an increased capacity to reach the extreme limits of life span. Two main data sets from four European populations were included in this study: unrelated healthy elderly individuals and ethnically matched young controls, and families with longevity members. Analysis was focused on HLA class I and II and cytokine gene polymorphisms. Preliminary results showed increased frequencies of DRB1*11- and DRB*16-associated haplotypes that were found to be protective for autoimmune diseases in some populations. Additionally, in families with longevity members, alleles and haplotypes positively associated with autoimmunity were not observed. Analysis of cytokine gene polymorphisms showed prevalence of anti-inflammatory profiles in healthy elderly individuals. Inheritance of extended haplotypes in families with longevity members allowed the identification of immunogenetic profiles that could be predictive for longevity. These preliminary studies indicate the relevance of genes regulating immune functions in human longevity and the importance of clarifying further their impact in successful aging.

  7. Harmonizing Measures of Cognitive Performance Across International Surveys of Aging Using Item Response Theory.

    Science.gov (United States)

    Chan, Kitty S; Gross, Alden L; Pezzin, Liliana E; Brandt, Jason; Kasper, Judith D

    2015-12-01

    To harmonize measures of cognitive performance using item response theory (IRT) across two international aging studies. Data for persons ≥65 years from the Health and Retirement Study (HRS, N = 9,471) and the English Longitudinal Study of Aging (ELSA, N = 5,444). Cognitive performance measures varied (HRS fielded 25, ELSA 13); 9 were in common. Measurement precision was examined for IRT scores based on (a) common items, (b) common items adjusted for differential item functioning (DIF), and (c) DIF-adjusted all items. Three common items (day of date, immediate word recall, and delayed word recall) demonstrated DIF by survey. Adding survey-specific items improved precision but mainly for HRS respondents at lower cognitive levels. IRT offers a feasible strategy for harmonizing cognitive performance measures across other surveys and for other multi-item constructs of interest in studies of aging. Practical implications depend on sample distribution and the difficulty mix of in-common and survey-specific items. © The Author(s) 2015.

  8. Advanced Non-Destructive Assessment Technology to Determine the Aging of Silicon Containing Materials for Generation IV Nuclear Reactors

    Science.gov (United States)

    Koenig, T. W.; Olson, D. L.; Mishra, B.; King, J. C.; Fletcher, J.; Gerstenberger, L.; Lawrence, S.; Martin, A.; Mejia, C.; Meyer, M. K.; Kennedy, R.; Hu, L.; Kohse, G.; Terry, J.

    2011-06-01

    To create an in-situ, real-time method of monitoring neutron damage within a nuclear reactor core, irradiated silicon carbide samples are examined to correlate measurable variations in the material properties with neutron fluence levels experienced by the silicon carbide (SiC) during the irradiation process. The reaction by which phosphorus doping via thermal neutrons occurs in the silicon carbide samples is known to increase electron carrier density. A number of techniques are used to probe the properties of the SiC, including ultrasonic and Hall coefficient measurements, as well as high frequency impedance analysis. Gamma spectroscopy is also used to examine residual radioactivity resulting from irradiation activation of elements in the samples. Hall coefficient measurements produce the expected trend of increasing carrier concentration with higher fluence levels, while high frequency impedance analysis shows an increase in sample impedance with increasing fluence.

  9. Internal Jugular Vein Cross-Sectional Area Enlargement Is Associated with Aging in Healthy Individuals.

    Directory of Open Access Journals (Sweden)

    Christopher Magnano

    Full Text Available Internal jugular vein (IJV narrowing has been implicated in central nervous system pathologies, however normal physiological age- and gender-related IJV variance in healthy individuals (HIs has not been adequately assessed.We assessed the relationship between IJV cross-sectional area (CSA and aging.This study involved 193 HIs (63 males and 130 females who received 2-dimensional magnetic resonance venography at 3T. The minimum CSA of the IJVs at cervical levels C2/C3, C4, C5/C6, and C7/T1 was obtained using a semi-automated contouring-thresholding technique. Subjects were grouped by decade. Pearson and partial correlation (controlled for cardiovascular risk factors, including hypertension, heart disease, smoking and body mass index and analysis of variance analyses were used, with paired t-tests comparing side differences.Mean right IJV CSA ranges were: in males, 41.6 mm2 (C2/C3 to 82.0 mm2 (C7/T1; in females, 38.0 mm2 (C2/C3 to 62.3 mm2 (C7/T1, while the equivalent left side ranges were: in males, 28.0 mm2 (C2/C3 to 52.2 mm2 (C7/T1; in females, 27.2 mm2 (C2/C3 to 47.8 mm2 (C7/T1. The CSA of the right IJVs was significantly larger (p<0.001 than the left at all cervical levels. Controlling for cardiovascular risk factors, the correlation between age and IJV CSA was more robust in males than in the females for all cervical levels.In HIs age, gender, hand side and cervical location all affect IJV CSA. These findings suggest that any definition of IJV stenosis needs to account for these factors.

  10. Reactor Neutrinos

    OpenAIRE

    Soo-Bong Kim; Thierry Lasserre; Yifang Wang

    2013-01-01

    We review the status and the results of reactor neutrino experiments. Short-baseline experiments have provided the measurement of the reactor neutrino spectrum, and their interest has been recently revived by the discovery of the reactor antineutrino anomaly, a discrepancy between the reactor neutrino flux state of the art prediction and the measurements at baselines shorter than one kilometer. Middle and long-baseline oscillation experiments at Daya Bay, Double Chooz, and RENO provided very ...

  11. BOILING REACTORS

    Science.gov (United States)

    Untermyer, S.

    1962-04-10

    A boiling reactor having a reactivity which is reduced by an increase in the volume of vaporized coolant therein is described. In this system unvaporized liquid coolant is extracted from the reactor, heat is extracted therefrom, and it is returned to the reactor as sub-cooled liquid coolant. This reduces a portion of the coolant which includes vaporized coolant within the core assembly thereby enhancing the power output of the assembly and rendering the reactor substantially self-regulating. (AEC)

  12. Preparing for Life in a Digital Age: The IEA International Computer and Information Literacy Study International Report

    National Research Council Canada - National Science Library

    Fraillon, Julian; Ainley, John; Schulz, Wolfram; Friedman, Tim; Gebhardt, Eveline

    2014-01-01

    ... this regard? The IEA International Computer and Information Literacy Study (ICILS) responded to this question by studying the extent to which young people have developed computer and information literacy (CIL...

  13. Development of a Versatile Ultrasonic Internal Pipe/Vessel Component Monitor for In-Service Inspection of Nuclear Reactor Components

    Energy Technology Data Exchange (ETDEWEB)

    Searfass, Clifford T. [Structural Integrity Associates, Inc., State College, PA (United States); Malinowski, Owen M. [Structural Integrity Associates, Inc., State College, PA (United States); Van Velsor, Jason K. [Structural Integrity Associates, Inc., State College, PA (United States)

    2015-03-22

    The stated goal of this work was to develop a versatile system which could accurately measure vessel and valve internal vibrations and cavitation formation under in-service conditions in nuclear power plants, ultrasonically. The developed technology will benefit the nuclear power generation industry by allowing plant operators to monitor valve and vessel internals during operation. This will help reduce planned outages and plant component failures. During the course of this work, Structural Integrity Associates, Inc. gathered information from industry experts that target vibration amplitudes to be detected should be in the range of 0.001-in to 0.005-in (0.025-mm to 0.127-mm) and target vibration frequency ranges which should be detected were found to be between 0-Hz and 300-Hz. During the performed work, an ultrasonic measuring system was developed which utilized ultrasonic pulse-echo time-of-flight measurements to measure vibration frequency and amplitude. The developed system has been shown to be able to measure vibration amplitudes as low as 0.0008-in (0.020-mm) with vibration frequencies in the range of 17-Hz to 1000-Hz. Therefore, the developed system was able to meet the industry needs for vibration measurement. The developed ultrasonic system was also to be able to measure cavitation formation by monitoring the received ultrasonic time- and frequency-domain signals. This work also demonstrated the survivability of commercially available probes at temperatures up to 300-F for several weeks.

  14. Conference Summary: First International Conference on Global Warming and the Next Ice Age

    Science.gov (United States)

    Wetzel, Peter J.; Chylek, Petr; Lesins, Glen; Starr, David OC. (Technical Monitor)

    2002-01-01

    The First International Conference on Global Warming and the Next Ice Age was convened in Halifax, Nova Scotia, August 19-24, 2001. The conference program began each day with a 30 minute live classical music performances of truly international quality before the beginning business. Ample time for panel discussions was also scheduled. The general public was invited to attend and participate in a special evening panel session on the last day of the conference. The unusual and somewhat provocative title of the conference was designed to attract diverse views on global climate change. This summary attempts to accurately reflect the tone and flavor of the lively discussions which resulted. Presentations ranged from factors forcing current climate to those in effect across the span of time from the Proterozoic "snowball Earth" epoch to 50,000 years in the future. Although, as should be expected, attendees at the conference arrived with opinions on some of the controversial issues regarding climate change, and no-one openly admitted to a 'conversion' from their initial point of view, the interdisciplinary nature of the formal presentations, poster discussions, panels, and abundant informal discourse helped to place the attendees' personal perspectives into a broader, more diversified context.

  15. Experiences of violence across life course and its effects on mobility among participants in the International Mobility in Aging Study

    OpenAIRE

    Guedes, Dimitri Taurino; Vafaei, Afshin; Alvarado, Beatriz Eugenia; Curcio, Carmen Lucia; Guralnik, Jack M.; Zunzunegui, María Victoria; Guerra,Ricardo Oliveira

    2016-01-01

    Background Life course exposure to violence may lead to disability in old age. We examine associations and pathways between life course violence and mobility disability in older participants of the International Mobility in Aging Study (IMIAS). Methods A cross-sectional study using IMIAS 2012 baseline. Men and women aged 65–74 years were recruited at 5 cities (n=1995): Kingston and Saint-Hyacinthe (Canada), Tirana (Albania), Manizales (Colombia) and Natal (Brazil). Mobility was assessed by th...

  16. Assessing the Age of an Asteroid's Surface with Data from the International Rosetta Mission

    Science.gov (United States)

    Lopez, Juan Carlos

    2011-01-01

    Rosetta is an international mission led by the European Space Agency (ESA) with key support and instrumentation from the National Aeronautics and Space Administration (NASA). Rosetta is currently on a ten-year mission to catch comet 67P/Churyumov-Gerasimenko (C-G); throughout its voyage, the spacecraft has performed flybys of two main belt asteroids (MBA): Steins and Lutetia. Data on the physical, chemical, and geological properties of these asteroids are currently being processed and analyzed. Accurate interpretation of such data is fundamental in the success of Rosetta's mission and overall objectives. Post-flyby data analyses strive to correlate the size, shape, volume, and rotational rate of Lutetia, in addition to interpreting its multi-color imagining, albedo, and spectral mapping. Although advancements in science have contributed to the examination of celestial bodies, methods to analyze asteroids remain largely empirical, not semi-empirical, nor ab initio. This study aims to interpret and document the scientific methods currently utilized in the characterization of asteroid (21) Lutetia in order to render these processes and methods accessible to the public. Examples include a standardized technique for assessing the age of an asteroid surface, complete with clickable reference maps, methodology of grouping surface characteristics together, and a standardized power law equation for the age. Other examples include determining the density of an object. Context for what both density and age mean is a bi-product of this study. Results of the study will aid in the development of pedagogical material on asteroids for public use, and in creation of an academic database for selected targets that might be used as a reference.

  17. The effects of thermal aging on material behavior and strength of CF8M in nuclear reactor coolant system

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Jae Do; Lee, Yong Seon; Nam, Uk Hui; Park, Jung Cheol; Pae, Yong Tak; In, Jae Hyeon; Woo, Seung Wan [Yeungnam Univ., Gyeongsan (Korea, Republic of)

    1998-03-15

    The following investigations are performed in order to estimate the mechanism of the thermal integrity, and the life prediction. The CF8M is observed a brittle behavior in the range of 475 .deg. C. The five classes of the thermally aged CF8M specimen are prepared using an artificially accelerated aging method. Namely, after the specimen are held for 100, 300, 900, 1800 and 3600 hrs. at 430 .deg. C respectively, the specimen are water cooled to room temperature. The impact energy variations are measures for both the aged and virgin specimen at -173, -70, -32, 27 and 100 .deg. C respectively through the Charpy impact tests in addition to the hardness tests. The tests results are to be a guide line to predict the life of CF8M, a RCS component material caused by thermal aging. The critical flaw size can be estimated by KIC obtained from the impact energy.

  18. Fusion Reactor Materials

    Energy Technology Data Exchange (ETDEWEB)

    Decreton, M

    2002-04-01

    The objective of SCK-CEN's programme on fusion reactor materials is to contribute to the knowledge on the radiation-induced behaviour of fusion reactor materials and components as well as to help the international community in building the scientific and technical basis needed for the construction of the future reactor. Ongoing projects include: the study of the mechanical and chemical (corrosion) behaviour of structural materials under neutron irradiation and water coolant environment; the investigation of the characteristics of irradiated first wall material such as beryllium; investigations on the management of materials resulting from the dismantling of fusion reactors including waste disposal. Progress and achievements in these areas in 2001 are discussed.

  19. Membrane reactor. Membrane reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shindo, Y.; Wakabayashi, K. (National Chemical Laboratory for Industry, Tsukuba (Japan))

    1990-08-05

    Many reaction examples were introduced of membrane reactor, to be on the point of forming a new region in the field of chemical technology. It is a reactor to exhibit excellent function, by its being installed with membrane therein, and is generally classified into catalyst function type and reaction promotion type. What firstly belongs to the former is stabilized zirconia, where oxygen, supplied to the cathodic side of membrane with voltage, impressed thereon, becomes O {sup 2 {minus}} to be diffused through the membrane and supplied, as variously activated oxygenous species, on the anodic side. Examples with many advantages can be given such as methane coupling, propylene oxidation, methanating reaction of carbon dioxide, etc. Apart, palladium film and naphion film also belong to the former. While examples of the latter comprise, among others, decomposition of hydrogen sulfide by porous glass film and dehydrogenation of cyclohexane or palladium alloy film, which are expected to be developed and materialized in the industry. 33 refs., 8 figs.

  20. The effects of thermal aging on material behavior and strength of CF8M in nuclear reactor coolant system

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Jae Do; Lee, Yong Seon; Park, Jung Cheol; In, Jae Hyeon; Woo, Seung Wan; Pae, Yong Tak; Nam, Uk Hui; Park, Yun Won [Yeungnam Univ., Gyeongsan (Korea, Republic of)

    1999-03-15

    The following investigations are performed in order to estimate the mechanism of the structural integrity, and the life prediction. The CF8M is observed a brittle behavior in the range of 475 .deg. C. The five classes of the thermally aged CF8M specimen are prepared using an artificially accelerated aging method. Namely, after the specimen are held for 100, 300, 900, 1800 and 3600 hrs. at 430 .deg. C respectively, the specimen are water cooled to room temperature. In addition to the thermally aged specimens the specimens associated with {delta}-phase degradation are prepared. After the specimens are maintained for 20 min, 5, 15, 50 and 150 hrs. at 700 .deg. C, respectively. which is in the range of {delta}-phase degradation, all specimens are cooled in water. The impact energy variations are measured for both the aged and virgin specimen at -173, -70, -32, 27 and 100 .deg. C, respectively, through the Charpy impact tests in addition to the hardness tests. The characteristics of the fatigue crack growth and low cycle fatigue tests are investigated using both aged and virgin specimens. Also fractured surfaces of the specimen are observed using the scanning electronic microscopy. J-R curve and J{sub IC} of the aged and virgin specimens are found J{sub IC} in order to predict the critical flaw size and fatigue life.

  1. The effects of thermal aging on material behavior and strength of CF8M in nuclear reactor coolant system

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Jae Do; Lee, Yong Seon; Park, Jung Cheol; In, Jae Hyeon; Woo, Seung Wan; Pae, Yong Tak; Nam, Uk Hui; Park, Yun Won [Yeungnam Univ., Gyeongsan (Korea, Republic of)

    1999-03-15

    The following investigations are performed in order to estimate the mechanism of the structural integrity, and the life prediction. The CF8M is observed a brittle behavior in the range of 475 .deg. C. The five classes of the thermally aged CF8M specimen are prepared using an artificially accelerated aging method. Namely, after the specimen are held for 100, 300, 900, 1800 and 3600 hrs. at 430 .deg. C respectively, the specimen are water cooled to room temperature. In addition to the thermally aged specimens the specimens associated with {delta}-phase degradation are prepared. After the specimens are maintained for 20 min, 5, 15, 50 and 150 hrs. at 700 .deg. C, respectively. which is in the range of {delta}-phase degradation, all specimens are cooled in water. The impact energy variations are measured for both the aged and virgin specimen at -173, -70, -32, 27 and 100 .deg. C, respectively, through the Charpy impact tests in addition to the hardness tests. The characteristics of the fatigue crack growth and low cycle fatigue tests are investigated using both aged and virgin specimens. Also fractured surfaces of the specimen are observed using the scanning electronic microscopy. J-R curve and J{sub IC} of the aged and virgin specimens are found J{sub IC} in order to predict the critical flaw size and fatigue life.

  2. Optimization of Mechanical Process of PWR Reactor Internals Baffle%压水堆堆内构件的围板机械加工工艺优化

    Institute of Scientific and Technical Information of China (English)

    青辉

    2013-01-01

    The structure characteristics and functions of baffles in the reactor internals in PWR were briefly introduced.The mechanical processing characteristic of baffle were described,which includes technique characters of milling and planing,processing difficulties of austenitic stainless steels,factors effecting the quality of mechanical processing,causes of residual stress produced in mechanical processing and their effects on products,etc.A preferable process was obtained through process optimization which increased the processing quality and pass rate of product and made the products meet the design and engineering requirements.%对压水堆堆内构件的围板的结构特点和功能进行了简述,介绍围板的机械加工工艺特点,包括铣削和刨削加工工艺的特点、奥氏体不锈钢加工难点、影响机加工质量的因素、机械加工残余应力产生的原因及其对产品的影响等.围板通过适宜的工艺优化方案提高其产品机加工质量和合格率,使产品达到其设计要求和满足其用途.

  3. Characteristic evaluation of high compression seismic isolator for International Thermonuclear Experimental Reactor (ITER). Verification test of sub-scaled rubber bearings. Contract research

    Energy Technology Data Exchange (ETDEWEB)

    Takahashi, Hiroyuki [Hitachi Ltd., Tokyo (Japan); Nakahira, Masataka [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Yabana, Shuichi; Matsuda, Akihiro; Ohtori, Yasuki [Central Research Inst. of Electric Power Industry, Tokyo (Japan)

    2001-11-01

    The International Thermonuclear Experimental Reactor (ITER) is designed to withstand the seismic load of 2 m/s{sup 2} at the ground level as a standard seismic condition. In case of severe seismic load over 2 m/s{sup 2}, an application of the seismic isolation to the tokamak building is studied so as to reduce the seismic load below 2 m/s{sup 2}. The seismic isolation with high compressive pressure of 7.35MPa to 14.7MPa is considered as a candidate, because the tokamak weight is large to the building size and the number of seismic isolator (rubber bearing) is limited in the available space of the building. Although many studies were executed in the past in order to apply the seismic isolation to the nuclear plant, the test data can not be applied to the ITER due to low compressive pressure of about 2.45MPa to 4.90MPa. Based on the above, it is therefore necessary to evaluate the various kinds of dynamic and mechanical characteristics of the rubber bearings under the high compressive pressure and to obtain the database for the design of the seismic isolation system of the ITER. The report describes the summary of the test results of the sub-scaled rubber bearings executed under the high compression condition in 1997 to 1999. (author)

  4. The Origins of Mental Toughness – Prosocial Behavior and Low Internalizing and Externalizing Problems at Age 5 Predict Higher Mental Toughness Scores at Age 14

    Science.gov (United States)

    Sadeghi Bahmani, Dena; Hatzinger, Martin; Gerber, Markus; Lemola, Sakari; Clough, Peter J.; Perren, Sonja; von Klitzing, Kay; von Wyl, Agnes; Holsboer-Trachsler, Edith; Brand, Serge

    2016-01-01

    Background: The concept of mental toughness (MT) has gained increasing importance among groups other than elite athletes by virtue of its psychological importance and explanatory power for a broad range of health-related behaviors. However, no study has focused so far on the psychological origins of MT. Therefore, the aims of the present study were: to explore, to what extent the psychological profiles of preschoolers aged five were associated with both (1) MT scores and (2) sleep disturbances at age 14, and 3) to explore possible gender differences. Method: Nine years after their first assessment at age five (preschoolers), a total of 77 adolescents (mean age: 14.35 years; SD = 1.22; 42% females) took part in this follow-up study. At baseline, both parents and teachers completed the Strengths and Difficulties Questionnaire (SDQ), covering internalizing and externalizing problems, hyperactivity, negative peer relationships, and prosocial behavior. At follow-up, participants completed a booklet of questionnaires covering socio-demographic data, MT, and sleep disturbances. Results: Higher prosocial behavior, lower negative peer relationships, and lower internalizing and externalizing problems at age five, as rated by parents and teachers, were associated with self-reported higher MT and lower sleep disturbances at age 14. At age 14, and relative to males, females had lower MT scores and reported more sleep disturbances. Conclusion: The pattern of results suggests that MT traits during adolescence may have their origins in the pre-school years. PMID:27605919

  5. Research on the Stress Calculation of Reactor Vessel Internal Upper Support Assembly under Accident Case%堆内构件上支承组件在事故工况下的应力计算

    Institute of Scientific and Technical Information of China (English)

    赵文清

    2014-01-01

    Reactor vessel internal upper support assembly adopt shell element combined with beam ele-ment to create model,shell element to create model and solid element to create model,which carry through respective finite element analysis calculation and stress evaluation.The stress calculation effect of reactor vessel internal upper support assembly adopting solid element model created is conservative and high precision by means of modification and simplification about preexistent stress analysis method related to reactor vessel internal,which can satisfy the demand of RCC-M criterion.The calculation method of reactor vessel internal upper support assembly adopting solid element to create model is simplification and practicality by comparing with different creating model mode about stress calculation mentioned above, which may apply to the stress analysis and evaluation of reactor vessel internal assembly related to differ-ent reactor core.%堆内构件上支承组件采用不同的建模方法,分别采用壳单元和梁单元相组合的建模模式、壳单元和壳单元相组合的建模模式、实体单元建模的模式,对堆内构件上支承组件进行了有限元应力计算,比较了不同建模模式下应力计算的各自特点,堆内构件上支承组件实体单元建模模式应力计算结果精确并能满足RC C-M规范应力评定要求,壳单元和梁单元相组合的建模模式、壳单元和壳单元相组合的建模模式应力计算结果保守且应力评定需等效处理其计算结果。堆内构件上支承组件采用整体实体单元全模型建模的计算方法,计算精确且应力评定简单直接,它可应用于其他工况和不同堆芯堆内构件应力计算及其应力评定。

  6. Treatment of landfill leachates by waste iron-scraps coupling with aged-refuse-based reactor%废铁屑耦合矿化垃圾反应床处理填埋场渗滤液研究

    Institute of Scientific and Technical Information of China (English)

    安晓英; 何岩; 陶锐锋; 李真; 郭翠香; 王力刚

    2011-01-01

    为了构建合理的废铁屑耦合矿化垃圾反应床工艺,分别进行了停留时间对铁屑处理填埋场渗滤液和矿化垃圾反应床出水的影响对比分析.并据此研究了前置铁屑固定床耦合矿化垃圾反应床处理填埋场渗滤液的效能.结果表明,在不进行pH调节的条件下,前置铁屑固定床和间歇曝气沉淀具有明显强化反应床体系去除COD和氨氮的能力,出水COD和氨氮分别小于150 mg/L和5 mg/L.随着铁屑内电解反应器的不断完善,前置铁屑固定床和间歇曝气沉淀可作为强化矿化垃圾反应床处理填埋场渗滤液的有效途径,为经济有效地处理填埋场渗滤液提供了一条新思路.%In order to build the optimal process of waste iron scraps coupling with aged-refuse-based reactor, the influences of retention time on the iron-scraps-based treatment of landfill leachate and the effluent from the aged-refuse-based reactor have been contrasted and analyzed, respectively. Subsequently, the treatment effect of the preceding waste-iron-scraps coupling with the aged-refuse-based reactor on the landfill leachate is investigated. The results show that without pH adjustment,the preceding iron-scraps-fixed reactor and the intermittent aeration-sedimentation have apparent capacity for strengthening the removal performance of the aged-refuse-based reactor in terms of COD and ammonium-nitrogen. The COD and ammonium-nitrogen concentrations in the effluent are less than 150 mg/L and 5 mg/L. With the continuous improvement in the iron-scrap-fixed reactor,the preceding iron-scraps-fixed reactor and the intermittent aeration-sedimentation could be considered as a promising way in strengthening the removal performance of the aged-refuse-based reactor,which is of great significance for the economic and effective treatment of landfill leachates.

  7. Paleomagnetic correlation and ages of basalt flow groups in coreholes at and near the Naval Reactors Facility, Idaho National Laboratory, Idaho

    Science.gov (United States)

    Champion, Duane E.; Davis, Linda C.; Hodges, Mary K.V.; Lanphere, Marvin A.

    2013-01-01

    Paleomagnetic inclination and polarity studies were conducted on subcore samples from eight coreholes located at and near the Naval Reactors Facility (NRF), Idaho National Laboratory (INL). These studies were used to characterize and to correlate successive stratigraphic basalt flow groups in each corehole to basalt flow groups with similar paleomagnetic inclinations in adjacent coreholes. Results were used to extend the subsurface geologic framework at the INL previously derived from paleomagnetic data for south INL coreholes. Geologic framework studies are used in conceptual and numerical models of groundwater flow and contaminant transport. Sample handling and demagnetization protocols are described, as well as the paleomagnetic data averaging process. Paleomagnetic inclination comparisons among NRF coreholes show comparable stratigraphic successions of mean inclination values over tens to hundreds of meters of depth. Corehole USGS 133 is more than 5 kilometers from the nearest NRF area corehole, and the mean inclination values of basalt flow groups in that corehole are somewhat less consistent than with NRF area basalt flow groups. Some basalt flow groups in USGS 133 are missing, additional basalt flow groups are present, or the basalt flow groups are at depths different from those of NRF area coreholes. Age experiments on young, low potassium olivine tholeiite basalts may yield inconclusive results; paleomagnetic and stratigraphic data were used to choose the most reasonable ages. Results of age experiments using conventional potassium argon and argon-40/argon-39 protocols indicate that the youngest and uppermost basalt flow group in the NRF area is 303 ± 30 ka and that the oldest and deepest basalt flow group analyzed is 884 ± 53 ka. A south to north line of cross-section drawn through the NRF coreholes shows corehole-to-corehole basalt flow group correlations derived from the paleomagnetic inclination data. From stratigraphic top to bottom, key results

  8. Reactor BR2. Introduction

    Energy Technology Data Exchange (ETDEWEB)

    Gubel, P

    2001-04-01

    The BR2 is a materials testing reactor and is still one of SCK-CEN's important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. During the last three years, the availability of the installation was maintained at an average level of 97.6 percent. In the year 2000, the reactor was operated for a total of 104 days at a mean power of 56 MW. In 2000, most irradiation experiments were performed in the CALLISTO PWR loop. The report describes irradiations achieved or under preparation in 2000, including the development of advanced facilities and concept studies for new programmes. An overview of the scientific irradiation programmes as well as of the R and D programme of the BR2 reactor in 2000 is given.

  9. Language and memory abilities of internationally adopted children from China: evidence for early age effects.

    Science.gov (United States)

    Delcenserie, Audrey; Genesee, Fred

    2014-11-01

    The goal of the present study was to examine if internationally adopted (IA) children from China (M = 10;8) adopted by French-speaking families exhibit lags in verbal memory in addition to lags in verbal abilities documented in previous studies (Gauthier & Genesee, 2011). Tests assessing verbal and non-verbal memory, language, non-verbal cognitive ability, and socio-emotional development were administered to thirty adoptees. Their results were compared to those of thirty non-adopted monolingual French-speaking children matched on age, gender, and socioeconomic status. The IA children scored significantly lower than the controls on language, verbal short-term memory, verbal working memory, and verbal long-term memory. No group differences were found on non-verbal memory, non-verbal cognitive ability, and socio-emotional development, suggesting language-specific difficulties. Despite extended exposure to French, adoptees may experience language difficulties due to limitations in verbal memory, possibly as a result of their delayed exposure to that language and/or attrition of the birth language.

  10. Analysis of internal structure changes in black human hair keratin fibers with aging using Raman spectroscopy.

    Science.gov (United States)

    Kuzuhara, Akio; Fujiwara, Nobuki; Hori, Teruo

    To investigate the internal structure changes in virgin black human hair keratin fibers due to aging, the structure of cross-sections at various depths of virgin black human hair (sections of new growth hair: 2 mm from the scalp) from a group of eight Japanese females in their twenties and another group of eight Japanese females in their fifties were analyzed using Raman spectroscopy. For the first time, we have succeeded in recording the Raman spectra of virgin black human hair, which had been impossible due to high melanin granule content. The key points of this method are to cross-section hair samples to a thickness of 1.50-microm, to select points at various depths of the cortex with the fewest possible melanin granules, and to optimize laser power, cross slit width as well as total acquisition time. The reproducibility of the Raman bands, namely the alpha-helix (alpha) content, the beta-sheet and/or random coil (beta/R) content, the disulfide (--SS--) content, and random coil content of two adjoining cross-sections of a single hair keratin fiber was clearly good. The --SS-- content of virgin black human hair from the Japanese females in their fifties for the cortex region decreased compared with that of the Japanese females in their twenties. On the other hand, the beta/R and alpha contents of the cortex region did not change.

  11. Management of the aging of critical safety-related concrete structures in light-water reactor plants

    Energy Technology Data Exchange (ETDEWEB)

    Naus, D.J.; Oland, C.B. (Oak Ridge National Lab., TN (USA)); Arndt, E.G. (Nuclear Regulatory Commission, Washington, DC (USA))

    1990-01-01

    The Structural Aging Program has the overall objective of providing the USNRC with an improved basis for evaluating nuclear power plant safety-related structures for continued service. The program consists of a management task and three technical tasks: materials property data base, structural component assessment/repair technology, and quantitative methodology for continued-service determinations. Objectives, accomplishments, and planned activities under each of these tasks are presented. Major program accomplishments include development of a materials property data base for structural materials as well as an aging assessment methodology for concrete structures in nuclear power plants. Furthermore, a review and assessment of inservice inspection techniques for concrete materials and structures has been complete, and work on development of a methodology which can be used for performing current as well as reliability-based future condition assessment of concrete structures is well under way. 43 refs., 3 tabs.

  12. Fast breeder reactors an engineering introduction

    CERN Document Server

    Judd, A M

    1981-01-01

    Fast Breeder Reactors: An Engineering Introduction is an introductory text to fast breeder reactors and covers topics ranging from reactor physics and design to engineering and safety considerations. Reactor fuels, coolant circuits, steam plants, and control systems are also discussed. This book is comprised of five chapters and opens with a brief summary of the history of fast reactors, with emphasis on international and the prospect of making accessible enormous reserves of energy. The next chapter deals with the physics of fast reactors and considers calculation methods, flux distribution,

  13. A homogenization method for the modal analysis of a nuclear reactor with internal structures modelling and fluid-structure interaction coupling; Une methode d'homogeneisation pour l'analyse modale d'un reacteur nucleaire avec modelisation des structures internes et de l'interaction fluide / structure

    Energy Technology Data Exchange (ETDEWEB)

    Sigrist, J.F. [DCN Propulsion, Service Technique et Scientifique, 44 - La Montagne (France); Broc, D. [CEA Saclay, Lab. d' Etude Mecanique et Sismique, 91 - Gif-sur-Yvette (France)

    2007-03-15

    A homogenization method is presented and validated in order to perform the dynamic analysis of a nuclear pressure vessel with a 'reduced' numerical model accounting for inertial fluid-structure coupling and describing the geometrical details of the internal structures, periodically embedded within the nuclear reactor. Homogenization techniques have been widely used in nuclear engineering to model confinement effects in reactor cores or tubes bundles. Application of such techniques to reactor internals is investigated in the present paper. The theory bases of the method are first recalled. Adaptation of the homogenization approach to the case of reactor internals is then exposed: it is shown that in such case, confinement effects can be modelled by a suitable modification of classical fluid-structure symmetric formulation. The method is then validated by comparison of 3D and 2D calculations. In the latter, a 'reduced' model with homogenized fluid is used, whereas in the former, a full finite element model of the nuclear pressure vessel with internal structures is elaborated. The homogenization approach is proved to be efficient from the numerical of view point and accurate from the physical point of view. Confinement effects in the industrial case can then be highlighted. (authors)

  14. The lunar Gruithuisen silicic extrusive domes: Topographic configuration, morphology, ages, and internal structure

    Science.gov (United States)

    Ivanov, M. A.; Head, J. W.; Bystrov, A.

    2016-07-01

    The Gruithuisen domes, situated on the western portion of the Imbrium basin rim, form three tall mountains (NW, Gamma, Delta) totaling ∼780 km3 in volume. The shapes of the domes are significantly different from that of mare-type domes elsewhere on the Moon. We use data from the Lunar Reconnaissance Orbiter (LRO) and Kaguya missions (LRO Lunar Orbiter Laser Altimeter, Lunar Reconnaissance Orbiter Camera, Diviner, and the Kaguya imager) to characterize the domes and assess models for their origin. The configuration of the domes (steep slopes, up to ∼18-20°) and their specific remote sensing characteristics (strong downturn in the UV, and results from the M3 and Diviner instruments) suggest that the domes formed by eruptions of highly viscous lava. The estimated surface volumes of the domes vary from ∼20 km3 (NW dome) to ∼290 km3 (Gamma dome) to ∼470 km3 (Delta dome). The domes occur on the portion of the Imbrium basin rim that is overlain by ejecta from the post-Imbrium Iridum crater. In some areas, relatively high albedo smooth volcanic plains are seen within the Iridum ejecta near the Gruithuisen domes, and low albedo mare deposits surround and embay the domes and Iridum crater. Dating of different units and features by crater counts indicates that impact melts from the Iridum basin are ∼3.9 Ga old, the domes Gamma and Delta are ∼3.8 Ga, and the ages of the plains near the domes vary from ∼2.3 to ∼3.6 Ga. A fresh impact crater exposes the internal structure of the Gamma dome. The most prominent features on the wall of the crater are rough, blocky layers that are typical of volcanic plains in the highlands and maria around the domes. The layers are interleaved with fine-grained materials of higher and lower albedo and the visible orientation of the layers changes over short (a few hundred meters) distances. These characteristics of the internal structure of the dome are consistent with eruptions of high viscosity lava (rough layers) that

  15. Light Water Reactor Sustainability Program: Evaluation of Localized Cable Test Methods for Nuclear Power Plant Cable Aging Management Programs

    Energy Technology Data Exchange (ETDEWEB)

    Glass, Samuel W. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Fifield, Leonard S. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Hartman, Trenton S. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2016-05-30

    This Pacific Northwest National Laboratory (PNNL) milestone report describes progress to date on the investigation of nondestructive test (NDE) methods focusing particularly on local measurements that provide key indicators of cable aging and damage. The work includes a review of relevant literature as well as hands-on experimental verification of inspection capabilities. As NPPs consider applying for second, or subsequent, license renewal (SLR) to extend their operating period from 60 years to 80 years, it important to understand how the materials installed in plant systems and components will age during that time and develop aging management programs (AMPs) to assure continued safe operation under normal and design basis events (DBE). Normal component and system tests typically confirm the cables can perform their normal operational function. The focus of the cable test program is directed toward the more demanding challenge of assuring the cable function under accident or DBE. Most utilities already have a program associated with their first life extension from 40 to 60 years. Regrettably, there is neither a clear guideline nor a single NDE that can assure cable function and integrity for all cables. Thankfully, however, practical implementation of a broad range of tests allows utilities to develop a practical program that assures cable function to a high degree. The industry has adopted 50% elongation at break (EAB) relative to the un-aged cable condition as the acceptability standard. All tests are benchmarked against the cable EAB test. EAB is a destructive test so the test programs must apply an array of other NDE tests to assure or infer the overall set of cable’s system integrity. These cable NDE programs vary in rigor and methodology. As the industry gains experience with the efficacy of these programs, it is expected that implementation practice will converge to a more common approach. This report addresses the range of local NDE cable tests that are

  16. Physics design of a 100 keV acceleration grid system for the diagnostic neutral beam for international tokamak experimental reactor.

    Science.gov (United States)

    Singh, M J; De Esch, H P L

    2010-01-01

    This paper describes the physics design of a 100 keV, 60 A H(-) accelerator for the diagnostic neutral beam (DNB) for international tokamak experimental reactor (ITER). The accelerator is a three grid system comprising of 1280 apertures, grouped in 16 groups with 80 apertures per beam group. Several computer codes have been used to optimize the design which follows the same philosophy as the ITER Design Description Document (DDD) 5.3 and the 1 MeV heating and current drive beam line [R. Hemsworth, H. Decamps, J. Graceffa, B. Schunke, M. Tanaka, M. Dremel, A. Tanga, H. P. L. De Esch, F. Geli, J. Milnes, T. Inoue, D. Marcuzzi, P. Sonato, and P. Zaccaria, Nucl. Fusion 49, 045006 (2009)]. The aperture shapes, intergrid distances, and the extractor voltage have been optimized to minimize the beamlet divergence. To suppress the acceleration of coextracted electrons, permanent magnets have been incorporated in the extraction grid, downstream of the cooling water channels. The electron power loads on the extractor and the grounded grids have been calculated assuming 1 coextracted electron per ion. The beamlet divergence is calculated to be 4 mrad. At present the design for the filter field of the RF based ion sources for ITER is not fixed, therefore a few configurations of the same have been considered. Their effect on the transmission of the electrons and beams through the accelerator has been studied. The OPERA-3D code has been used to estimate the aperture offset steering constant of the grounded grid and the extraction grid, the space charge interaction between the beamlets and the kerb design required to compensate for this interaction. All beamlets in the DNB must be focused to a single point in the duct, 20.665 m from the grounded grid, and the required geometrical aimings and aperture offsets have been calculated.

  17. Physics design of a 100 keV acceleration grid system for the diagnostic neutral beam for international tokamak experimental reactor

    Science.gov (United States)

    Singh, M. J.; De Esch, H. P. L.

    2010-01-01

    This paper describes the physics design of a 100 keV, 60 A H- accelerator for the diagnostic neutral beam (DNB) for international tokamak experimental reactor (ITER). The accelerator is a three grid system comprising of 1280 apertures, grouped in 16 groups with 80 apertures per beam group. Several computer codes have been used to optimize the design which follows the same philosophy as the ITER Design Description Document (DDD) 5.3 and the 1 MeV heating and current drive beam line [R. Hemsworth, H. Decamps, J. Graceffa, B. Schunke, M. Tanaka, M. Dremel, A. Tanga, H. P. L. De Esch, F. Geli, J. Milnes, T. Inoue, D. Marcuzzi, P. Sonato, and P. Zaccaria, Nucl. Fusion 49, 045006 (2009)]. The aperture shapes, intergrid distances, and the extractor voltage have been optimized to minimize the beamlet divergence. To suppress the acceleration of coextracted electrons, permanent magnets have been incorporated in the extraction grid, downstream of the cooling water channels. The electron power loads on the extractor and the grounded grids have been calculated assuming 1 coextracted electron per ion. The beamlet divergence is calculated to be 4 mrad. At present the design for the filter field of the RF based ion sources for ITER is not fixed, therefore a few configurations of the same have been considered. Their effect on the transmission of the electrons and beams through the accelerator has been studied. The OPERA-3D code has been used to estimate the aperture offset steering constant of the grounded grid and the extraction grid, the space charge interaction between the beamlets and the kerb design required to compensate for this interaction. All beamlets in the DNB must be focused to a single point in the duct, 20.665 m from the grounded grid, and the required geometrical aimings and aperture offsets have been calculated.

  18. A report on older-age bipolar disorder from the International Society for Bipolar Disorders Task Force

    DEFF Research Database (Denmark)

    Sajatovic, Martha; Strejilevich, Sergio A; Gildengers, Ariel G

    2015-01-01

    OBJECTIVES: In the coming generation, older adults with bipolar disorder (BD) will increase in absolute numbers as well as proportion of the general population. This is the first report of the International Society for Bipolar Disorder (ISBD) Task Force on Older-Age Bipolar Disorder (OABD). METHODS...

  19. Multifunctional reactors

    NARCIS (Netherlands)

    Westerterp, K.R.

    1992-01-01

    Multifunctional reactors are single pieces of equipment in which, besides the reaction, other functions are carried out simultaneously. The other functions can be a heat, mass or momentum transfer operation and even another reaction. Multifunctional reactors are not new, but they have received much

  20. Radiation effects and tritium technology for fusion reactors. Volume I. Proceedings of the international conference, Gatlinburg, Tennessee, October 1--3, 1975

    Energy Technology Data Exchange (ETDEWEB)

    Watson, J.S.; Wiffen, F.W.; Bishop, J.L.; Breeden, B.K. (eds.)

    1976-03-01

    Separate abstracts were prepared for the 29 included papers in Vol. I. The topics covered in this volume include swelling and microstructures in thermonuclear reactor materials. Some papers on modeling and damage analysis are included. (MOW)

  1. Development of Improved Models and Designs for Coated-Particle Gas Reactor Fuels -- Final Report under the International Nuclear Energy Research Initiative (I-NERI)

    Energy Technology Data Exchange (ETDEWEB)

    Petti, David [Idaho National Lab. (INL), Idaho Falls, ID (United States). Idaho National Engineering and Environmental Lab. (INEEL); Martin, Philippe [Commissariat a l' Energie Atomique et aux Energies Alternatives (CEA-Saclay), Gif-sur-Yvette (France); Phelip, Mayeul [Commissariat a l' Energie Atomique et aux Energies Alternatives (CEA-Saclay), Gif-sur-Yvette (France); Ballinger, Ronald [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)

    2004-12-01

    The objective of this INERI project was to develop improved fuel behavior models for gas reactor coated-particle fuels and to explore improved coated-particle fuel designs that could be used reliably at very high burnups and potentially in gas-cooled fast reactors. Project participants included the Idaho National Engineering Laboratory (INEEL), Centre Étude Atomique (CEA), and the Massachusetts Institute of Technology (MIT). To accomplish the project objectives, work was organized into five tasks.

  2. Numerical simulation and optimization of internal circulation anaerobic reactor by Fluent software%内循环厌氧反应器Fluent数值模拟与优化

    Institute of Scientific and Technical Information of China (English)

    蔡会勇; 刘永红; 李婷; 于兴峰

    2014-01-01

    采用Fluent技术对25L内循环(IC)厌氧反应器内气液两相的流动过程进行了数值模拟,考察了提升管直径(0.006m、0.009m、0.012m、0.015m、0.018m、0.021m )和反应器容积负荷[8.64kgCOD/(m3·d)、10.08kgCOD/(m3·d)、11.52kgCOD/(m3·d)、12.96kgCOD/(m3·d)、14.40kgCOD/(m3·d)、15.84kgCOD/(m3·d)]变化对其内循环量的影响。研究结果表明:①当反应器容积负荷为11.52kgCOD/(m3·d)、提升管直径为0.015m 时,该反应器内循环量达到最大值0.0079m3/h;②当提升管直径为0.015m、容积负荷为12.96kgCOD/(m3·d)时,该反应器内循环量增幅达到最大值9.28%,通过拟合获得了内循环量Y与产气量X间的经验关联式为Y=1.0514X+0.004。%The process of gas-liquid two-phase flow of internal circulation anaerobic (IC) reactor (25L)was simulated by Fluent software. The influence of internal circulation flow rate with different riser pipe diameters (0.006m,0.009m,0.012m,0.015m,0.018m,0.021m)and volume loading rates [8.64kgCOD/(m3·d) , 10.08kgCOD/(m3·d) , 11.52kgCOD/(m3·d) , 12.96kgCOD/(m3·d) , 14.40 kgCOD/(m3·d) , 15.84kgCOD/(m3·d)] were investigated. When volume loading rate was 11.52kgCOD/(m3·d),and riser diameter was 0.015m,a maximum internal circulation flow rate of 0.0079m3/h was observed. When riser diameter was 0.015m,and volume loading rate was 12.96 kgCOD/(m3·d) , a maximum increase amplitude of 9.28% was observed. The correlation of Y=1.0514X+0.004 between internal circulation flow rate(Y) and biogas production(X) was obtained by fitting.

  3. PHASS99: A software program for retrieving and decoding the radiometric ages of igneous rocks from the international database IGBADAT

    Science.gov (United States)

    Al-Mishwat, Ali T.

    2016-05-01

    PHASS99 is a FORTRAN program designed to retrieve and decode radiometric and other physical age information of igneous rocks contained in the international database IGBADAT (Igneous Base Data File). In the database, ages are stored in a proprietary format using mnemonic representations. The program can handle up to 99 ages in an igneous rock specimen and caters to forty radiometric age systems. The radiometric age alphanumeric strings assigned to each specimen description in the database consist of four components: the numeric age and its exponential modifier, a four-character mnemonic method identification, a two-character mnemonic name of analysed material, and the reference number in the rock group bibliography vector. For each specimen, the program searches for radiometric age strings, extracts them, parses them, decodes the different age components, and converts them to high-level English equivalents. IGBADAT and similarly-structured files are used for input. The output includes three files: a flat raw ASCII text file containing retrieved radiometric age information, a generic spreadsheet-compatible file for data import to spreadsheets, and an error file. PHASS99 builds on the old program TSTPHA (Test Physical Age) decoder program and expands greatly its capabilities. PHASS99 is simple, user friendly, fast, efficient, and does not require users to have knowledge of programing.

  4. A Comprehensive Study on the Degradation of Lithium-Ion Batteries during Calendar Ageing: The Internal Resistance Increase

    DEFF Research Database (Denmark)

    Stroe, Daniel Loan; Swierczynski, Maciej Jozef; Kær, Søren Knudsen

    2016-01-01

    Lithium-ion batteries are regarded as the key energy storage technology for both e-mobility and stationary renewable energy storage applications. Nevertheless, the Lithium-ion batteries are complex energy storage devices, which are characterized by a complex degradation behavior, which affects both...... their capacity and internal resistance. This paper investigates, based on extended laboratory calendar ageing tests, the degradation of the internal resistance of a Lithium-ion battery. The dependence of the internal resistance increase on the temperature and state-of-charge level have been extensive studied...... and quantified. Based on the obtained laboratory results, an accurate semi-empirical lifetime model, which is able to predict with high accuracy the internal resistance increase of the Lithium-ion battery over a wide temperature range and for all state-of-charge levels was proposed and validated....

  5. Cortisol Awakening Response and Internalizing Symptoms across Childhood: Exploring the Role of Age and Externalizing Symptoms

    Science.gov (United States)

    McGinnis, Ellen W.; Lopez-Duran, Nestor; Martinez-Torteya, Cecilia; Abelson, James L.; Muzik, Maria

    2016-01-01

    Efforts to identify biological correlates of internalizing symptoms in childhood have involved examinations of HPA-axis functioning, namely Cortisol Awakening Response (CAR). However, research has not assessed the relationship between CAR and internalizing problems among children younger than 8 years. Findings with older samples have been somewhat…

  6. Form of the male and female corpus callosum internal organization at the mature age

    Directory of Open Access Journals (Sweden)

    Юрий Петрович Костиленко

    2016-04-01

    Full Text Available Aim: to study the special features of the male and female corpus callosum internal organization at the mature age.Materials and methods: the total preparations of the male and female corpus callosum (10 preparation of each sex at 45–60 years old were used as the material. The given preparations were used to get from it the plate cuts in the two mutually perpendicular planes with 2 mm. thick. Then the received tissue plates of the corpus callosum underwent plastination in the epoxy. Then the preparations were extracted from the non-polymerized epoxy and placed on the polyethylene film that was covered with the other film of the same size. Further this stratified block was placed amid the two glasses of the equal size that shrunk together by placing the small load on it. After the complete polymerization the received epoxy plates with the corpus callosum tissue contained in it underwent the gentle grinding and the accurate polish and as the result was obtained the surface denudation of its tissue structures that were colored with the 1 % solution of blue methylene for 1% borax solution.Results of research: at the study of the corpus callosum plastinated cuts in saggital plane was revealed that the transverse platen-form elevations of its higher surface are the cord-form tenias standing out from within and going through the corpus callosum. At its studying in the transverse cut was established that in adults can be separated two types of corpus callosum by its density: the dense one and disperse one.At the large increases of the binocular loupe (microscope MBS-9 can be seen the gaps between the adjacent commissural cords. Within it can be detected the blood vessels. On the transverse cut of commissural cords in its depth are revealed the thinnest streaks which totality consists of the two alternate dark and light lines that form the layered striation. Among the series of the light lines are visible the interlayer that separate the whole depth of

  7. Poverty-Related Adversity and Emotion Regulation Predict Internalizing Behavior Problems among Low-Income Children Ages 8-11.

    Science.gov (United States)

    Raver, C Cybele; Roy, Amanda L; Pressler, Emily; Ursache, Alexandra M; Charles McCoy, Dana

    2016-12-29

    The current study examines the additive and joint roles of chronic poverty-related adversity and three candidate neurocognitive processes of emotion regulation (ER)-including: (i) attention bias to threat (ABT); (ii) accuracy of facial emotion appraisal (FEA); and (iii) negative affect (NA)-for low-income, ethnic minority children's internalizing problems (N = 338). Children were enrolled in the current study from publicly funded preschools, with poverty-related adversity assessed at multiple time points from early to middle childhood. Field-based administration of neurocognitively-informed assessments of ABT, FEA and NA as well as parental report of internalizing symptoms were collected when children were ages 8-11, 6 years after baseline. Results suggest that chronic exposure to poverty-related adversity from early to middle childhood predicted higher levels of internalizing symptomatology when children are ages 8-11, even after controlling for initial poverty status and early internalizing symptoms in preschool. Moreover, each of the 3 hypothesized components of ER played an independent and statistically significant role in predicting children's parent-reported internalizing symptoms at the 6-year follow-up, even after controlling for early and chronic poverty-related adversity.

  8. Poverty-Related Adversity and Emotion Regulation Predict Internalizing Behavior Problems among Low-Income Children Ages 8–11

    Science.gov (United States)

    Raver, C. Cybele; Roy, Amanda L.; Pressler, Emily; Ursache, Alexandra M.; Charles McCoy, Dana

    2016-01-01

    The current study examines the additive and joint roles of chronic poverty-related adversity and three candidate neurocognitive processes of emotion regulation (ER)—including: (i) attention bias to threat (ABT); (ii) accuracy of facial emotion appraisal (FEA); and (iii) negative affect (NA)—for low-income, ethnic minority children’s internalizing problems (N = 338). Children were enrolled in the current study from publicly funded preschools, with poverty-related adversity assessed at multiple time points from early to middle childhood. Field-based administration of neurocognitively-informed assessments of ABT, FEA and NA as well as parental report of internalizing symptoms were collected when children were ages 8–11, 6 years after baseline. Results suggest that chronic exposure to poverty-related adversity from early to middle childhood predicted higher levels of internalizing symptomatology when children are ages 8–11, even after controlling for initial poverty status and early internalizing symptoms in preschool. Moreover, each of the 3 hypothesized components of ER played an independent and statistically significant role in predicting children’s parent-reported internalizing symptoms at the 6-year follow-up, even after controlling for early and chronic poverty-related adversity. PMID:28036091

  9. Reactor vessel

    OpenAIRE

    Makkee, M.; Kapteijn, F.; Moulijn, J.A

    1999-01-01

    A reactor vessel (1) comprises a reactor body (2) through which channels (3) are provided whose surface comprises longitudinal inwardly directed parts (4) and is provided with a catalyst (6), as well as buffer bodies (8, 12) connected to the channels (3) on both sides of the reactor body (2) and comprising connections for supplying (9, 10, 11) and discharging (13, 14, 15) via the channels (3) gases and/or liquids entering into a reaction with each other and substances formed upon this reactio...

  10. Chemical Reactors.

    Science.gov (United States)

    Kenney, C. N.

    1980-01-01

    Describes a course, including content, reading list, and presentation on chemical reactors at Cambridge University, England. A brief comparison of chemical engineering education between the United States and England is also given. (JN)

  11. Reactor Neutrinos

    Directory of Open Access Journals (Sweden)

    Soo-Bong Kim

    2013-01-01

    Full Text Available We review the status and the results of reactor neutrino experiments. Short-baseline experiments have provided the measurement of the reactor neutrino spectrum, and their interest has been recently revived by the discovery of the reactor antineutrino anomaly, a discrepancy between the reactor neutrino flux state of the art prediction and the measurements at baselines shorter than one kilometer. Middle and long-baseline oscillation experiments at Daya Bay, Double Chooz, and RENO provided very recently the most precise determination of the neutrino mixing angle θ13. This paper provides an overview of the upcoming experiments and of the projects under development, including the determination of the neutrino mass hierarchy and the possible use of neutrinos for society, for nonproliferation of nuclear materials, and geophysics.

  12. NUCLEAR REACTOR

    Science.gov (United States)

    Miller, H.I.; Smith, R.C.

    1958-01-21

    This patent relates to nuclear reactors of the type which use a liquid fuel, such as a solution of uranyl sulfate in ordinary water which acts as the moderator. The reactor is comprised of a spherical vessel having a diameter of about 12 inches substantially surrounded by a reflector of beryllium oxide. Conventionnl control rods and safety rods are operated in slots in the reflector outside the vessel to control the operation of the reactor. An additional means for increasing the safety factor of the reactor by raising the ratio of delayed neutrons to prompt neutrons, is provided and consists of a soluble sulfate salt of beryllium dissolved in the liquid fuel in the proper proportion to obtain the result desired.

  13. Reactor Engineering

    Science.gov (United States)

    Lema, Juan M.; López, Carmen; Eibes, Gemma; Taboada-Puig, Roberto; Moreira, M. Teresa; Feijoo, Gumersindo

    In this chapter, the engineering aspects of processes catalyzed by peroxidases will be presented. In particular, a discussion of the existing technologies that utilize peroxidases for different purposes, such as the removal of recalcitrant compounds or the synthesis of polymers, is analyzed. In the first section, the essential variables controlling the process will be investigated, not only those that are common in any enzymatic system but also those specific to peroxidative reactions. Next, different reactor configurations and operational modes will be proposed, emphasizing their suitability and unsuitability for different systems. Finally, two specific reactors will be described in detail: enzymatic membrane reactors and biphasic reactors. These configurations are especially valuable for the treatment of xenobiotics with high and poor water solubility, respectively.

  14. Bio-terrorism, human security and public health: can international law bring them together in an age of globalization?

    Science.gov (United States)

    Aginam, Obijiofor

    2005-09-01

    Bio-terrorism, the use of a microorganism with the deliberate intent of causing infection, before and since the anthrax attacks in the United States in October 2001, has emerged as a real medical and public health threat. The link between bio-terrorism, human security and public health raises complex questions on the normative trajectories of international law, the mandates of international organizations, and global health governance. In May 2001, the World Health Assembly of the World Health Organization (WHO) passed a resolution entitled "Global Health Security: Epidemic Alert and Response" which inter alia, urged WHO member states to participate actively in the verification and validation of surveillance data and information concerning health emergencies of international concern. This article explores the links between bio-terrorism, human security and public health, and investigates the effectiveness of international legal mechanisms that link them in an age of globalization of public health. The article explores the interaction of WHO's 'soft-law' approaches to global health security, and the 'moribund' negotiations of the verification and monitoring protocol to the Biological Weapons Convention 1972. Can international law link bio-terrorism, public health and human security? Does the WHO collaborate with other international organizations within and outside the United Nations system to develop effective legal and governance approaches to bio-terrorism and global health security? The article concludes that the globalization of public health threats like bio-terrorism requires globalized legal approaches.

  15. Nuclear Reactors and Technology

    Energy Technology Data Exchange (ETDEWEB)

    Cason, D.L.; Hicks, S.C. [eds.

    1992-01-01

    This publication Nuclear Reactors and Technology (NRT) announces on a monthly basis the current worldwide information available from the open literature on nuclear reactors and technology, including all aspects of power reactors, components and accessories, fuel elements, control systems, and materials. This publication contains the abstracts of DOE reports, journal articles, conference papers, patents, theses, and monographs added to the Energy Science and Technology Database during the past month. Also included are US information obtained through acquisition programs or interagency agreements and international information obtained through the International Energy Agency`s Energy Technology Data Exchange or government-to-government agreements. The digests in NRT and other citations to information on nuclear reactors back to 1948 are available for online searching and retrieval on the Energy Science and Technology Database and Nuclear Science Abstracts (NSA) database. Current information, added daily to the Energy Science and Technology Database, is available to DOE and its contractors through the DOE Integrated Technical Information System. Customized profiles can be developed to provide current information to meet each user`s needs.

  16. Reactor Neutrinos

    OpenAIRE

    Lasserre, T.; Sobel, H.W.

    2005-01-01

    We review the status and the results of reactor neutrino experiments, that toe the cutting edge of neutrino research. Short baseline experiments have provided the measurement of the reactor neutrino spectrum, and are still searching for important phenomena such as the neutrino magnetic moment. They could open the door to the measurement of coherent neutrino scattering in a near future. Middle and long baseline oscillation experiments at Chooz and KamLAND have played a relevant role in neutrin...

  17. Height, weight and BMI percentiles and nutritional status relative to the international growth references among Pakistani school-aged children

    Directory of Open Access Journals (Sweden)

    Mushtaq Muhammad Umair

    2012-03-01

    Full Text Available Abstract Background Child growth is internationally recognized as an important indicator of nutritional status and health in populations. This study was aimed to compare age- and gender-specific height, weight and BMI percentiles and nutritional status relative to the international growth references among Pakistani school-aged children. Methods A population-based study was conducted with a multistage cluster sample of 1860 children aged five to twelve years in Lahore, Pakistan. Smoothed height, weight and BMI percentile curves were obtained and comparison was made with the World Health Organization 2007 (WHO and United States' Centers for Disease Control and Prevention 2000 (USCDC references. Over- and under-nutrition were defined according to the WHO and USCDC references, and the International Obesity Task Force (IOTF cut-offs. Simple descriptive statistics were used and statistical significance was considered at P Results Height, weight and BMI percentiles increased with age among both boys and girls, and both had approximately the same height and a lower weight and BMI as compared to the WHO and USCDC references. Mean differences from zero for height-, weight- and BMI-for-age z score values relative to the WHO and USCDC references were significant (P Conclusion Pakistani school-aged children significantly differed from the WHO and USCDC references. However, z score means relative to the WHO reference were closer to zero and the present study as compared to the USCDC reference. Overweight and obesity were significantly higher while underweight and thinness/wasting were significantly lower relative to the WHO reference as compared to the USCDC reference and the IOTF cut-offs. New growth charts for Pakistani children based on a nationally representative sample should be developed. Nevertheless, shifting to use of the 2007 WHO child growth reference might have important implications for child health programs and primary care pediatric clinics.

  18. Metallic fuels for advanced reactors

    Science.gov (United States)

    Carmack, W. J.; Porter, D. L.; Chang, Y. I.; Hayes, S. L.; Meyer, M. K.; Burkes, D. E.; Lee, C. B.; Mizuno, T.; Delage, F.; Somers, J.

    2009-07-01

    In the framework of the Generation IV Sodium Fast Reactor Program, the Advanced Fuel Project has conducted an evaluation of the available fuel systems supporting future sodium cooled fast reactors. This paper presents an evaluation of metallic alloy fuels. Early US fast reactor developers originally favored metal alloy fuel due to its high fissile density and compatibility with sodium. The goal of fast reactor fuel development programs is to develop and qualify a nuclear fuel system that performs all of the functions of a conventional fast spectrum nuclear fuel while destroying recycled actinides. This will provide a mechanism for closure of the nuclear fuel cycle. Metal fuels are candidates for this application, based on documented performance of metallic fast reactor fuels and the early results of tests currently being conducted in US and international transmutation fuel development programs.

  19. Internal morphology of human facet joints: comparing cervical and lumbar spine with regard to age, gender and the vertebral core.

    Science.gov (United States)

    Wilke, Hans-Joachim; Zanker, Daniel; Wolfram, Uwe

    2012-03-01

    Back pain constitutes a major problem in modern societies. Facet joints are increasingly recognised as a source of such pain. Knowledge about the internal morphology and its changes with age may make it possible to include the facets more in therapeutic strategies, for instance joint replacements or immobilisation. In total, 168 facets from C6/7 and L4/5 segments were scanned in a micro-computed tomography. Image analysis was used to investigate the internal morphology with regard to donor age and gender. Additional data from trabecular bone of the vertebral core allowed a semi-quantitative comparison of the morphology of the vertebral core and the facets. Porosity and pore spacing of the cortical sub-chondral bone does not appear to change with age for either males or females. In contrast, bone volume fraction decreases in females from approximately 0.4 to 0.2 , whereas it is constant in males. Trabecular thickness decreases during the ageing process in females and stays constant in males , whereas trabecular separation increases during the ageing process in both genders. The results of this study may help to improve the understanding of pathophysiological changes in the facet joints. Such results could be of value for understanding back pain and its treatment.

  20. Caloric restriction increases internal iliac artery and penil nitric oxide synthase expression in rat: Comparison of aged and adult rats

    Directory of Open Access Journals (Sweden)

    Emin Ozbek

    2013-09-01

    Full Text Available Because of the positive corelation between healthy cardiovascular system and sexual life we aimed to evaluate the effect of caloric restriction (CR on endothelial and neuronal nitric oxide synthase (eNOS, nNOS expression in cavernousal tissues and eNOS expression in the internal iliac artery in young and aged rats. Young (3 mo, n = 7 and aged (24 mo, n = 7 male Sprague-Dawley rats were subjected to 40% CR and were allowed free access to water for 3 months. Control rats (n = 14 fed ad libitum had free access to food and water at all times. On day 90, rats were sacrified and internal iliac arteries and penis were removed and parafinized, eNOS and nNOS expression evaluated with immunohistochemistry. Results were evaluated semiquantitatively. eNOS and nNOS expression in cavernousal tis- sue in CR rats were more strong than in control group in both young and old rats. eNOS expression was also higher in the internal iliac arteries of CR rats than in control in young and old rats. As a result of our study we can say that there is a positive link between CR and neurotransmitter of erection in cavernousal tissues and internal iliac arteries. CR has beneficial effect to prevent sexual dysfunction in young and old animals and possible humans.

  1. Reactor monitoring and safeguards using antineutrino detectors

    CERN Document Server

    Bowden, N S

    2008-01-01

    Nuclear reactors have served as the antineutrino source for many fundamental physics experiments. The techniques developed by these experiments make it possible to use these very weakly interacting particles for a practical purpose. The large flux of antineutrinos that leaves a reactor carries information about two quantities of interest for safeguards: the reactor power and fissile inventory. Measurements made with antineutrino detectors could therefore offer an alternative means for verifying the power history and fissile inventory of a reactors, as part of International Atomic Energy Agency (IAEA) and other reactor safeguards regimes. Several efforts to develop this monitoring technique are underway across the globe.

  2. Decontamination of the reactor pressure vessel and further internals and auxiliary systems in the German boiling water reactor Isar-1; Dekontamination des RDB inkl. der Einbauten wie Dampftrockner und Wasserabscheider sowie der angeschlossenen Hilfssysteme im deutschen Siedewasserreaktor ISAR 1

    Energy Technology Data Exchange (ETDEWEB)

    Fischer, Michael; Sempere Belda, Luis; Basu, Ashim; Topf, Christian [AREVA GmbH, Erlangen (Germany). Abt. Chemistry Services; Erbacher, Thomas; Hiermer, Thomas; Schnurr, Bernhard; Appeldorn, Thomas van [E.ON Kernkraft GmbH, Kernkraftwerk ISAR, Essenbach (Germany). Abt. Maschinentechnik; Volkmann, Christian [ESG Engineering Services GmbH, Greifswald (Germany)

    2015-12-15

    The German nuclear power plant ISAR 1 (KKI 1), a 878 MWe boiling water reactor of KWU design, was shut down on March 17{sup th}, 2011. With the objective to minimize the plants activity inventory accompanied by the reduction of contact dose rates of systems and components the project 'decontamination of the RPV incl. steam dryer and water separator and the connected auxiliary systems' was implemented in the first quarter of 2015. One major focus within the project was the specific in-situ decontamination of the steam dryer.

  3. Concept for LEU Burst Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Klein, Steven Karl [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Kimpland, Robert Herbert [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-03-07

    Design and performance of a proposed LEU burst reactor are sketched. Salient conclusions reached are the following: size would be ~1,500 kg or greater, depending on the size of the central cavity; internal stresses during burst require split rings for relief; the reactor would likely require multiple control and safety rods for fine control; the energy spectrum would be comparable to that of HEU machines; and burst yields and steady-state power levels will be significantly greater in an LEU reactor.

  4. An internal report: Electron Spectroscopy of the Oxidation and Aging of U and Pu

    Energy Technology Data Exchange (ETDEWEB)

    Tobin, J. G. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2015-04-27

    Uranium and Plutonium are highly reactive elements that undergo not only chemical reactions but also nuclear reactions. This can lead to possibly significant materials degradation, a matter of potentially great concern. Here, the issue of the electronic structure changes that occur with oxidation and radiological aging will be addressed, in a fairly empirical manner. In essence, the sensitivity of various electron spectroscopic techniques to oxidation and aging will be surveyed and discussed, including the apparent limitations. It will be found that 5d and 4d X-ray absorption and electron energy loss are essentially blind to the changes corresponding to oxidation and aging in U and Pu.

  5. Flow mixing model of liquid phase in an internal airlift loop reactor%气升式内环流反应器的液相流动混合模型

    Institute of Scientific and Technical Information of China (English)

    聂大仕; 沈文豪; 刘群姣; 张震宇; 王欣

    2006-01-01

    Residence time distribution (RTD) analysis of liquid phase was conducted in an internal airlift loop reactor (AL) and a bubble column (BC) with the tracer response technique. These data were simulated and compared through several flow mixing models. The modeling results of two-parameter model indicated that there were higher ratio of full mixing zones and lower ratio of bypass flow in AL than in BC. Then a completely mixed-plug flow parallel combined (four-parameter) model was established. Modeling results show that it is more precise and more obvious than two-parameter model.

  6. The XIIIth International Physiological Congress in Boston in 1929: American Physiology Comes of Age

    Science.gov (United States)

    Rall, Jack A.

    2016-01-01

    In the 19th century, the concept of experimental physiology originated in France with Claude Bernard, evolved in Germany stimulated by the teaching of Carl Ludwig, and later spread to Britain and then to the United States. The goal was to develop a physicochemical understanding of physiological phenomena. The first International Physiological…

  7. SBR短程硝化处理老龄化垃圾渗滤液%Treatment of Aged Landfill Leachate by Shortcut Nitrification in Sequencing Batch Reactor

    Institute of Scientific and Technical Information of China (English)

    周海妙; 解庆林; 黄国玲; 杨永东

    2012-01-01

    采用序批式反应器(SBR)短程硝化系统处理老龄化垃圾渗滤液,研究有机物浓度、水力停留时间(HRT)、pH值、温度对短程硝化系统的影响.以硝化污泥接种反应器,在溶解氧为1.0~1.2 mg/L和温度为(35±1)℃下达到亚硝酸氮的快速积累.结果表明,在进水氨氮为300mg/L、COD为600 mg/L、HRT为24 h、pH值为7.5~8.5、温度为(35±1)℃、溶解氧浓度保持不变的条件下,出水氨氮平均为134.0 mg/L,出水亚硝酸氮平均为142.5 mg/L,对氨氮的平均去除率为55.3%,NO2--N/NH4+-N平均值为1.06,出水硝酸氮平均为10.2 mg/L,亚硝酸氮的平均积累率为93.3%,对COD的去除率稳定在38%左右.%The shortcut nitrification process was used to treat aged landfill leachate in a sequencing batch reactor (SBR). The effect of organic concentration, hydraulic retention time (HRT) , pH and temperature was investigated. The reactor was inoculated with nitrifying sludge, and nitrite accumulation was achieved at DO concentration of 1. 0 to 1. 2 mg/L and temperature of (35 ± 1) ℃. The results showed that the average concentrations of NH4+ - N, NO2- - N and NO3- - N in effluent were 134. 0 mg/L, 142.5 mg/L and 10.2 mg/L, the average NO2- - N/NH4+ -N ratio was 1.06, the average nitrite accumulation was 93. 3% , and the average removal rates of NH4+ - N and COD were 55. 3% and about 38% respectively when the influent NH4+ - N and COD were 300 mg/L and 600 mg/L respectively. The reaction conditions were controlled as follows: HRT at 24 h, pH at 7. 5 to 8. 5 , temperature at (35 ± 1) ℃ and unchanged DO concentration.

  8. Bioconversion reactor

    Science.gov (United States)

    McCarty, Perry L.; Bachmann, Andre

    1992-01-01

    A bioconversion reactor for the anaerobic fermentation of organic material. The bioconversion reactor comprises a shell enclosing a predetermined volume, an inlet port through which a liquid stream containing organic materials enters the shell, and an outlet port through which the stream exits the shell. A series of vertical and spaced-apart baffles are positioned within the shell to force the stream to flow under and over them as it passes from the inlet to the outlet port. The baffles present a barrier to the microorganisms within the shell causing them to rise and fall within the reactor but to move horizontally at a very slow rate. Treatment detention times of one day or less are possible.

  9. Beyond The InternationalAccounting Standards Towards a Global Age of Accounting

    Directory of Open Access Journals (Sweden)

    TATIANA DANESCU

    2011-09-01

    Full Text Available Since the birth of the European Union the way people make business has changed. Globalization is no longer a key concept in economic development, but it has become a necessity. Accounting provides useful information to decision makers in companies and large corporations, therefore it has to keep pace with the International Accounting Standards. The European countries members of the EU and other countries are still in the process of implementing these standards.

  10. Age-specific risk factor profiles of adenocarcinomas of the esophagus: A pooled analysis from the international BEACON consortium.

    Science.gov (United States)

    Drahos, Jennifer; Xiao, Qian; Risch, Harvey A; Freedman, Neal D; Abnet, Christian C; Anderson, Lesley A; Bernstein, Leslie; Brown, Linda; Chow, Wong-Ho; Gammon, Marilie D; Kamangar, Farin; Liao, Linda M; Murray, Liam J; Ward, Mary H; Ye, Weimin; Wu, Anna H; Vaughan, Thomas L; Whiteman, David C; Cook, Michael B

    2016-01-01

    Esophageal (EA) and esophagogastric junction (EGJA) adenocarcinoma have been steadily increasing in frequency in younger people; however, the etiology of these cancers is poorly understood. We therefore investigated associations of body mass index (BMI), cigarette smoking, alcohol consumption, gastroesophageal reflux and use of nonsteroidal anti-inflammatory drugs (NSAIDs) in relation to age-specific risks of EA and EGJA. We pooled individual participant data from eight population-based, case-control studies within the international Barrett's and Esophageal Adenocarcinoma Consortium (BEACON). The analysis included 1,363 EA patients, 1,472 EGJA patients and 5,728 control participants. Multivariable logistic regression was used to estimate odds ratios (ORs) and 95% confidence intervals (CIs) for age-specific (<50, 50-59, 60-69, ≥70 years) cancer outcomes, as well as interactions by age. BMI, smoking status and pack-years, recurrent gastroesophageal reflux and frequency of gastroesophageal reflux were positively associated with EA and EGJA in each age group. Early-onset EA (<50 years) had stronger associations with recurrent gastroesophageal reflux (OR = 8.06, 95% CI: 4.52, 14.37; peffect modification  = 0.01) and BMI (ORBMI ≥ 30 vs . <25  = 4.19, 95% CI: 2.23, 7.87; peffect modification  = 0.04), relative to older age groups. In contrast, inverse associations of NSAID use were strongest in the oldest age group (≥70 years), although this apparent difference was not statistically significant. Age-specific associations with EGJA showed similar, but slightly weaker patterns and no statistically significant differences by age were observed. Our study provides evidence that associations between obesity and gastroesophageal reflux are stronger among earlier onset EA cancers.

  11. Nutrition and the biology of human ageing: Proceedings of the ninth nestle international nutrition symposium

    Science.gov (United States)

    This 9th Nestle Nutrition Symposium on “Nutrition and the Biology of Human Ageing” is presented at a time of unprecedented demographic change worldwide. The UN population division forecasts that the number of people living over age 65 will rise to almost 1 billion (12% percent of the world’s populat...

  12. School-Age Children Obesity in Rural Communities with International Migration in Central Mexico.

    OpenAIRE

    Carmona González, Maricela; Viczcarra Bordi., Ivonne

    2009-01-01

    The objective of the study was to evaluate the prevalence of obesity in school children between 6 and 12 years old. The study was carried out in communities with international migration tradition in the south of the Mexico state. Changes on food habits and life style were also evaluated. The Index of Corporal Mass (ICM) of 276 students was recorded. The 276 students were divided into two groups according to migrant or non-migrant father and by gender to identify differences between groups. Pa...

  13. Sonochemical Reactors.

    Science.gov (United States)

    Gogate, Parag R; Patil, Pankaj N

    2016-10-01

    Sonochemical reactors are based on the generation of cavitational events using ultrasound and offer immense potential for the intensification of physical and chemical processing applications. The present work presents a critical analysis of the underlying mechanisms for intensification, available reactor configurations and overview of the different applications exploited successfully, though mostly at laboratory scales. Guidelines have also been presented for optimum selection of the important operating parameters (frequency and intensity of irradiation, temperature and liquid physicochemical properties) as well as the geometric parameters (type of reactor configuration and the number/position of the transducers) so as to maximize the process intensification benefits. The key areas for future work so as to transform the successful technique at laboratory/pilot scale into commercial technology have also been discussed. Overall, it has been established that there is immense potential for sonochemical reactors for process intensification leading to greener processing and economic benefits. Combined efforts from a wide range of disciplines such as material science, physics, chemistry and chemical engineers are required to harness the benefits at commercial scale operation.

  14. Old-Age Pension and Extended Families: How is Adult Children's Internal Migration Affected?

    Science.gov (United States)

    Chen, Xi

    2016-10-01

    This paper makes use of the most recent social pension reform in rural China to examine whether receipt of the pension payment equips adult children of pensioners to migrate. Employing a regression discontinuity (hereafter RD) design to a primary longitudinal survey, this paper overcomes challenges in the literature that households eligible for pension payment might be systematically different from ineligible households and that it is difficult to separate the effect of pension from that of age or cohort heterogeneity. Around the pension eligibility age cut-off, results reveal large and significant increase among adult sons (but not daughters) to migrate out of their home county. Meanwhile, adult children are more likely to migrate out if their parents are healthy. Our Fuzzy RD estimations survive a standard set of key placebo tests and robustness checks.

  15. JHR Project: a future Material Testing Reactor working as an International user Facility: The key-role of instrumentation in support to the development of modern experimental capacity

    Energy Technology Data Exchange (ETDEWEB)

    Bignan, G. [CEA, DEN, DER, JHR user Facility Interface Manager' , Cadarache, F-13108 St-Paul-Lez-Durance (France); Gonnier, C. [CEA, DEN, DER, SRJH Jules Horowitz Reactor Service, Cadarache, F-13108 St-Paul-Lez-Durance (France); Lyoussi, A.; Villard, J.F.; Destouches, C. [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-Lez-Durance (France); Chauvin, J.P. [CEA,DEN, DER, SPEX, Experimental Physics Service, Cadarache, F-13108 St-Paul-Lez-Durance (France); Maugard, B. [CEA, DEN, DER, Reactor Department Studies, Cadarache, F-13108 St-Paul-Lez-Durance (France)

    2015-07-01

    Research and development on fuel and material behaviour under irradiation is a key issue for sustainable nuclear energy in order to meet specific needs by keeping the best level of safety. These needs mainly deal with a constant improvement of performances and safety in order to optimize the fuel cycle and hence to reach nuclear energy sustainable objectives. A sustainable nuclear energy requires a high level of performances in order to meet specific needs such as: - Pursuing improvement of the performances and safety of present and coming water cooled reactor technologies. This will require a continuous R and D support following a long-term trend driven by the plant life management, safety demonstration, flexibility and economics improvement. Experimental irradiations of structure materials are necessary to anticipate these material behaviours and will contribute to their optimisation. - Upgrading continuously nuclear fuel technology in present and future nuclear power plants to achieve better performances and to optimise the fuel cycle keeping the best level of safety. Fuel evolution for generation II, III and III+ is a key stake requiring developments, qualification tests and safety experiments to ensure the competitiveness and safety: experimental tests exploring the full range of fuel behaviour determine fuel stability limits and safety margins, as a major input for the fuel reliability analysis. To perform such accurate and innovative progress and developments, specific and ad hoc instrumentation, irradiation devices, measurement methods are necessary to be set up inside or beside the material testing reactor (MTR) core. These experiments require beforehand in situ and on line sophisticated measurements to accurately determine different key parameters such as thermal and fast neutron fluxes and nuclear heating in order to precisely monitor and control the conducted assays. The new Material Testing Reactor JHR (Jules Horowitz Reactor) currently under

  16. [Gerontosocial work in the context of Russian culture and recommendations of Madrid International Plan of Action on Ageing].

    Science.gov (United States)

    Danilova, R I; Golubeva, E Iu

    2007-01-01

    The definition of gerontosocial work adopted in the majority of the European countries in the context of the Russian culture and priority trends of the Madrid International Plan of Action on Ageing has been discussed. In the article the interaction and differences between terms being used in social work with the elderly people in Russia have been showed; tasks and peculiarities of gerontosocial work have been defined. Peculiarities of interaction between personnel and elderly patients have been studied; the increasing requirements to the professionalism of gerontosocial workers have been proved.

  17. International recognition for ageing research: John Scott Award-2014 to Leonard Hayflick and Paul Moorhead

    OpenAIRE

    Rattan, Suresh

    2014-01-01

    It is with great pleasure and pride that we share the news of the award of the 2014 “City of Philadelphia John Scott Award”, to Dr. Leonard Hayflick and Dr. Paul Moorhead, for their research on ageing. The press release announcing the award states that: “from the first awarded in 1822, the Award is the oldest scientific award in the United States and, as a legacy to Benjamin Franklin, they are in the historic company of past winners who include Marie Curie, Thomas Edison, Jonas Salk, Irving L...

  18. Review of the proposed materials of construction for the SBWR and AP600 advanced reactors

    Energy Technology Data Exchange (ETDEWEB)

    Diercks, D.R.; Shack, W.J.; Chung, H.M.; Kassner, T.F. [Argonne National Lab., IL (United States)

    1994-06-01

    Two advanced light water reactor (LWR) concepts, namely the General Electric Simplified Boiling Water Reactor (SBWR) and the Westinghouse Advanced Passive 600 MWe Reactor (AP600), were reviewed in detail by Argonne National Laboratory. The objectives of these reviews were to (a) evaluate proposed advanced-reactor designs and the materials of construction for the safety systems, (b) identify all aging and environmentally related degradation mechanisms for the materials of construction, and (c) evaluate from the safety viewpoint the suitability of the proposed materials for the design application. Safety-related systems selected for review for these two LWRs included (a) reactor pressure vessel, (b) control rod drive system and reactor internals, (c) coolant pressure boundary, (d) engineered safety systems, (e) steam generators (AP600 only), (f) turbines, and (g) fuel storage and handling system. In addition, the use of cobalt-based alloys in these plants was reviewed. The selected materials for both reactors were generally sound, and no major selection errors were found. It was apparent that considerable thought had been given to the materials selection process, making use of lessons learned from previous LWR experience. The review resulted in the suggestion of alternate an possibly better materials choices in a number of cases, and several potential problem areas have been cited.

  19. Integration of improved decontamination and characterization technologies in the decommissioning of the CP-5 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bhattacharyya, S. K.; Boing, L. E.

    2000-02-17

    The aging of research reactors worldwide has resulted in a heightened awareness in the international technical decommissioning community of the timeliness to review and address the needs of these research institutes in planning for and eventually performing the decommissioning of these facilities. By using the reactors already undergoing decommissioning as test beds for evaluating enhanced or new/innovative technologies for decommissioning, it is possible that new techniques could be made available for those future research reactor decommissioning projects. Potentially, the new technologies will result in: reduced radiation doses to the work force, larger safety margins in performing decommissioning and cost and schedule savings to the research institutes in performing the decommissioning of these facilities. Testing of these enhanced technologies for decontamination, dismantling, characterization, remote operations and worker protection are critical to furthering advancements in the technical specialty of decommissioning. Furthermore, regulatory acceptance and routine utilization for future research reactor decommissioning will be assured by testing and developing these technologies in realistically contaminated environments prior to use in the research reactors. The decommissioning of the CP-5 Research Reactor is currently in the final phase of dismantlement. In this paper the authors present results of work performed at Argonne National Laboratory (ANL) in the development, testing and deployment of innovative and/or enhanced technologies for the decommissioning of research reactors.

  20. Compact fusion reactors

    CERN Document Server

    CERN. Geneva

    2015-01-01

    Fusion research is currently to a large extent focused on tokamak (ITER) and inertial confinement (NIF) research. In addition to these large international or national efforts there are private companies performing fusion research using much smaller devices than ITER or NIF. The attempt to achieve fusion energy production through relatively small and compact devices compared to tokamaks decreases the costs and building time of the reactors and this has allowed some private companies to enter the field, like EMC2, General Fusion, Helion Energy, Lawrenceville Plasma Physics and Lockheed Martin. Some of these companies are trying to demonstrate net energy production within the next few years. If they are successful their next step is to attempt to commercialize their technology. In this presentation an overview of compact fusion reactor concepts is given.

  1. Hoarseness in School-Aged Children and Effectiveness of Voice Therapy in International Classification of Functioning Framework.

    Science.gov (United States)

    Akın Şenkal, Özgül; Özer, Cem

    2015-09-01

    The hoarseness in school-aged children disrupts the educational process because it affects the social progress, communication skills, and self-esteem of children. Besides otorhinolaryngological examination, the first treatment option is voice therapy when hoarseness occurs. The aim of the study was to determine the factors increasing the hoarseness in school-aged children by parental interview and to know preferable voice therapy on school-aged children within the frame of International Classification of Functioning (ICF). Retrospective analysis of data gathered from patient files. A total of 75 children (56 boys and 19 girls) were examined retrospectively. The age range of school-aged children is 7-14 years and average is 10.86 ± 2.51. A detailed history was taken from parents of children involved in this study. Information about vocal habits of children was gathered within the frame of ICF and then the voice therapies of children were started by scheduling appointments by an experienced speech-language pathologist. The differences between before and after voice therapy according to applied voice therapy methods, statistically significant differences were determined between maximum phonation time values and s/z rate. The relationship between voice therapy sessions and s/z rate with middle degree significance was found with physiological voice therapy sessions. According to ICF labels, most of voice complaints are matching with "body functions" and "activity and limitations." The appropriate voice therapy methods for hoarseness in school-aged children must be chosen and applied by speech-language therapists. The detailed history, which is received from family during the examination, within the frame of ICF affects the processes of choosing the voice therapy method and application of them positively. Child's family is very important for a successful management. Copyright © 2015 The Voice Foundation. Published by Elsevier Inc. All rights reserved.

  2. Proceedings of the fourth international conference on CANDU maintenance

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-07-01

    These proceedings record the information presented at the 4th International Conference on CANDU Maintenance held November 16-18,1997 in Toronto, Canada. The papers for these proceedings were prepared on component maintenance, human performance, steam generator leak detection, fuel channel inspections, rotating equipment maintenance, surveillance programs, inspection techniques, valve maintenance, steam generator repairs and performance, reactor aging management and preventative maintenance.

  3. Brazilian multipurpose reactor

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2014-07-01

    The Brazilian Multipurpose Reactor (RMB) Project is an action of the Federal Government, through the Ministry of Science Technology and Innovation (MCTI) and has its execution under the responsibility of the Brazilian National Nuclear Energy Commission (CNEN). Within the CNEN, the project is coordinated by the Research and Development Directorate (DPD) and developed through research units of this board: Institute of Nuclear Energy Research (IPEN); Nuclear Engineering Institute (IEN); Centre for Development of Nuclear Technology (CDTN); Regional Center of Nuclear Sciences (CRCN-NE); and Institute of Radiation Protection and Dosimetry (IRD). The Navy Technological Center in Sao Paulo (CTMSP) and also the participation of other research centers, universities, laboratories and companies in the nuclear sector are important and strategic partnerships. The conceptual design and the safety analysis of the reactor and main facilities, related to nuclear and environmental licensing, are performed by technicians of the research units of DPD / CNEN. The basic design was contracted to engineering companies as INTERTHECNE from Brazil and INVAP from Argentine. The research units from DPD/CNEN are also responsible for the design verification on all engineering documents developed by the contracted companies. The construction and installation should be performed by specific national companies and international partnerships. The Nuclear Reactor RMB will be a open pool type reactor with maximum power of 30 MW and have the OPAL nuclear reactor of 20 MW, built in Australia and designed by INVAP, as reference. The RMB reactor core will have a 5x5 configuration, consisting of 23 elements fuels (EC) of U{sub 3}Si{sub 2} dispersion-type Al having a density of up to 3.5 gU/cm{sup 3} and enrichment of 19.75% by weight of {sup 23{sup 5}}U. Two positions will be available in the core for materials irradiation devices. The main objectives of the RMB Reactor and the other nuclear and radioactive

  4. Proceedings of the International Association for Development of the Information Society (IADIS) International Conference on Cognition and Exploratory Learning in the Digital Age (CELDA) (Fort Worth, Texas, October 22-24, 2013)

    Science.gov (United States)

    Sampson, Demetrios G., Ed.; Spector, J. Michael, Ed.; Ifenthaler, Dirk, Ed.; Isaias, Pedro, Ed.

    2013-01-01

    These proceedings contain the papers of the IADIS International Conference on Cognition and Exploratory Learning in the Digital Age (CELDA 2013), October 22-24, 2013, which has been organized by the International Association for Development of the Information Society (IADIS), co-organized by The University of North Texas (UNT), sponsored by the…

  5. Proceedings of the International Association for Development of the Information Society (IADIS) International Conference on Cognition and Exploratory Learning in the Digital Age (CELDA) (13th, Mannheim, Germany, October 28-30, 2016)

    Science.gov (United States)

    Sampson, Demetrios G., Ed.; Spector, J. Michael, Ed.; Ifenthaler, Dirk, Ed.; Isaias, Pedro, Ed.

    2016-01-01

    These proceedings contain the papers of the 13th International Conference on Cognition and Exploratory Learning in the Digital Age (CELDA 2016), October 28-30, 2016, which has been organized by the International Association for Development of the Information Society (IADIS), co-organized by the University of Mannheim, Germany, and endorsed by the…

  6. Proceedings of the International Association for Development of the Information Society (IADIS) International Conference on Cognition and Exploratory Learning in the Digital Age (CELDA) (12th, Maynooth, Greater Dublin, Ireland, October 24-26, 2015)

    Science.gov (United States)

    Sampson, Demetrios G., Ed.; Spector, J. Michael, Ed.; Ifenthaler, Dirk, Ed.; Isaias, Pedro, Ed.

    2015-01-01

    These proceedings contain the papers of the 12th International Conference on Cognition and Exploratory Learning in the Digital Age (CELDA 2015), October 24-26, 2015, which has been organized by the International Association for Development of the Information Society (IADIS), co-organized by Maynooth University, Ireland, and endorsed by the…

  7. Setaria Comes of Age: Meeting Report on the Second International Setaria Genetics Conference

    Directory of Open Access Journals (Sweden)

    Chuanmei Zhu

    2017-09-01

    Full Text Available Setaria viridis is an emerging model for cereal and bioenergy grasses because of its short stature, rapid life cycle and expanding genetic and genomic toolkits. Its close phylogenetic relationship with economically important crops such as maize and sorghum positions Setaria as an ideal model system for accelerating discovery and characterization of crop genes that control agronomically important traits. The Second International Setaria Genetics Conference was held on March 6–8, 2017 at the Donald Danforth Plant Science Center, St. Louis, MO, United States to discuss recent technological breakthroughs and research directions in Setaria (presentation abstracts can be downloaded at https://www.brutnelllab.org/setaria. Here, we highlight topics presented in the conference including inflorescence architecture, C4 photosynthesis and abiotic stress. Genetic and genomic toolsets including germplasm, mutant populations, transformation and gene editing technologies are also discussed. Since the last meeting in 2014, the Setaria community has matured greatly in the quality of research being conducted. Outreach and increased communication with maize and other plant communities will allow broader adoption of Setaria as a model system to translate fundamental discovery research to crop improvement.

  8. Depression, Sex and Gender Roles in Older Adult Populations: The International Mobility in Aging Study (IMIAS).

    Science.gov (United States)

    Vafaei, Afshin; Ahmed, Tamer; Freire, Aline do N Falcão; Zunzunegui, Maria Victoria; Guerra, Ricardo O

    2016-01-01

    To assess the associations between gender roles and depression in older men and women and whether gender roles are independent risk factors for depression. International cross-sectional study of adults between 65 and 74 years old (n = 1,967). Depression was defined by a score of 16 or over in the Center for Epidemiologic Studies Depression Scale (CES-D). A validated 12-item Bem Sex Role Inventory (BSRI) was used to classify participants in gender roles (Masculine, Feminine, Androgynous, and Undifferentiated) using research site medians of femininity and masculinity as cut-off points. Poisson regressions were fitted to estimate the prevalence ratios (PR) of depression for each gender role compared to the masculine role, adjusting for sex, sufficiency of income, education, marital status, self-rated health, and chronic conditions. Among men, 31.2% were androgynous, 26% were masculine, 14.4% were feminine, and 28.4% were undifferentiated; among women, the corresponding percentages were 32.7%, 14.9%, 27%, and 25.4%. Both in men and in women, depressive symptoms (CES-D≥16) were more prevalent in those endorsing the undifferentiated type, compared to masculine, feminine or androgynous groups. However, after adjusting for potential confounders, compared to the masculine group only those endorsing the androgynous role were 28% less likely to suffer from depression: PR of 0.72 (95% CI: 0.55-0.93). In fully adjusted models, prevalence rates of depression were not different from masculine participants in the two other gender groups of feminine and undifferentiated. Androgynous roles were associated with lower rates of depression in older adults, independently of being a man or a woman.

  9. Depression, Sex and Gender Roles in Older Adult Populations: The International Mobility in Aging Study (IMIAS.

    Directory of Open Access Journals (Sweden)

    Afshin Vafaei

    Full Text Available To assess the associations between gender roles and depression in older men and women and whether gender roles are independent risk factors for depression.International cross-sectional study of adults between 65 and 74 years old (n = 1,967. Depression was defined by a score of 16 or over in the Center for Epidemiologic Studies Depression Scale (CES-D. A validated 12-item Bem Sex Role Inventory (BSRI was used to classify participants in gender roles (Masculine, Feminine, Androgynous, and Undifferentiated using research site medians of femininity and masculinity as cut-off points. Poisson regressions were fitted to estimate the prevalence ratios (PR of depression for each gender role compared to the masculine role, adjusting for sex, sufficiency of income, education, marital status, self-rated health, and chronic conditions.Among men, 31.2% were androgynous, 26% were masculine, 14.4% were feminine, and 28.4% were undifferentiated; among women, the corresponding percentages were 32.7%, 14.9%, 27%, and 25.4%. Both in men and in women, depressive symptoms (CES-D≥16 were more prevalent in those endorsing the undifferentiated type, compared to masculine, feminine or androgynous groups. However, after adjusting for potential confounders, compared to the masculine group only those endorsing the androgynous role were 28% less likely to suffer from depression: PR of 0.72 (95% CI: 0.55-0.93. In fully adjusted models, prevalence rates of depression were not different from masculine participants in the two other gender groups of feminine and undifferentiated.Androgynous roles were associated with lower rates of depression in older adults, independently of being a man or a woman.

  10. The conceptualization and organization of the first International Neurological Congress (1931): the coming of age of neurology.

    Science.gov (United States)

    Louis, Elan D

    2010-07-01

    The first International Neurological Congress (Berne, Switzerland, 1931), attended by individuals from 42 countries, signified a global presence of world neurology; a coming of age. The aim of this study was to trace the history of that Congress, an important episode in the emergence of our discipline. The historical literature was reviewed and a detailed study conducted of the Henry Alsop Riley Papers, Columbia University. These papers contain primary source material from the Berne conference. In 1927, two neurologists, Bernard Sachs (American, 1858-1944) and Otto Marburg (Austrian, 1874-1948) met at an Austrian spa town and began to consider the creation of a meeting with a truly international character. The Americans were to play a seminal role in the organization of the Congress. In 1928, an introductory letter from Sachs went out to the international community and, in 1929, a planning meeting was held and the general principles of the Congress were established. Several earlier attempted congresses had been thwarted by World War I and European tensions would also influence the organization of the Berne Congress. Gordon Holmes (1876-1965) wrote: 'It would be certainly wiser to have the meeting in Scandinavia, Holland or Switzerland, as the only difficulty may be to get the French and Germans to mix.' Interest in the congress was immense and subsequent international congresses (London, Copenhagen, Paris, Lisbon and Brussels) became a central event in world neurology. In summary, the Berne Congress brought together individuals from several continents, thereby facilitating the exchange of ideas across entire schools. More broadly, the congress reflected a growing trend toward specialization in neurology and medicine.

  11. MR image assessment of disc configuration and degree of anterior disc displacement in internal derangement related to age

    Energy Technology Data Exchange (ETDEWEB)

    Igarashi, Chinami; Kobayashi, Kaoru; Imanaka, Masahiro; Yuasa, Masao; Yamamoto, Akira [Tsurumi Univ., Yokohama (Japan). School of Dental Medicine

    1999-03-01

    This study was designed to evaluate the configuration of the articular disc and degree of anterior disc displacement on magnetic resonance (MR) imagings in temporomandibular joints (TMJs) with internal derangement. A total of 363 joints diagnosed as having anterior disc displacement with reduction (ADD w R) and 523 joints diagnosed as having anterior disc displacement without reduction (ADD wo R) by MR imaging were examined. These joints did not show severe osseous changes on the condylar head or glenoid fossa. We assessed the configuration of the articular disc and degree of anterior disc displacement. In the ADD w R group, 82.6% of the articular discs showed biconcave configuration; enlargement of the posterior band in 4.6%, biconvex configuration in 0.5%, and others in 10.7%. Moreover 62.5% of the discs showed a slight degree of anterior disc displacement; were 27.2% moderately displaced and were 10.2% severe displaced. The prevalence of slightly displaced discs was higher in the TMJs of cases over 50 years of age than in cases under 30 years in the ADD w R group. On the other hand, in the ADD wo R group 35.9% of the articular discs showed biconcave configuration; enlargement of the posterior band in 12.6%, biconvex configuration in 25.4%, and others in 22.3%. Furthermore, 4.4% of the discs were slightly displaced; 43.9% moderately displaced and 51.6% were severely displaced. The prevalence of severely displaced and deformed discs in joints of cases over 40 years of age was high in the ADD wo R group. The prevalence of slightly displaced biconcave discs was higher in the ADD w R group. The other hand, the prevalence of severely displaced deformed discs was higher in the ADD wo R group. MR findings of internal derangement of the TMJ were found to be significantly correlated with age. (author)

  12. The internal structure of foster-parent completed SDQ for school-aged children.

    Science.gov (United States)

    Lehmann, Stine; Bøe, Tormod; Breivik, Kyrre

    2017-01-01

    Mental health problems are common in foster-children, and tools to measure the mental health of these children are needed. One candidate instrument is the Strengths and Difficulties Questionnaire (SDQ), a measure of child psychological adjustment that is increasingly being employed by Child Protection services. The aim of the current study was to examine the structural validity of the foster parent completed SDQ in a sample of 237 school aged foster children. Confirmatory factor analysis demonstrated an excellent fit of the foster parent completed SDQ data to a five-factor model (CFI = 0.96, RMSEA = 0.05, 90% CI [0.04, 0.06]), thus confirming the structural validity of the five-factor model for the parent-version of the SDQ in Norwegian foster children. Measurement invariance analyses indicated that boys had lower thresholds for fighting with or bullying other children than girls. Girls were on their side more likely to be rated as less popular than boys with a similar level of peer problems.

  13. Proceedings of the International Conference on Cognition and Exploratory Learning in Digital Age (CELDA) (11th, Porto, Portugal, October 25-27, 2014)

    Science.gov (United States)

    Sampson, Demetrios G., Ed.; Spector, J. Michael, Ed.; Ifenthaler, Dirk, Ed.; Isaias, Pedro, Ed.

    2014-01-01

    These proceedings contain the papers of the 11th International Conference on Cognition and Exploratory Learning in the Digital Age (CELDA 2014), October 25-27, 2014, which has been organized by the International Association for Development of the Information Society (IADIS) and endorsed by the Japanese Society for Information and Systems in…

  14. Plant maintenance and advanced reactors, 2006

    Energy Technology Data Exchange (ETDEWEB)

    Agnihotri, Newal (ed.)

    2006-09-15

    The focus of the September-October issue is on plant maintenance and advanced reactors. Major articles/reports in this issue include: Advanced plants to meet rising expectations, by John Cleveland, International Atomic Energy Agency, Vienna; A flexible and economic small reactor, by Mario D. Carelli and Bojan Petrovic, Westinghouse Electric Company; A simple and passively safe reactor, by Yury N. Kuznetsov, Research and Development Institute of Power Engineering (NIKIET), Russia; Gas-cooled reactors, by Jeffrey S. Merrifield, U.S. Nuclear Regulatory Commission; ISI project managment in the PRC, by Chen Chanbing, RINPO, China; and, Fort Calhoun refurbishment, by Sudesh Cambhir, Omaha Public Power District.

  15. Osiris, an irradiation reactor for material and nuclear fuel testing; Osiris, reacteur d'irradiation pour materiaux et combustibles

    Energy Technology Data Exchange (ETDEWEB)

    Loubiere, S.; Durande-Ayme, P. [CEA Saclay, Div. Nucleaire Energie, Dept. Reacteurs et Nucleaire Service, 91 - Gif-sur-Yvette (France)

    2005-07-01

    Since 1966 the Osiris reactor located at Saclay has been participating in French and international irradiation programs for research and development in the field of nuclear fuel and materials. Today the French atomic commission (Cea) pursues irradiation programs in support of existing reactors, mainly PWR, strengthening its own knowledge and the one of its clients on fuel and material behaviour under irradiation, pertaining to plant life-time issues and high burn-up. For instance important programs have been performed on pressure vessel steel aging, pellet-clad interaction, internal component aging and mox fuel qualification. With the arising of the Generation 4 research and development programs, the Osiris reactor has developed capacities to undertake material and fuel irradiation under high temperature conditions. Routine irradiations such as the doping of silicon or the production of radio-nuclides for medical or imaging purposes are made on a daily basis. The specificities of the Osiris reactor are presented in the first part of this paper while the second part focuses on the experimental devices available in Osiris to perform irradiation in light water reactor conditions and in high temperature reactor conditions and on their associated programs.

  16. Integral reactor system and method for fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Fernandes, Neil Edward; Brown, Michael S.; Cheekatamaria, Praveen; Deng, Thomas; Dimitrakopoulos, James; Litka, Anthony F.

    2017-03-07

    A reactor system is integrated internally within an anode-side cavity of a fuel cell. The reactor system is configured to convert higher hydrocarbons to smaller species while mitigating the lower production of solid carbon. The reactor system may incorporate one or more of a pre-reforming section, an anode exhaust gas recirculation device, and a reforming section.

  17. Integral reactor system and method for fuel cells

    Science.gov (United States)

    Fernandes, Neil Edward; Brown, Michael S; Cheekatamarla, Praveen; Deng, Thomas; Dimitrakopoulos, James; Litka, Anthony F

    2013-11-19

    A reactor system is integrated internally within an anode-side cavity of a fuel cell. The reactor system is configured to convert hydrocarbons to smaller species while mitigating the lower production of solid carbon. The reactor system may incorporate one or more of a pre-reforming section, an anode exhaust gas recirculation device, and a reforming section.

  18. Proceedings of the NEACRP/IAEA Specialists meeting on the international comparison calculation of a large sodium-cooled fast breeder reactor at Argonne National Laboratory on February 7-9, 1978

    Energy Technology Data Exchange (ETDEWEB)

    LeSage, L.G.; McKnight, R.D.; Wade, D.C.; Freese, K.E.; Collins, P.J.

    1980-08-01

    The results of an international comparison calculation of a large (1250 MWe) LMFBR benchmark model are presented and discussed. Eight reactor configurations were calculated. Parameters included with the comparison were: eigenvalue, k/sub infinity/, neutron balance data, breeding reaction rate ratios, reactivity worths, central control rod worth, regional sodium void reactivity, core Doppler and effective delayed neutron fraction. Ten countries participated in the comparison, and sixteen solutions were contributed. The discussion focuses on the variation in parameter values, the degree of consistency among the various parameters and solutions, and the identification of unexpected results. The results are displayed and discussed both by individual participants and by groupings of participants (e.g., results from adjusted data sets versus non-adjusted data sets).

  19. A comparison of radioactive waste from first generation fusion reactors and fast fission reactors with actinide recycling

    Energy Technology Data Exchange (ETDEWEB)

    Koch, M.; Kazimi, M.S.

    1991-04-01

    Limitations of the fission fuel resources will presumably mandate the replacement of thermal fission reactors by fast fission reactors that operate on a self-sufficient closed fuel cycle. This replacement might take place within the next one hundred years, so the direct competitors of fusion reactors will be fission reactors of the latter rather than the former type. Also, fast fission reactors, in contrast to thermal fission reactors, have the potential for transmuting long-lived actinides into short-lived fission products. The associated reduction of the long-term activation of radioactive waste due to actinides makes the comparison of radioactive waste from fast fission reactors to that from fusion reactors more rewarding than the comparison of radioactive waste from thermal fission reactors to that from fusion reactors. Radioactive waste from an experimental and a commercial fast fission reactor and an experimental and a commercial fusion reactor has been characterized. The fast fission reactors chosen for this study were the Experimental Breeder Reactor 2 and the Integral Fast Reactor. The fusion reactors chosen for this study were the International Thermonuclear Experimental Reactor and a Reduced Activation Ferrite Helium Tokamak. The comparison of radioactive waste parameters shows that radioactive waste from the experimental fast fission reactor may be less hazardous than that from the experimental fusion reactor. Inclusion of the actinides would reverse this conclusion only in the long-term. Radioactive waste from the commercial fusion reactor may always be less hazardous than that from the commercial fast fission reactor, irrespective of the inclusion or exclusion of the actinides. The fusion waste would even be far less hazardous, if advanced structural materials, like silicon carbide or vanadium alloy, were employed.

  20. The OPAL reactor

    Energy Technology Data Exchange (ETDEWEB)

    Miller, R.; Irwin, T. [Australian Nuclear Science and Technology Organisation, Sydney (Australia); Ordonez, J.P. [INVAP SE, Bariloche (Argentina)

    2007-07-01

    The project to provide a replacement for Australia's HIFAR reactor began with governmental approval in September 1997 and reached its latest milestone with the achievement of the first full power operation of the OPAL reactor in November 2006. OPAL is a pool-type reactor with a thermal power of 20 MW and a fuel enrichment maximum of 20 per cent. This has been a successful project for both ANSTO (Australian Nuclear Science and Technology Organisation) and the contractor INVAP SE. This project was characterised by extensive interaction with the project's stake-holders during project definition and the use of a performance-based turnkey contract which gave the contractor the maximum opportunity to optimise the design to achieve performance and cost effectiveness. The contactor provided significant in-house resources as well as capacity to manage an international team of suppliers and sub-contractors. A key contributor to the project's successful outcomes has been the development and maintenance of an excellent working relationship between ANSTO and INVAP project teams. Commissioning was undertaken in accordance with the IAEA recommended stages. This paper presents the approaches used to define the project requirements, to choose the supplier and to deliver the project. The main results of hot commissioning are reviewed and the problems encountered examined. Operational experience since hot commissioning is also reviewed.

  1. Telomere Length in Aged Mayak PA Nuclear Workers Chronically Exposed to Internal Alpha and External Gamma Radiation.

    Science.gov (United States)

    Scherthan, Harry; Sotnik, Natalia; Peper, Michel; Schrock, Gerrit; Azizova, Tamara; Abend, Michael

    2016-06-01

    Telomeres consist of GC-rich DNA repeats and the "shelterin" protein complex that together protect chromosome ends from fusion and degradation. Telomeres shorten with age due to incomplete end replication and upon exposure to environmental and intrinsic stressors. Exposure to ionizing radiation is known to modulate telomere length. However, the response of telomere length in humans chronically exposed to radiation is poorly understood. Here, we studied relative telomere length (RTL) by IQ-FISH to leukocyte nuclei in a group of 100 workers from the plutonium production facility at the Mayak Production Association (PA) who were chronically exposed to alpha-emitting ((239)Pu) radiation and/or gamma (photon) radiation, and 51 local residents serving as controls, with a similar mean age of about 80 years. We applied generalized linear statistical models adjusted for age at biosampling and the second exposure type on a linear scale and observed an age-dependent telomere length reduction. In those individuals with the lowest exposure, a significant reduction of about 20% RTL was observed, both for external gamma radiation (≤1 Gy) and internal alpha radiation (≤0.05-0.1 Gy to the red bone marrow). In highly exposed individuals (>0.1 Gy alpha, 1-1.5 Gy gamma), the RTL was similar to control. Stratification by gender revealed a significant (∼30%) telomere reduction in low-dose-exposed males, which was absent in females. While the gender differences in RTL may reflect different working conditions, lifestyle and/or telomere biology, absence of a dose response in the highly exposed individuals may reflect selection against cells with short telomeres or induction of telomere-protective effects. Our observations suggest that chronic systemic exposure to radiation leads to variable dose-dependent effects on telomere length.

  2. Design of the internal-recycling circumfluence fluid-bed reactor based on the PLC control%基于PLC控制的内循环回流式流化床反应器的设计

    Institute of Scientific and Technical Information of China (English)

    刘为国; 李松; 陈清华; 吴建锋

    2011-01-01

    The present paper is to introduce our design of the internal-recycle circumfluence fluid-bed reactor based on PLC control. The aim of our design project is to overcome the currently existing problems with the traditional reactor in the organic sewage treatment based on our study of the characteristic features of the Up-flow Anaerobic Sludge Blanket ( UASB) and the Internal Circulation (IC ) reactor in hoping to heighten the treating efficiency of high ammonia organic sewage by installing water distributors to balance the inflow. Besides, an immobilized microorganism filler is added to the reactor to prevent its blockage with a three-phase separator. Besides, the sewage recycling rate has been improved by using an automatic controller and the anaerobic and aerobic integrated device. An internal-recycle return pipe has been plugged to reduce the head loss with the layout of the multi-level sampling ports so as to facilitate the collection of the different height samples. Moreover, with the new design, the reactor can operate at the necessary parameters of the inflow and outflow, upflow and downflow of the water, as well as the aeration intensity, while the aeration time and hydraulic retention time were controlled accurately by the Programmable Logic Controller (PLC). The results of our laboratory experiments show that under the condition of the inflow volume of 1 m3/d, pH 7.2, the operating temperature of 25 ℃ and hydraulic retention time of 3 h, the reactor can work automatically in the alternative anaerobic/aerobic conditions. As a result of the above improvements, the dissolved oxygen (DO) can be kept about 2 mg/L by the micro-aeration, whereas the upflow water velocity is controlled as the inflow can return conveniently at the rate of 40% with no need to use the power, and at the rate of 1.0 m/h. After 60 days of acclimation, the removing rate of CODCr, NH4+ -N, fat and chroma can be made to reach 87%, 63.6%, 74.7% and 84.7%, respectively. Moreover, about 0

  3. Hybrid adsorptive membrane reactor

    Science.gov (United States)

    Tsotsis, Theodore T. (Inventor); Sahimi, Muhammad (Inventor); Fayyaz-Najafi, Babak (Inventor); Harale, Aadesh (Inventor); Park, Byoung-Gi (Inventor); Liu, Paul K. T. (Inventor)

    2011-01-01

    A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

  4. D and DR Reactors

    Data.gov (United States)

    Federal Laboratory Consortium — The world's second full-scale nuclear reactor was the D Reactor at Hanford which was built in the early 1940's and went operational in December of 1944.D Reactor ran...

  5. Hybrid adsorptive membrane reactor

    Science.gov (United States)

    Tsotsis, Theodore T.; Sahimi, Muhammad; Fayyaz-Najafi, Babak; Harale, Aadesh; Park, Byoung-Gi; Liu, Paul K. T.

    2011-03-01

    A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

  6. 瞬发和缓发γ射线对堆内构件释热率影响的研究%Study on the influence of prompt fission γ-ray and delayed γ-ray on reactor internals heating rate

    Institute of Scientific and Technical Information of China (English)

    苏耿华; 石秀安; 蔡德昌; 李雷

    2012-01-01

    To improve the accuracy of the calculated reactor internals heating rate in the design of nuclear power plants, this paper studied the contribution of prompt fission γ to the reactor internals heating rate based on the original method of MCNP external neutron source model. The results revealed that the reactor internals heating rate increased by 9% ~38% with prompt fission γ taken into account and the internals nearer to the core had a lager increment. In addition, it is believed after analysis that the contribution of the delayed γ on reactor internals heating rate is similar to the prompt fission γ. Therefore, when calculating reactor internals heating rate, in addition to the neutron source and neutron capture γ, prompt fission γ and delayed γ should also be considered.%为提高核电设计中反应堆堆内构件释热率计算的准确性,本文在原来MCNP外中子源模型计算方法的基础上,计算分析瞬发裂变γ对堆内构件释热率的贡献.计算结果显示,考虑瞬发裂变γ使得堆内构件的释热率增加9%~38%,离堆芯越近的堆内构件的增加值越大.另外,分析认为缓发γ对堆内构件释热率的贡献与瞬发裂变γ相当.因而反应堆堆内构件释热率计算中除了考虑中子及中子俘获所生γ的贡献,还应该考虑瞬发裂变γ和缓发γ的贡献.

  7. Experiences of violence across life course and its effects on mobility among participants in the International Mobility in Aging Study

    Science.gov (United States)

    Guedes, Dimitri Taurino; Vafaei, Afshin; Alvarado, Beatriz Eugenia; Curcio, Carmen Lucia; Guralnik, Jack M; Zunzunegui, María Victoria; Guerra, Ricardo Oliveira

    2016-01-01

    Background Life course exposure to violence may lead to disability in old age. We examine associations and pathways between life course violence and mobility disability in older participants of the International Mobility in Aging Study (IMIAS). Methods A cross-sectional study using IMIAS 2012 baseline. Men and women aged 65–74 years were recruited at 5 cities (n=1995): Kingston and Saint-Hyacinthe (Canada), Tirana (Albania), Manizales (Colombia) and Natal (Brazil). Mobility was assessed by the Short Physical Performance Battery (SPPB) and by 2 questions on difficulty in walking and climbing stairs. Childhood physical abuse history and the HITS instrument were used to gather information on childhood exposure to violence and violence by intimate partners or family members. Multivariate logistic regression and mediation analysis models were constructed to explore the significance of direct and indirect effects of violence on mobility. Interaction effects of gender on violence and on each of the mediators were tested. Results Experiences of physical violence at any point of life were associated with mobility disability (defined as SPPB<8 or limitation in walking/climbing stairs) while psychological violence was not. Chronic conditions, C reactive protein, physical activity and depression mediated the effect of childhood exposure to violence on both mobility outcomes. Chronic conditions and depression were pathways between family and partner violence and both mobility outcomes. Physical activity was a significant pathway linking family violence to mobility. Gender interactions were not significant. Conclusions Our results provide evidence for the detrimental effects of life course exposure to violence on mobility in later life. PMID:27737884

  8. Uraninite recrystallization and Pb loss in the Oklo and Bangombé natural fission reactors, Gabon

    Science.gov (United States)

    Evins, Lena Z.; Jensen, Keld A.; Ewing, Rodney C.

    2005-03-01

    The Oklo and Bangombé natural fossil fission reactors formed ca. 2 Ga ago in the Franceville basin, Gabon. The response of uraninite in the natural reactors to different geological conditions has implications for the disposal of the UO 2 in spent nuclear fuel. Uraninite and galena from two reactor zones, RZ16 at Oklo and RZB at Bangombé, were studied to clarify the chronology and effect of alteration events on the reactor zones. In addition, ion microprobe U-Pb analysis of zircons from a dolerite dyke in the Oklo deposit were completed to better constrain the age of the dyke, and thereby testing the link between the dyke and an important alteration event in the reactor zones. The analyzed uraninite from RZ16 and RZB contains ca. 6 wt% PbO, indicating a substantial loss of radiogenic Pb. Transmission electron microscopy showed that microscopic uraninite grains in the reactor zones consist of mainly defect-free nanocrystalline to microcrystalline aggregates. However, the nanocrystalline regions have elevated Si contents and lower Pb contents than coarser uraninite crystallites. Single stage model ages of large, millimeter-sized galena grains at both RZ16 and RZB correlate well with the age of the Oklo dolerite dyke, 860 ± 39 Ma (2σ). Thus, the first major Pb loss from uraninite occurred at both Oklo and Bangombé during regional extension and the intrusion of a dyke swarm in the Franceville basin, ˜860-890 Ma ago. Uraninite Pb isotopes from RZ16 and RZB give lower ages of ca. 500 Ma. These ages agree with the "chemical" ages of the uraninite, and show that an ancient Pb loss occurred after the intrusion of the dolerite dykes. The presence of nanocrystallites in the reactor uraninite indicates internal recrystallization, which may have occurred around 500 Ma, resulting in the 6wt% PbO uraninite. It is suggested that leaching by fluid interaction triggered by the Pan-African orogeny was important during this second Pb-loss event. Thus, there are indications that

  9. Fission-Produced (99)Mo Without a Nuclear Reactor.

    Science.gov (United States)

    Youker, Amanda J; Chemerisov, Sergey D; Tkac, Peter; Kalensky, Michael; Heltemes, Thad A; Rotsch, David A; Vandegrift, George F; Krebs, John F; Makarashvili, Vakho; Stepinski, Dominique C

    2017-03-01

    (99)Mo, the parent of the widely used medical isotope (99m)Tc, is currently produced by irradiation of enriched uranium in nuclear reactors. The supply of this isotope is encumbered by the aging of these reactors and concerns about international transportation and nuclear proliferation. Methods: We report results for the production of (99)Mo from the accelerator-driven subcritical fission of an aqueous solution containing low enriched uranium. The predominately fast neutrons generated by impinging high-energy electrons onto a tantalum convertor are moderated to thermal energies to increase fission processes. The separation, recovery, and purification of (99)Mo were demonstrated using a recycled uranyl sulfate solution. Conclusion: The (99)Mo yield and purity were found to be unaffected by reuse of the previously irradiated and processed uranyl sulfate solution. Results from a 51.8-GBq (99)Mo production run are presented.

  10. Degradation of 2,4-dichlorophenoxyacetate isopropyl amine (2,4-D IPA) by O3/AC/UV in an internally slurry airlift photo-reactor.

    Science.gov (United States)

    Farhadian, Negin; Behin, Jamshid

    2017-02-22

    An externally illuminated slurry airlift reactor (ALR) was used to decompose 2,4-dichlorophenoxyacetate isopropyl amine during catalytic ozonation with activated carbon. The effect of superficial gas velocity (0.05-0.15 cm/s), UVAB irradiation (0-60 W), treatment period (10-30 min) and amount of activated carbon (0-0.8 g/l) on removal efficiency was investigated using response surface methodology (RSM) based on the Box-Behnken surface statistical design. Well-defined circulation pattern in the ALR allowed all the fluid elements to be exposed to high light intensity zone and achieve sufficient contact between the solid catalyst and the pollutant. Treatment period appeared as the most influential variable followed by the amount of activated carbon, superficial gas velocity and UV irradiation. A kinetic study was also carried out to evaluate the degradation efficiency versus the O3, O3/AC, O3/UV and O3/AC/UV combinations in which the last one had the highest impact. Efficient suspensions of AC in the ALR resulted in the high efficiency of the O3/AC system. No significant difference was observed between the overall kinetic constants determined in O3/AC and O3/AC/UV systems due to the light transmission obstacle of solid suspension.

  11. Application of Porous Nickel-Coated TiO2 for the Photocatalytic Degradation of Aqueous Quinoline in an Internal Airlift Loop Reactor

    Directory of Open Access Journals (Sweden)

    Mingxin Huo

    2012-02-01

    Full Text Available P25 film, prepared by a facile dip-coating method without any binder, was further developed in a recirculating reactor for quinoline removal from synthetic wastewater. Macroporous foam Ni, which has an open three-dimensional network structure, was utilized as a substrate to make good use of UV rays. Field emission scanning electron microscopy and X-ray diffraction analysis showed that the coated/calcinated P25 films consisted of two crystal phases, and had a number of uniform microcracks on the surface. The effects of initial quinoline concentration, light intensity, reaction temperature, aeration, and initial pH were studied. Increased reaction time, light intensity, environmental temperature, and gas aeration were found to significantly improve the quinoline removal efficiency. The aeration effect of oxygen dependency on the quinoline degradation had the trend pure oxygen > air > no gas > pure nitrogen with free O2. The solution pH crucially affected quinoline photodegradation; the high electrostatic adsorption of quinoline molecules on the TiO2 surface was strongly pH dependent. 2-Pyridine-carboxaldehyde, 3-pyridinecarboxaldehyde, and 2(1H-quinolinone were identified as the major intermediates of quinoline degradation. Based on these intermediates, a primary degradation mechanism was proposed. This reusable P25 film benefits the photodegradation of water contaminants and has potential in other various applications.

  12. Nuclear reactor neutron shielding

    Energy Technology Data Exchange (ETDEWEB)

    Speaker, Daniel P; Neeley, Gary W; Inman, James B

    2017-09-12

    A nuclear reactor includes a reactor pressure vessel and a nuclear reactor core comprising fissile material disposed in a lower portion of the reactor pressure vessel. The lower portion of the reactor pressure vessel is disposed in a reactor cavity. An annular neutron stop is located at an elevation above the uppermost elevation of the nuclear reactor core. The annular neutron stop comprises neutron absorbing material filling an annular gap between the reactor pressure vessel and the wall of the reactor cavity. The annular neutron stop may comprise an outer neutron stop ring attached to the wall of the reactor cavity, and an inner neutron stop ring attached to the reactor pressure vessel. An excore instrument guide tube penetrates through the annular neutron stop, and a neutron plug comprising neutron absorbing material is disposed in the tube at the penetration through the neutron stop.

  13. Role of decommissioning plan and its progress for the PUSPATI TRIGA Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zakaria, Norasalwa, E-mail: norasalwa@nuclearmalaysia.gov.my; Mustafa, Muhammad Khairul Ariff, E-mail: norasalwa@nuclearmalaysia.gov.my; Anuar, Abul Adli, E-mail: norasalwa@nuclearmalaysia.gov.my; Idris, Hairul Nizam, E-mail: norasalwa@nuclearmalaysia.gov.my; Ba' an, Rohyiza, E-mail: norasalwa@nuclearmalaysia.gov.my [Malaysian Nuclear Agency, 43000 Kajang, Selangor (Malaysia)

    2014-02-12

    Malaysian nuclear research reactor, the PUSPATI TRIGA Reactor, reached its first criticality in 1982, and since then, it has been serving for more than 30 years for training, radioisotope production and research purposes. Realizing the age and the need for its decommissioning sometime in the future, a ground basis of assessment and an elaborative project management need to be established, covering the entire process from termination of reactor operation to the establishment of final status, documented as the Decommissioning Plan. At international level, IAEA recognizes the absence of Decommissioning Plan as one of the factors hampering progress in decommissioning of nuclear facilities in the world. Throughout the years, IAEA has taken initiatives and drawn out projects in promoting progress in decommissioning programmes, like CIDER, DACCORD and R2D2P, for which Malaysia is participating in these projects. This paper highlights the concept of Decommissioning plan and its significances to the Agency. It will also address the progress, way forward and challenges faced in developing the Decommissioning Plan for the PUSPATI TRIGA Reactor. The efforts in the establishment of this plan helps to provide continual national contribution at the international level, as well as meeting the regulatory requirement, if need be. The existing license for the operation of PUSPATI TRIGA Reactor does not impose a requirement for a decommissioning plan; however, the renewal of license may call for a decommissioning plan to be submitted for approval in future.

  14. Role of decommissioning plan and its progress for the PUSPATI TRIGA Reactor

    Science.gov (United States)

    Zakaria, Norasalwa; Mustafa, Muhammad Khairul Ariff; Anuar, Abul Adli; Idris, Hairul Nizam; Ba'an, Rohyiza

    2014-02-01

    Malaysian nuclear research reactor, the PUSPATI TRIGA Reactor, reached its first criticality in 1982, and since then, it has been serving for more than 30 years for training, radioisotope production and research purposes. Realizing the age and the need for its decommissioning sometime in the future, a ground basis of assessment and an elaborative project management need to be established, covering the entire process from termination of reactor operation to the establishment of final status, documented as the Decommissioning Plan. At international level, IAEA recognizes the absence of Decommissioning Plan as one of the factors hampering progress in decommissioning of nuclear facilities in the world. Throughout the years, IAEA has taken initiatives and drawn out projects in promoting progress in decommissioning programmes, like CIDER, DACCORD and R2D2P, for which Malaysia is participating in these projects. This paper highlights the concept of Decommissioning plan and its significances to the Agency. It will also address the progress, way forward and challenges faced in developing the Decommissioning Plan for the PUSPATI TRIGA Reactor. The efforts in the establishment of this plan helps to provide continual national contribution at the international level, as well as meeting the regulatory requirement, if need be. The existing license for the operation of PUSPATI TRIGA Reactor does not impose a requirement for a decommissioning plan; however, the renewal of license may call for a decommissioning plan to be submitted for approval in future.

  15. Reactor and method of operation

    Science.gov (United States)

    Wheeler, John A.

    1976-08-10

    A nuclear reactor having a flattened reactor activity curve across the reactor includes fuel extending over a lesser portion of the fuel channels in the central portion of the reactor than in the remainder of the reactor.

  16. Cervical range of motion, cervical and shoulder strength in senior versus age-grade Rugby Union International front-row forwards.

    Science.gov (United States)

    Davies, Mark; Moore, Isabel S; Moran, Patrick; Mathema, Prabhat; Ranson, Craig A

    2016-05-01

    To provide normative values for cervical range of motion (CROM), isometric cervical and shoulder strength for; International Senior professional, and International Age-grade Rugby Union front-row forwards. Cross-sectional population study. All international level front-row players within a Rugby Union Tier 1 Nation. Nineteen Senior and 21 Age-grade front-row forwards underwent CROM, cervical and shoulder strength testing. CROM was measured using the CROM device and the Gatherer System was used to measure multi-directional isometric cervical and shoulder strength. The Age-grade players had significantly lower; cervical strength (26-57% deficits), cervical flexion to extension strength ratios (0.5 vs. 0.6), and shoulder strength (2-36% deficits) than the Senior players. However, there were no differences between front-row positions within each age group. Additionally, there were no differences between age groups or front-row positions in the CROM measurements. Senior Rugby Union front-row forwards have greater cervical and shoulder strength than Age-grade players, with the biggest differences being in cervical strength, highlighting the need for age specific normative values. Importantly, Age-grade players should be evaluated to ensure they have developed sufficient cervical strength prior to entering professional level Rugby Union. Copyright © 2015 Elsevier Ltd. All rights reserved.

  17. COSMIC (Cohort Studies of Memory in an International Consortium): An international consortium to identify risk and protective factors and biomarkers of cognitive ageing and dementia in diverse ethnic and sociocultural groups

    Science.gov (United States)

    2013-01-01

    Background A large number of longitudinal studies of population-based ageing cohorts are in progress internationally, but the insights from these studies into the risk and protective factors for cognitive ageing and conditions like mild cognitive impairment and dementia have been inconsistent. Some of the problems confounding this research can be reduced by harmonising and pooling data across studies. COSMIC (Cohort Studies of Memory in an International Consortium) aims to harmonise data from international cohort studies of cognitive ageing, in order to better understand the determinants of cognitive ageing and neurocognitive disorders. Methods/Design Longitudinal studies of cognitive ageing and dementia with at least 500 individuals aged 60 years or over are eligible and invited to be members of COSMIC. There are currently 17 member studies, from regions that include Asia, Australia, Europe, and North America. A Research Steering Committee has been established, two meetings of study leaders held, and a website developed. The initial attempts at harmonising key variables like neuropsychological test scores are in progress. Discussion The challenges of international consortia like COSMIC include efficient communication among members, extended use of resources, and data harmonisation. Successful harmonisation will facilitate projects investigating rates of cognitive decline, risk and protective factors for mild cognitive impairment, and biomarkers of mild cognitive impairment and dementia. Extended implications of COSMIC could include standardised ways of collecting and reporting data, and a rich cognitive ageing database being made available to other researchers. COSMIC could potentially transform our understanding of the epidemiology of cognitive ageing, and have a world-wide impact on promoting successful ageing. PMID:24195705

  18. Minimizing the fissile inventory of the molten salt fast reactor

    OpenAIRE

    Merle-Lucotte, E.; Heuer, D.; Allibert, M.; Doligez, X.; Ghetta, V.

    2009-01-01

    International audience; Molten salt reactors in the configurations presented here, called Molten Salt Fast Reactors (MSFR), have been selected for further studies by the Generation IV International Forum. These reactors may be operated in simplified and safe conditions in the Th/233U fuel cycle with fluoride salts. We present here the concept, before focusing on a possible optimization in term of minimization of the initial fissile inventory. Our studies demonstrate that an inventory of 233U ...

  19. Bioprinting is coming of age: Report from the International Conference on Bioprinting and Biofabrication in Bordeaux (3B'09).

    Science.gov (United States)

    Guillemot, Fabien; Mironov, Vladimir; Nakamura, Makoto

    2010-03-01

    The International Conference on Bioprinting and Biofabrication in Bordeaux (3B'09) demonstrated that the field of bioprinting and biofabrication continues to evolve. The increasing number and broadening geography of participants, the emergence of new exciting bioprinting technologies, and the attraction of young investigators indicates the strong growth potential of this emerging field. Bioprinting can be defined as the use of computer-aided transfer processes for patterning and assembling living and non-living materials with a prescribed 2D or 3D organization in order to produce bio-engineered structures serving in regenerative medicine, pharmacokinetic and basic cell biology studies. The use of bioprinting technology for biofabrication of in vitro assay has been shown to be a realistic short-term application. At the same time, the principal feasibility of bioprinting vascularized human organs as well as in vivo bioprinting has been demonstrated. The bioprinting of complex 3D human tissues and constructs in vitro and especially in vivo are exciting, but long-term, applications. It was decided that the 5th International Conference on Bioprinting and Biofabrication would be held in Philadelphia, USA in October 2010. The specially appointed 'Eploratory Committee' will consider the possibility of turning the growing bioprinting community into a more organized entity by creating a new bioprinting and biofabrication society. The new journal Biofabrication was also presented at 3B'09. This is an important milestone per se which provides additional objective evidence that the bioprinting and biofabrication field is consolidating and maturing. Thus, it is safe to state that bioprinting technology is coming of age.

  20. Twenty-First Water Reactor Safety Information Meeting. Volume 3, Primary system integrity; Aging research, products and applications; Structural and seismic engineering; Seismology and geology: Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    Monteleone, S. [comp.] [Brookhaven National Lab., Upton, NY (United States)

    1994-04-01

    This three-volume report contains 90 papers out of the 102 that were presented at the Twenty-First Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 25-27, 1993. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Germany, Japan, Russia, Switzerland, Taiwan, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. Selected papers were indexed separately for inclusion in the Energy Science and Technology Database.

  1. New observations by visualizing age stratification and internal dynamics of freshwater lenses in heterogeneous media - laboratory experiments and numerical simulations

    Science.gov (United States)

    Stoeckl, L.; Dose, E.; Houben, G.; Himmelsbach, T.

    2012-12-01

    We performed a series of multi-tracer laboratory scale experiments in a transparent sand-box model to visualize (a) processes during the genesis of freshwater lenses and (b) their internal dynamics. For physical modeling an acrylic glass box was used to simulate a cross section of an island, similar to Stoeckl & Houben (2012). Degassed salt water with a density of 1023 kg/m3 was injected from the bottom, saturating the sand inside the model. Fluorescent tracer dyes uranine, eosine and indigotine were used to mark the infiltrating fresh water from the top. All experiments were filmed and analyzed using fast motion mode. We performed two different types of experimental set-up according to Vacher (1988): (1) Layers of different hydraulic conductivity: By filling the sand-box model with sand of different grain sizes, layers of different hydraulic conductivity could be simulated. (2) Recharge distribution: By recharging the island heterogeneously we could observe shifts in the geometry of the freshwater lens. A novel approach of using different tracer colors and varying them spatially and over time within the recharge waters allowed us to visualize and measure internal flow processes. Age stratification and flow paths could therefore be investigated. Moreover, a combination of temporal and spatial tracer color variation in one single experiment enabled us to measure flow velocities of freshwater movement. Additionally, by injecting small amounts of tracer in the salt water environment, movements near the interface between fresh- and saltwater could be observed. Using the finite element model FEFLOW we could model the density driven dynamics of our small scale freshwater lens, including its formation and the degradation after turning off the recharge water. This is important to fill the gap between our physical sand-box model and ongoing field investigations. The main focus of this work is the effects of climate change as well as geological and morphological

  2. A study of a zone approach to IAEA (International Atomic Energy Agency) safeguards: The low-enriched-uranium zone of a light-water-reactor fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Fishbone, L.G.; Higinbotham, W.A.

    1986-06-01

    At present the IAEA designs its safeguards approach with regard to each type of nuclear facility so that the safeguards activities and effort are essentially the same for a given type and size of nuclear facility wherever it may be located. Conclusions regarding a state are derived by combining the conclusions regarding the effectiveness of safeguards for the individual facilities within a state. In this study it was convenient to define three zones in a state with a closed light-water-reactor nuclear fuel cycle. Each zone contains those facilities or parts thereof which use or process nuclear materials of the same safeguards significance: low-enriched uranium, radioactive spent fuel, or recovered plutonium. The possibility that each zone might be treated as an extended material balance area for safeguards purposes is under investigation. The approach includes defining the relevant features of the facilities in the three zones and listing the safeguards activities which are now practiced. This study has focussed on the fresh-fuel zone, the several facilities of which use or process low-enriched uranium. At one extreme, flows and inventories would be verified at each material balance area. At the other extreme, the flows into and out of the zone and the inventory of the whole zone would be verified. There are a number of possible safeguards approaches which fall between the two extremes. The intention is to develop a rational approach which will make it possible to compare the technical effectiveness and the inspection effort for the facility-oriented approach, for the approach involving the zone as a material balance area, and for some reasonable intermediate safeguards approaches.

  3. Experimental study on the treatment of wastewater from food waste by a new type of internal circulation reactor%新型IC反应器处理餐厨垃圾废水的实验研究

    Institute of Scientific and Technical Information of China (English)

    王罕; 蒋文化; 顾礼炜; 马三剑

    2014-01-01

    采用内循环厌氧反应器(IC)处理餐厨垃圾废水。结果表明:采用快速提升负荷至5 kg/(m3·d)并稳定运行19 d这一启动方式有利于提高污泥的活性。负荷提升中后期,出水pH高于进水pH。IC处理餐厨垃圾废水的最大容积负荷为25.2 kg/(m3·d),此时COD去除率下降到86%。稳定运行期,当进水COD达到22.4 mg/L,出水COD稳定在1650~1950 mg/L,COD去除率高达91.8%。%The new type of internal circulation (IC ) reactor has been used for treating the wastewater from food waste water. The results show that in the start-up period,the start-up form of raising the load rapidly to 5 kg/(m3·d) and running the system steadily for 19 d,is good for improving the sludge activity. In the mid late period of load lifting,the pH of effluent is higher than that of influent. The maximum volume load of food-waste wastewater treated by IC reactor is 25.2 kg/(m3·d). At this time,the COD removing rate declines to 86%. In the steadily running period,when COD concentration of influent reaches 22.4 mg/L,the COD concentration of effluent stabilizes between 1 650-1 950 mg/L,and the COD removing rate reaches 91.8%.

  4. Gender Roles and Physical Function in Older Adults: Cross-Sectional Analysis of the International Mobility in Aging Study (IMIAS).

    Science.gov (United States)

    Ahmed, Tamer; Vafaei, Afshin; Auais, Mohammad; Guralnik, Jack; Zunzunegui, Maria Victoria

    2016-01-01

    To examine the relationships between physical function and gender-stereotyped traits and whether these relationships are modified by sex or social context. A total of 1995 community-dwelling older adults from the International Mobility in Aging Study (IMIAS) aged 65 to 74 years were recruited in Natal (Brazil), Manizales (Colombia), Tirana (Albania), Kingston (Ontario, Canada), and Saint-Hyacinthe (Quebec, Canada). We performed a cross-sectional analysis. Study outcomes were mobility disability, defined as having difficulty in walking 400 meters without assistance or climbing a flight of stairs without resting, and low physical performance, defined as a score gender roles (Masculine, Feminine, Androgynous, and Undifferentiated) using site-specific medians of femininity and masculinity as cut-off points. Poisson regression models were used to estimate prevalence rate ratios (PRR) of mobility disability and poor physical performance according to gender roles. In models adjusted for sex, marital status, education, income, and research site, when comparing to the androgynous role, we found higher prevalence of mobility disability and poor physical performance among participants endorsing the feminine role (PRR = 1.20, 95% confidence interval (CI) 1.03-1.39 and PRR = 1.37, CI 1.01-1.88, respectively) or the undifferentiated role (PRR = 1.23, 95% CI 1.07-1.42 and PRR = 1.58, CI 1.18-2.12, respectively). Participants classified as masculine did not differ from androgynous participants in prevalence rates of mobility disability or low physical performance. None of the multiplicative interactions by sex and research site were significant. Feminine and undifferentiated gender roles are independent risk factors for mobility disability and low physical performance in older adults. Longitudinal research is needed to assess the mediation pathways through which gender-stereotyped traits influence functional limitations and to investigate the longitudinal nature of these

  5. Gender, age, and place of residence as moderators of the internalized homophobia-depressive symptoms relation among Australian gay men and lesbians.

    Science.gov (United States)

    McLaren, Suzanne

    2015-01-01

    Internalized homophobia is a risk factor for depression among gay men and lesbians. The aim of the study was to test whether the internalized homophobia-depression relation was moderated by gender (stronger among gay men compared with lesbians), age (stronger among younger compared with older gay men and lesbians), and place of residence (stronger among gay men and lesbians who live in rural areas compared with those who live in urban areas). An Australian sample of 311 self-identified gay men and 570 self-identified lesbians, aged 18 to 70 years, completed the Internalized Homophobia Scale and the Centre for Epidemiological Studies Depression Scale. Results indicated that age and gender did not moderate the internalized homophobia-depressive symptoms relation. Place of residence was a significant moderator for gay men but not lesbians. In contrast to the hypothesis, the internalized homophobia-depression relation was significant only among gay men who resided in urban areas. Those who work with gay men should be particularly aware of the significant relationship between internalized homophobia and depressive symptoms among gay men who reside in urban areas.

  6. Irradiation testing of internally pressurized and/or graphite coated Zircaloy-4 clad fuel rods in the NRX Reactor (AWBA Development Program). [LWBR

    Energy Technology Data Exchange (ETDEWEB)

    Hoffman, R.C.; Sherman, J.

    1978-11-01

    Irradiation tests on 0.612 inch O.D. by 117-inch long Zircaloy-4 clad fuel rods were performed to assess the effects on fuel rod performance of (1) internal helium pre-pressurization to 500 psi as fabricated, (2) the presence of a graphite barrier coating on the inside cladding surface, and (3) combined pre-pressurization and graphite coating. Periodic dimensional examinations were performed on the test rods, and the results were compared with data obtained from two previously irradiated test rods--both unpressurized and uncoated and one intentionally defected. These comparisons indicate that both pre-pressurization and graphite coating can substantially improve fuel element performance capability.

  7. Nuclear Reactors and Technology; (USA)

    Energy Technology Data Exchange (ETDEWEB)

    Cason, D.L.; Hicks, S.C. (eds.)

    1991-01-01

    Nuclear Reactors and Technology (NRT) announces on a monthly basis the current worldwide information available from the open literature on nuclear reactors and technology, including all aspects of power reactors, components and accessories, fuel elements, control systems, and materials. This publication contains the abstracts of DOE reports, journal articles, conference papers, patents, theses, and monographs added to the Energy Science and Technology Database (EDB) during the past month. Also included are US information obtained through acquisition programs or interagency agreements and international information obtained through the International Energy Agency's Energy Technology Data Exchange or government-to-government agreements. The digests in NRT and other citations to information on nuclear reactors back to 1948 are available for online searching and retrieval on EDB and Nuclear Science Abstracts (NSA) database. Current information, added daily to EDB, is available to DOE and its contractors through the DOE integrated Technical Information System. Customized profiles can be developed to provide current information to meet each user's needs.

  8. Internal filtration in liquid fuel synthesis stirred tank slurry reactor%液体燃料合成浆态反应器内部过滤的研究

    Institute of Scientific and Technical Information of China (English)

    李建文; 钱炜鑫; 李涛; 应卫勇; 房鼎业

    2012-01-01

    The internal filter made of sintered metal plate as filter media was installed to achieve solid-liquid separation in the stirred tank slurry reactor. The influences of pressure drop, temperature, rotational speed, solid content,filter media pore size, and particle diameter on filtration rate were investigated. The results show that clarified and stable filtrate is gained through internal filtration in the slurry reactor, and the cake thickness reaches dynamic balance and the filtration rate tends to stabilize with the operation carrying out. The filtration rate increases with the increase of pressure drop, temperature , filter media pore diameter, particle diameter and the decrease of rotational speed, solid content. The present investigation is useful to practical production. Filtration model is established through dimensional analysis, and values of parameter are gained by using Michael Marquardt algorithm. The statistical test and the comparison between calculated data and experimental data reveal that the model is reliable.%在液体燃料合成浆态搅拌反应器中设置了以金属烧结板为过滤介质的内过滤器,实现固液分离.研究了压降、温度、搅拌转速、固含率、过滤介质孔径及颗粒粒径对过滤速率的影响.实验结果表明:在浆态反应器中进行内过滤可以得到澄清稳定的滤液;随着操作进行,滤饼厚度达到动态平衡,过滤速率最终趋于稳定;过滤速率随着压降、温度、过滤介质孔径、颗粒粒径的增大及搅拌转速、固含率的减小而增大.应用因次分析法建立过滤模型,通过麦夸特算法对实验数据进行最优拟合得到模型的参数值,统计检验以及模型计算值与实验值的比较表明该模型可靠.

  9. The prevalence of internal and external parasites in pigs of different ages and sexes in Southeast District, Botswana.

    Science.gov (United States)

    Nsoso, S J; Mosala, K P; Ndebele, R T; Ramabu, S S

    2000-09-01

    Botswana imports most pig-based products from neighbouring countries. Pig farming is limited by, among other things, the negative effect of parasites and diseases on production. The object of this study was to determine the prevalence of ecto- and endoparasites in pigs of different ages and sexes in the Southeast District of Botswana. Thirty-nine pigs were sampled for endoparasites and 19 for ectoparasites during a period of 2 1/2 months. Of all the pigs sampled, 54,55% were infected with Ascaris suum, 20,45% with Trichostrongylus spp. and 6,82% with Trichuris suis. Ascaris suum was found to be the most common endoparasite infesting both mature, i.e. 12 months and older, and young, i.e. less than 12 months old, pigs. Although not significantly different (P > 0,05), the prevalence of this parasite species was slightly higher (68,42% with an average of 1,023 +/- 545 eggs per gram (EPG) of faeces per pig) in mature than in young pigs (55% with an of average 1,500 +/- 846 EPG of faeces per pig). The prevalence of Trichostrongylus spp. was lower in mature (5,26% with 20 +/- 14 EPG of faeces per pig) than in young pigs (25% with 22 +/- 9 EPG of faeces per pig). The prevalence of T. suis was also lower in mature (0% infection) than in young pigs (15% with 9 +/- 4 EPG of faeces per pig). The prevalence of the three endoparasite species was not significantly different between the sexes A. suum (1,020 +/- 883 v. 1,503 +/- 522 EPG of faeces per pig), Trichostrongylus spp. (24 +/- 14 v. 18 +/- 8 EPG of faeces per pig) and T. suis (11 +/- 6 v. 2 +/- 4 EPG of faeces per pig) for male and female pigs respectively. Sarcoptes scabiei was the only ectoparasite identified on the pigs sampled for external parasites. It infested 40% of all pigs but the infestation on young pigs (70%) was higher than on the mature ones (33,33%). Since the infection of internal and external parasites was similar in young and old pigs of both sexes, controlling parasites is of great importance since these

  10. Gender Roles and Physical Function in Older Adults: Cross-Sectional Analysis of the International Mobility in Aging Study (IMIAS.

    Directory of Open Access Journals (Sweden)

    Tamer Ahmed

    Full Text Available To examine the relationships between physical function and gender-stereotyped traits and whether these relationships are modified by sex or social context.A total of 1995 community-dwelling older adults from the International Mobility in Aging Study (IMIAS aged 65 to 74 years were recruited in Natal (Brazil, Manizales (Colombia, Tirana (Albania, Kingston (Ontario, Canada, and Saint-Hyacinthe (Quebec, Canada. We performed a cross-sectional analysis. Study outcomes were mobility disability, defined as having difficulty in walking 400 meters without assistance or climbing a flight of stairs without resting, and low physical performance, defined as a score < 8 on the Short Physical Performance Battery. The 12-item Bem Sex Role Inventory (BSRI was used to classify participants into four gender roles (Masculine, Feminine, Androgynous, and Undifferentiated using site-specific medians of femininity and masculinity as cut-off points. Poisson regression models were used to estimate prevalence rate ratios (PRR of mobility disability and poor physical performance according to gender roles.In models adjusted for sex, marital status, education, income, and research site, when comparing to the androgynous role, we found higher prevalence of mobility disability and poor physical performance among participants endorsing the feminine role (PRR = 1.20, 95% confidence interval (CI 1.03-1.39 and PRR = 1.37, CI 1.01-1.88, respectively or the undifferentiated role (PRR = 1.23, 95% CI 1.07-1.42 and PRR = 1.58, CI 1.18-2.12, respectively. Participants classified as masculine did not differ from androgynous participants in prevalence rates of mobility disability or low physical performance. None of the multiplicative interactions by sex and research site were significant.Feminine and undifferentiated gender roles are independent risk factors for mobility disability and low physical performance in older adults. Longitudinal research is needed to assess the mediation

  11. Maternal prenatal anxiety and child brain-derived neurotrophic factor (BDNF) genotype: effects on internalizing symptoms from 4 to 15 years of age.

    Science.gov (United States)

    O'Donnell, Kieran J; Glover, Vivette; Holbrook, Joanna D; O'Connor, Thomas G

    2014-11-01

    Multiple behavioral and health outcomes, including internalizing symptoms, may be predicted from prenatal maternal anxiety, depression, or stress. However, not all children are affected, and those that are can be affected in different ways. Here we test the hypothesis that the effects of prenatal anxiety are moderated by genetic variation in the child's brain-derived neurotrophic factor (BDNF) gene, using the Avon Longitudinal Study of Parents and Children population cohort. Internalizing symptoms were assessed from 4 to 13 years of age using the Strengths and Difficulties Questionnaire (n = 8,584); a clinical interview with the adolescents was conducted at age 15 years (n = 4,704). Obstetric and psychosocial risk and postnatal maternal symptoms were included as covariates. Results show that prenatal maternal anxiety predicted internalizing symptoms, including with the diagnostic assessment at 15 years. There was a main effect of two BDNF polymorphisms (rs6265 [val66met] and rs11030104) on internalizing symptoms up to age 13. There was also genetic moderation of the prenatal anxiety effect by different BDNF polymorphisms (rs11030121 and rs7124442), although significant effects were limited to preadolescence. The findings suggest a role for BDNF gene-environment interactions in individual vulnerability to the effects of prenatal anxiety on child internalizing symptoms.

  12. Current status of fast reactor physics

    Energy Technology Data Exchange (ETDEWEB)

    Hummel, H.H.

    1979-01-01

    The subject of calculation of reactivity coefficients for fast reactors is developed, starting with a discussion of the status of relevant nuclear data and proceeding to the subjects of group cross section generation and of methods of obtaining reactivity coefficients from group cross sections. Reactivity coefficients measured in critical experiments are compared with calculated values. Dependence of reactivity coefficients on reactor design is discussed. Finally, results of the recent international comparison of calculated reactivity coefficients are presented.

  13. Does the School Performance Variable Used in the International Health Behavior in School-Aged Children (HBSC) Study Reflect Students' School Grades?

    Science.gov (United States)

    Felder-Puig, Rosemarie; Griebler, Robert; Samdal, Oddrun; King, Matthew A.; Freeman, John; Duer, Wolfgang

    2012-01-01

    Background: Given the pressure that educators and policy makers are under to achieve academic standards for students, understanding the relationship of academic success to various aspects of health is important. The international Health Behavior in School-Aged Children (HBSC) questionnaire, being used in 41 countries with different school and…

  14. Does the School Performance Variable Used in the International Health Behavior in School-Aged Children (HBSC) Study Reflect Students' School Grades?

    Science.gov (United States)

    Felder-Puig, Rosemarie; Griebler, Robert; Samdal, Oddrun; King, Matthew A.; Freeman, John; Duer, Wolfgang

    2012-01-01

    Background: Given the pressure that educators and policy makers are under to achieve academic standards for students, understanding the relationship of academic success to various aspects of health is important. The international Health Behavior in School-Aged Children (HBSC) questionnaire, being used in 41 countries with different school and…

  15. The Relationship between Age of Post-Graduate Adult Learning Students and Learning Style Preferences: A Case of Africa International University, Kenya

    Science.gov (United States)

    Ngala, Francisca Wavinya

    2017-01-01

    This paper sought to examine the relationship between age and learning preferences of post- graduate students at Africa International University (AIU). The study employed a descriptive survey design which used cross-sectional approach to data collection. The population of the study consisted of all the 397 post-graduate students at Africa…

  16. 核电站反应堆冷却剂平均温度内模控制系统仿真%Simulation Study of Internal Model Control System of Refrigerant Average Temperature of Nuclear Reactor

    Institute of Scientific and Technical Information of China (English)

    许天舒; 史小平

    2001-01-01

    采用系统开环脉冲响应序列,创造了一种非参数建模方法,解决了复杂非线性系统的建模难题,并应用内模控制原理创造了一种反应堆冷却剂平均温度恒定控制的非参数模型.该模型设计简单、跟踪调节性能好、鲁棒性强、能消除不可测干扰.仿真实验证明了本设计方法的正确性和有效性,实现了系统的高精度控制.%This paper applies the internal model control method to the refrigerant average temperature control problem of a nuclear reactor, using the open-loop impulse response series of the system as its non-parameter model. The control method presented in this paper is rather simple and has good tracking performance and robustness. The simulation experiment testifies the correctness and effectiveness of the method. The high accuracy control of the system is achieved

  17. Light Water Reactor Sustainability (LWRS) Program – Non-Destructive Evaluation (NDE) R&D Roadmap for Determining Remaining Useful Life of Aging Cables in Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Simmons, Kevin L.; Ramuhalli, Pradeep; Brenchley, David L.; Coble, Jamie B.; Hashemian, Hash; Konnik, Robert; Ray, Sheila

    2012-09-14

    The purpose of the non-destructive evaluation (NDE) R&D Roadmap for Cables is to support the Materials Aging and Degradation (MAaD) R&D pathway. The focus of the workshop was to identify the technical gaps in detecting aging cables and predicting their remaining life expectancy. The workshop was held in Knoxville, Tennessee, on July 30, 2012, at Analysis and Measurement Services Corporation (AMS) headquarters. The workshop was attended by 30 experts in materials, electrical engineering, U.S. Nuclear Regulatory Commission (NRC), U.S. Department of Energy (DOE) National Laboratories (Oak Ridge National Laboratory, Pacific Northwest National Laboratory, Argonne National Laboratory, and Idaho National Engineering Laboratory), NDE instrumentation development, universities, commercial NDE services and cable manufacturers, and Electric Power Research Institute (EPRI). The motivation for the R&D roadmap comes from the need to address the aging management of in-containment cables at nuclear power plants (NPPs).

  18. Fe-15Ni-13Cr austenitic stainless steels for fission and fusion reactor applications - Part II: Effects of minor elements on precipitate phase stability during thermal aging

    Science.gov (United States)

    Lee, E. H.; Mansur, L. K.

    2000-01-01

    The precipitate phase stability in Fe-15Ni-13Cr base austenitic alloys was investigated as a function of minor alloying additions after thermally aging at 600°C and 675°C for times ranging from 24 h to one year. Seven major precipitate phases were found in aged specimens, including M 23C 6, Laves, Eta (η), TiO, NbC, MC, and M 2P. The types and amounts of precipitate phases varied with alloying element additions, aging temperature, and aging time. By analyzing the composition of each individual particle, it was possible to determine the essential constituent elements for each phase. From this information, a strategy to promote or suppress certain precipitate phases was developed. Among the seven phases, the most desirable precipitate phases were considered to be MC and M 2P, because these particles form on a fine scale with a high number density and, therefore, can serve as effective gas atom trap sites under irradiation.

  19. Reactor Physics Programme

    Energy Technology Data Exchange (ETDEWEB)

    De Raedt, C

    2000-07-01

    The Reactor Physics and Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis on reactor fuel. This expertise is applied within the Reactor Physics and MYRRHA Research Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments. Progress and achievements in 1999 in the following areas are reported on: (1) investigations on the use of military plutonium in commercial power reactors; (2) neutron and gamma calculations performed for BR-2 and for other reactors; (3) the updating of neutron and gamma cross-section libraries; (4) the implementation of reactor codes; (6) the management of the UNIX workstations; and (6) fuel cycle studies.

  20. Attrition reactor system

    Science.gov (United States)

    Scott, Charles D.; Davison, Brian H.

    1993-01-01

    A reactor vessel for reacting a solid particulate with a liquid reactant has a centrifugal pump in circulatory flow communication with the reactor vessel for providing particulate attrition, resulting in additional fresh surface where the reaction can occur.

  1. Light Water Reactor Sustainability (LWRS) Program – Non-Destructive Evaluation (NDE) R&D Roadmap for Determining Remaining Useful Life of Aging Cables in Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Simmons, K.L.; Ramuhali, P.; Brenchley, D.L.; Coble, J.B.; Hashemian, H.M.; Konnick, R.; Ray, S.

    2012-09-01

    Executive Summary [partial] The purpose of the non-destructive evaluation (NDE) R&D Roadmap for Cables is to support the Materials Aging and Degradation (MAaD) R&D pathway. A workshop was held to gather subject matter experts to develop the NDE R&D Roadmap for Cables. The focus of the workshop was to identify the technical gaps in detecting aging cables and predicting their remaining life expectancy. The workshop was held in Knoxville, Tennessee, on July 30, 2012, at Analysis and Measurement Services Corporation (AMS) headquarters. The workshop was attended by 30 experts in materials, electrical engineering, and NDE instrumentation development from the U.S. Nuclear Regulatory Commission (NRC), U.S. Department of Energy (DOE) National Laboratories (Oak Ridge National Laboratory, Pacific Northwest National Laboratory, Argonne National Laboratory, and Idaho National Engineering Laboratory), universities, commercial NDE service vendors and cable manufacturers, and the Electric Power Research Institute (EPRI).

  2. Study on microstructural changes in thermally-aged stainless steel weld-overlay cladding of nuclear reactor pressure vessels by atom probe tomography

    Energy Technology Data Exchange (ETDEWEB)

    Takeuchi, T., E-mail: takeuchi.tomoaki@jaea.go.jp [Japan Atomic Energy Agency, Tokai, Ibaraki 319-1195 (Japan); Kameda, J. [National Institute for Materials Science, Sengen, Tsukuba 305-0047 (Japan); Nagai, Y.; Toyama, T. [Oarai Center, Institute for Materials Research, Tohoku University, Oarai, Ibaraki 311-1313 (Japan); Nishiyama, Y.; Onizawa, K. [Japan Atomic Energy Agency, Tokai, Ibaraki 319-1195 (Japan)

    2011-08-15

    Highlights: > Microstructural changes in stainless steel electroslag weld-overlay cladding. > Thermal aging caused progress of spinodal decomposition and precipitation of G phases in the {delta}-ferrite phase. > The degree of the spinodal decomposition had a linear relationship to the hardness. - Abstract: The effect of thermal aging on microstructural changes was investigated in stainless steel weld-overlay cladding composed of 90% austenite and 10% {delta}-ferrite phases using atom probe tomography (APT). In as-received materials subjected to cooling process after post-welding heat treatments (PWHT), a slight fluctuation of the Cr concentration was already observed due to spinodal decomposition in the ferrite phase but not in the austenitic phase. Thermal aging at 400 deg. C for 10,000 h caused not only an increase in the amplitude of spinodal decomposition but also the precipitation of G phases with composition ratios of Ni:Si:Mn = 16:7:6 in the ferrite phase. The chemical compositions of M{sub 23}C{sub 6} type carbides seemed to be formed at the austenite/ferrite interface were analyzed. The analyses of the magnitude of the spinodal decomposition and the hardness implied that the spinodal decomposition was the main cause of the hardening.

  3. Long-Run Labor Market Effects of Japanese American Internment during World War II on Working-Age Male Internees

    OpenAIRE

    Aimee Chin

    2005-01-01

    In 1942, all Japanese were evacuated from the West Coast and incarcerated in internment camps. To investigate the long-run economic consequences of this historic episode, I exploit the fact that Hawaiian Japanese were not subject to mass internment. I find that the labor market withdrawal induced by the internment reduced the annual earnings of males by as much as 9%13% 25 years afterward. This is consistent with the predictions of an economic model that equates the labor market withdrawal in...

  4. LBB application in the US operating and advanced reactors

    Energy Technology Data Exchange (ETDEWEB)

    Wichman, K.; Tsao, J.; Mayfield, M.

    1997-04-01

    The regulatory application of leak before break (LBB) for operating and advanced reactors in the U.S. is described. The U.S. Nuclear Regulatory Commission (NRC) has approved the application of LBB for six piping systems in operating reactors: reactor coolant system primary loop piping, pressurizer surge, safety injection accumulator, residual heat removal, safety injection, and reactor coolant loop bypass. The LBB concept has also been applied in the design of advanced light water reactors. LBB applications, and regulatory considerations, for pressurized water reactors and advanced light water reactors are summarized in this paper. Technology development for LBB performed by the NRC and the International Piping Integrity Research Group is also briefly summarized.

  5. Progress with OPAL, the new Australian research reactor

    Indian Academy of Sciences (India)

    R A Robinson

    2008-11-01

    Australian science is entering a new `golden age', with the start-up of bright new neutron and photon sources in Sydney and Melbourne, in 2006 and 2007 respectively. The OPAL reactor and the Australian Synchrotron can be considered as the greatest single investment in scientific infrastructure in Australia's history. They will essentially be `sister' facilities, with a common open user ethos, and a vision to play a major role in international science. Fuel was loaded into the reactor in August 2006, and full power was (20 MW) achieved in November 2006. The first call for proposals was made in 2007, and commissioning experiments have taken place well before then. The first three instruments in operation are high-resolution powder diffractometer (for materials discovery), high-intensity powder diffractometer (for kinetics experiments and small samples) and a strain scanner (for mechanical engineering and industrial applications). These are closely followed by four more instruments with broad application in nanoscience, condensed matter physics and other scientific disciplines. Instrument performance will be competitive with the best research-reactor facilities anywhere. To date there is committed funding for nine instruments, with a capacity to install a total of ∼ 18 beamlines. An update will be given on the status of OPAL, its thermal and cold neutron sources, its instruments and the first results.

  6. LMFBR type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kawakami, Hiroto

    1995-02-07

    A reactor container of the present invention has a structure that the reactor container is entirely at the same temperature as that at the inlet of the reactor and, a hot pool is incorporated therein, and the reactor container has is entirely at the same temperature and has substantially uniform temperature follow-up property transiently. Namely, if the temperature at the inlet of the reactor core changes, the temperature of the entire reactor container changes following this change, but no great temperature gradient is caused in the axial direction and no great heat stresses due to axial temperature distribution is caused. Occurrence of thermal stresses caused by the axial temperature distribution can be suppressed to improve the reliability of the reactor container. In addition, since the laying of the reactor inlet pipelines over the inside of the reactor is eliminated, the reactor container is made compact and the heat shielding structures above the reactor and a protection structure of container walls are simplified. Further, secondary coolants are filled to the outside of the reactor container to simplify the shieldings. The combined effects described above can improve economical property and reliability. (N.H.).

  7. Improved safety fast reactor with “reservoir” for delayed neutrons generating

    Science.gov (United States)

    Kulikov, G. G.; Apse, V. A.; Shmelev, A. N.; Kulikov, E. G.

    2017-01-01

    The paper considers the possibility to improve safety of fast reactors by using weak neutron absorber with large atomic weight as a material for external neutron reflector and for internal cavity in the reactor core (the neutron “reservoir”) where generation of some additional “delayed” neutron takes place. The effects produced by the external neutron reflector and the internal neutron “reservoir” on kinetic behavior of fast reactors are inter-compared. It is demonstrated that neutron kinetics of fast reactors with such external and internal zones becomes the quieter as compared with neutron kinetics of thermal reactors.

  8. Parent behaviors moderate the relationship between neonatal pain and internalizing behaviors at 18 months corrected age in children born very prematurely.

    Science.gov (United States)

    Vinall, Jillian; Miller, Steven P; Synnes, Anne R; Grunau, Ruth E

    2013-09-01

    Children born very preterm (≤ 32 weeks gestation) exhibit greater internalizing (anxious/depressed) behaviors compared to term-born peers as early as 2 years corrected age (CA); however, the role of early stress in the etiology of internalizing problems in preterm children remains unknown. Therefore, we examined the relationship between neonatal pain and internalizing behavior at 18 months CA in children born very preterm and examined whether parent behavior and stress moderated this relationship. Participants were 145 children (96 very preterm, 49 full term) assessed at 18 months CA. Neonatal data were obtained from medical and nursing chart review. Neonatal pain was defined as the number of skin-breaking procedures. Cognitive ability was measured with the Bayley Scales of Infant Development II. Parents completed the Parenting Stress Index III, Child Behavior Checklist 1.5-5, and participated in a videotaped play session with their child, which was coded using the Emotional Availability Scale IV. Very preterm children displayed greater Internalizing behaviors compared to full-term control children (P=.02). Parent Sensitivity and Nonhostility moderated the relationship between neonatal pain and Internalizing behavior (all Pneonatal medical confounders, gender, and child cognitive ability (all P>.05). Parent Emotional Availability and stress were not associated with Internalizing behaviors in full-term control children. Positive parent interaction and lower stress appears to ameliorate negative effects of neonatal pain on stress-sensitive behaviors in this vulnerable population.

  9. Conduction heat transfer in a cylindrical dielectric barrier discharge reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sadat, H. [Laboratoire d' Etudes Thermiques, Universite de Poitiers, 40 Avenue du Recteur Pineau, 86022 Poitiers (France)], E-mail: hamou.sadat@univ-poitiers.fr; Dubus, N. [Laboratoire d' Etudes Thermiques, Universite de Poitiers, 40 Avenue du Recteur Pineau, 86022 Poitiers (France); Pinard, L.; Tatibouet, J.M.; Barrault, J. [Laboratoire en catalyse et chimie organique, Universite de Poitiers, 40 Avenue du Recteur Pineau, 86022 Poitiers (France)

    2009-04-15

    The thermal behaviour of a dielectric barrier discharge reactor is studied. The experimental tests are performed on a laboratory reactor with two working fluids: helium and air. A simple heat conduction model for calculating the heat loss is developed. By using temperature measurements in the internal and external electrodes, a thermal resistance of the reactor is defined. Finally, the percentage of the input power that is dissipated to the environment is given.

  10. Using thermal balance model to determine optimal reactor volume and insulation material needed in a laboratory-scale composting reactor.

    Science.gov (United States)

    Wang, Yongjiang; Pang, Li; Liu, Xinyu; Wang, Yuansheng; Zhou, Kexun; Luo, Fei

    2016-04-01

    A comprehensive model of thermal balance and degradation kinetics was developed to determine the optimal reactor volume and insulation material. Biological heat production and five channels of heat loss were considered in the thermal balance model for a representative reactor. Degradation kinetics was developed to make the model applicable to different types of substrates. Simulation of the model showed that the internal energy accumulation of compost was the significant heat loss channel, following by heat loss through reactor wall, and latent heat of water evaporation. Lower proportion of heat loss occurred through the reactor wall when the reactor volume was larger. Insulating materials with low densities and low conductive coefficients were more desirable for building small reactor systems. Model developed could be used to determine the optimal reactor volume and insulation material needed before the fabrication of a lab-scale composting system.

  11. Light water reactor safety

    CERN Document Server

    Pershagen, B

    2013-01-01

    This book describes the principles and practices of reactor safety as applied to the design, regulation and operation of light water reactors, combining a historical approach with an up-to-date account of the safety, technology and operating experience of both pressurized water reactors and boiling water reactors. The introductory chapters set out the basic facts upon which the safety of light water reactors depend. The central section is devoted to the methods and results of safety analysis. The accidents at Three Mile Island and Chernobyl are reviewed and their implications for light wate

  12. Nuclear reactor physics

    CERN Document Server

    Stacey, Weston M

    2010-01-01

    Nuclear reactor physics is the core discipline of nuclear engineering. Nuclear reactors now account for a significant portion of the electrical power generated worldwide, and new power reactors with improved fuel cycles are being developed. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. The second edition of this successful comprehensive textbook and reference on basic and advanced nuclear reactor physics has been completely updated, revised and enlarged to include the latest developme

  13. Safety enhancement by transposition of the nitration of toluene from semi-batch reactor to continuous intensified heat exchanger reactor

    OpenAIRE

    Di Miceli Raimondi, Nathalie; Olivier Maget, Nelly; Gabas, Nadine; Cabassud, Michel; GOURDON, Christophe

    2015-01-01

    International audience; The behaviour of a continuous intensified heat exchanger (HEX) reactor in case of process failure is analysed and compared to the behaviour of a semi-continuous reactor. The nitration of toluene is considered as test reaction to identify the main failure scenarios that can lead to thermal runaway in both processes using the HAZOP method.No flow rate of process fluid and utility fluid in the continuous process. No stirring during feeding of the reactor followed by norma...

  14. Validity and internal consistency of the Ages and Stages Questionnaire 60-month version and the effect of three scoring methods

    NARCIS (Netherlands)

    Hornman, Jorijn; Kerstjens, Jorien M; de Winter, Andrea F; Bos, Arend F; Reijneveld, Sijmen A

    2013-01-01

    Background: The Ages and Stages Questionnaire (ASQ) is currently the most used parent-completed developmental screener consisting of different age-specific questionnaires. Psychometric evaluation of the ASQ 60-month version (ASQ-60) is limited. Furthermore, it is unclear which of the available scori

  15. A summary of the proceedings of the Eleventh International Symposium on the Neurobiology and Neuroendocrinology of Aging, Bregenz, Austria, July 29-August 3, 2012.

    Science.gov (United States)

    Brown-Borg, Holly M; Borg, Kurt E

    2013-07-01

    A summary of the Eleventh International Symposium on the Neurobiology and Neuroendocrinology of Aging that was held in July 29-August 3 in Bregenz, Austria, is presented. Sixteen of the speakers who presented at the conference submitted review papers covering the topic of their presentation as well as an overview of their respective fields and are included in this special issue. The abstracts from each poster presentation are also included at the end of the special issue.

  16. Spinning fluids reactor

    Science.gov (United States)

    Miller, Jan D; Hupka, Jan; Aranowski, Robert

    2012-11-20

    A spinning fluids reactor, includes a reactor body (24) having a circular cross-section and a fluid contactor screen (26) within the reactor body (24). The fluid contactor screen (26) having a plurality of apertures and a circular cross-section concentric with the reactor body (24) for a length thus forming an inner volume (28) bound by the fluid contactor screen (26) and an outer volume (30) bound by the reactor body (24) and the fluid contactor screen (26). A primary inlet (20) can be operatively connected to the reactor body (24) and can be configured to produce flow-through first spinning flow of a first fluid within the inner volume (28). A secondary inlet (22) can similarly be operatively connected to the reactor body (24) and can be configured to produce a second flow of a second fluid within the outer volume (30) which is optionally spinning.

  17. Nuclear reactor alignment plate configuration

    Energy Technology Data Exchange (ETDEWEB)

    Altman, David A; Forsyth, David R; Smith, Richard E; Singleton, Norman R

    2014-01-28

    An alignment plate that is attached to a core barrel of a pressurized water reactor and fits within slots within a top plate of a lower core shroud and upper core plate to maintain lateral alignment of the reactor internals. The alignment plate is connected to the core barrel through two vertically-spaced dowel pins that extend from the outside surface of the core barrel through a reinforcement pad and into corresponding holes in the alignment plate. Additionally, threaded fasteners are inserted around the perimeter of the reinforcement pad and into the alignment plate to further secure the alignment plate to the core barrel. A fillet weld also is deposited around the perimeter of the reinforcement pad. To accomodate thermal growth between the alignment plate and the core barrel, a gap is left above, below and at both sides of one of the dowel pins in the alignment plate holes through with the dowel pins pass.

  18. Development of fault diagnostic technique using reactor noise analysis

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jin Ho; Kim, J. S.; Oh, I. S.; Ryu, J. S.; Joo, Y. S.; Choi, S.; Yoon, D. B

    1999-04-01

    The ultimate goal of this project is to establish the analysis technique to diagnose the integrity of reactor internals using reactor noise. The reactor noise analyses techniques for the PWR and CANDU NPP(Nuclear Power Plants) were established by which the dynamic characteristics of reactor internals and SPND instrumentations could be identified, and the noise database corresponding to each plant(both Korean and foreign one) was constructed and compared. Also the change of dynamic characteristics of the Ulchin 1 and 2 reactor internals were simulated under presumed fault conditions. Additionally portable reactor noise analysis system was developed so that real time noise analysis could directly be able to be performed at plant site. The reactor noise analyses techniques developed and the database obtained from the fault simulation, can be used to establish a knowledge based expert system to diagnose the NPP's abnormal conditions. And the portable reactor noise analysis system may be utilized as a substitute for plant IVMS(Internal Vibration Monitoring System). (author)

  19. Small reactors in the Canadian context: opportunities and challenges

    Energy Technology Data Exchange (ETDEWEB)

    Walker, R.S. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2013-07-01

    This presentation discusses the opportunities and challenges for small reactors in Canada. It concludes by suggesting that the success of small reactors in Canada will depend on a number of factors including private sector investment, access to international markets, stable, equitable and adaptable regulatory regime, public trust and technology.

  20. Normal sizes of internal jugular veins in children/adolescents aged birth to 18 years at rest and during the Valsalva maneuver

    Energy Technology Data Exchange (ETDEWEB)

    Eksioglu, Ayse Secil, E-mail: yucelsecil@yahoo.com [Dr. Sami Ulus Women and Children' s Hospital, Radiology Department, Babür Caddesi No:4, 06080 Altındağ, Ankara (Turkey); Tasci Yildiz, Yasemin, E-mail: ytasciyildizl@yahoo.com [Dr. Sami Ulus Women and Children' s Hospital, Radiology Department, Babür Caddesi No:4, 06080 Altındağ, Ankara (Turkey); Senel, Saliha, E-mail: drsaliha007@yahoo.com.tr [Dr. Sami Ulus Women and Children' s Hospital, Department of Pediatrics, Babür Caddesi No:4, 06080 Altındağ, Ankara (Turkey)

    2014-04-15

    Objectives: We aimed to establish normal ultrasonographic (US) values of internal jugular vein (IJV) sizes in children/adolescents aged birth to 18 years and to determine the correlation of US measurements with age, height, weight and body surface area (BSA) of children in different age groups. Methods: Two hundred and thirty-six healthy children (0–18 years) were divided into four groups according to their age (0–2, 3–6, 7–12, and 13–18 years). US measurements (transverse, anteroposterior diameter, and cross-sectional area at rest and during the Valsalva maneuver) of bilateral IJVs were taken at the level of cricoid cartilage. Results: Our study gives information about the reference values in children between birth to 18 years of age. There were significant differences between measurements taken at rest and during the Valsalva maneuver in all age groups. Moderate to strong correlations (clinically significant) between age, height and BSA of the subjects and IJV measurements were detected only in the 0–2 years age group. The strength of the correlations decreased with increasing age. Pearson's correlation revealed that height had the strongest and weight had the weakest correlation with US measurements. ‘Height’ was an independent variable on the right, and ‘age’ on the left side, except for rest CSA, when a regression analysis was performed for clinically significant correlations. Conclusions: Determination of normal reference values for US measurements of the IJV and knowledge of correlation with age, height, weight and BSA might be valuable during interventional procedures and for the diagnosis of phlebectasia in children/adolescents.

  1. Technological status of reactor coolant pumps in generation III+ pressurized nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Brecht, Bernhard; Bross, Stephan [KSB Aktiengesellschaft, Frankenthal (Germany)

    2016-05-15

    KSB has been developing and producing pumps for thermal power plants for nearly 90 years. Consequently, KSB also started to develop and manufacture pumps for all kinds of nuclear power plants from the very beginning of the civil use of nuclear energy. This is especially true for reactor coolant pumps for pressurized water reactors. For the generation of advanced evolutionary reactors (Generation III+ reactors), KSB developed an advanced shaft seal system which is also able to fulfill the requirements of station blackout conditions. The tests in the KSB test rigs, which were successfully completed in December 2015, proved the full functionality of the new design. For generation III+ passive plant reactors KSB developed a new reactor coolant pump type called RUV, which is based on the experience of classic reactor coolant pumps and reactor internal pumps. It is a very compact, hermetically sealed vertical pump-motor unit with a wet winding motor. A full scale prototype successfully passed the 1st stage qualification test program in October 2015.

  2. Helmholtz in Gilded-Age America: The International Electrical Congress of 1893 and the Relations of Science and Technology

    Science.gov (United States)

    Cahan, David

    2010-01-01

    This essay recounts Hermann von Helmholtz's trip to represent Germany at the International Electrical Congress in Chicago in 1893 as well as his reception by various members of the American scientific, technological, and cultural elite in several other American cities. In doing so, it seeks to portray something of the vitality of the youthful and…

  3. Age at menarche and menopause and breast cancer risk in the International BRCA1/2 Carrier Cohort Study.

    NARCIS (Netherlands)

    Chang-Claude, J.; Andrieu, N.; Rookus, M.A.; Brohet, R.M.; Antoniou, A.C.; Peock, S.; Davidson, R.; Izatt, L.; Cole, T.; Nogues, C.; Luporsi, E.; Huiart, L.; Hoogerbrugge, N.; Leeuwen, F.E. van; Osorio, A.; Eyfjord, J.; Radice, P.; Goldgar, D.E.; Easton, D.F.

    2007-01-01

    BACKGROUND: Early menarche and late menopause are important risk factors for breast cancer, but their effects on breast cancer risk in BRCA1 and BRCA2 carriers are unknown. METHODS: We assessed breast cancer risk in a large series of 1,187 BRCA1 and 414 BRCA2 carriers from the International BRCA1/2

  4. Reactor Vessel Surveillance Program for Advanced Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Kyeong-Hoon; Kim, Tae-Wan; Lee, Gyu-Mahn; Kim, Jong-Wook; Park, Keun-Bae; Kim, Keung-Koo

    2008-10-15

    This report provides the design requirements of an integral type reactor vessel surveillance program for an integral type reactor in accordance with the requirements of Korean MEST (Ministry of Education, Science and Technology Development) Notice 2008-18. This report covers the requirements for the design of surveillance capsule assemblies including their test specimens, test block materials, handling tools, and monitors of the surveillance capsule neutron fluence and temperature. In addition, this report provides design requirements for the program for irradiation surveillance of reactor vessel materials, a layout of specimens and monitors in the surveillance capsule, procedures of installation and retrieval of the surveillance capsule assemblies, and the layout of the surveillance capsule assemblies in the reactor.

  5. Generation IV Reactors Integrated Materials Technology Program Plan: Focus on Very High Temperature Reactor Materials

    Energy Technology Data Exchange (ETDEWEB)

    Corwin, William R [ORNL; Burchell, Timothy D [ORNL; Katoh, Yutai [ORNL; McGreevy, Timothy E [ORNL; Nanstad, Randy K [ORNL; Ren, Weiju [ORNL; Snead, Lance Lewis [ORNL; Wilson, Dane F [ORNL

    2008-08-01

    requirements. (4) Pressure Vessel Steels: (a) Qualification of short-term, high-temperature properties of light water reactor steels for anticipated VHTR off-normal conditions must be determined, as well as the effects of aging on tensile, creep, and toughness properties, and on thermal emissivity. (b) Large-scale fabrication process for higher temperature alloys, such as 9Cr-1MoV, including ensuring thick-section and weldment integrity must be developed, as well as improved definitions of creep-fatigue and negligible creep behavior. (5) High-Temperature Alloys: (a) Qualification and codification of materials for the intermediate heat exchanger, such as Alloys 617 or 230, for long-term very high-temperature creep, creep-fatigue, and environmental aging degradation must be done, especially in thin sections for compact designs, for both base metal and weldments. (b) Constitutive models and an improved methodology for high-temperature design must be developed.

  6. Summary and Outlook of the International Workshop on Aging Phenomena in Gaseous Detectors (DESY, Hamburg, October, 2001)

    CERN Document Server

    Titov, M L; Padilla, C; Tesch, N

    2002-01-01

    High Energy Physics experiments are currently entering a new era which requires the operation of gaseous particle detectors at unprecedented high rates and integrated particle fluxes. Full functionality of such detectors over the lifetime of an experiment in a harsh radiation environment is of prime concern to the involved experimenters. New classes of gaseous detectors such as large-scale straw-type detectors, Micro-pattern Gas Detectors and related detector types with their own specific aging effects have evolved since the first workshop on wire chamber aging was held at LBL, Berkeley in 1986. In light of these developments and as detector aging is a notoriously complex field, the goal of the workshop was to provide a forum for interested experimentalists to review the progress in understanding of aging effects and to exchange recent experiences. A brief summary of the main results and experiences reported at the 2001 workshop is presented, with the goal of providing a systematic review of aging effects in ...

  7. 77 FR 39521 - Application for a License To Export Nuclear Reactor Major Components and Equipment

    Science.gov (United States)

    2012-07-03

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Application for a License To Export Nuclear Reactor Major Components and Equipment Pursuant to 10... Reactor internals, Components and For use in Braka nuclear power Company LLC reactor coolant equipment...

  8. Light Water Reactor Sustainability Constellation Pilot Project FY11 Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    R. Johansen

    2011-09-01

    Summary report for Fiscal Year 2011 activities associated with the Constellation Pilot Project. The project is a joint effor between Constellation Nuclear Energy Group (CENG), EPRI, and the DOE Light Water Reactor Sustainability Program. The project utilizes two CENG reactor stations: R.E. Ginna and Nine Point Unit 1. Included in the report are activities associate with reactor internals and concrete containments.

  9. Effect of age and sex on the production of internal and external hydrocarbons and pheromones in the housefly, Musca domestica.

    Science.gov (United States)

    Mpuru, S; Blomquist, G J; Schal, C; Roux, M; Kuenzli, M; Dusticier, G; Clément, J L; Bagnères, A G

    2001-02-01

    The epicuticular and internal waxes of male and female houseflies were examined by capillary gas chromatography-mass spectrometry at closely timed intervals from emergence until day-6 of adulthood. New components identified included tricosan-10-one, 9,10-epoxyheptacosane, heptacosen-12-one, a series of odd-carbon numbered dienes from C31 to C39, several positional isomers of monoenes including (Z)-9- and 7-pentacosene and a number of methyl- and dimethylalkanes. (Z)-9-tricosene appears in internal lipids prior to appearing on the surface of the insect, suggesting that it is transported in the hemolymph to its site of deposition on the epicuticle. The large increases in the amount of (Z)-9-tricosene in females from day-2 until day-6 is compensated for by a concomitant decrease in (Z)-9-heptacosene. The C23 epoxide and ketone only appear in females after the production of (Z)-9-tricosene is induced, and are only abundant in epicuticular waxes, suggesting they are formed after (Z)-9-tricosene is transported to the cells which are involved in taking them to the surface of the insect. Mathematical analysis indicated that the time shift between internal production and external accumulation in females is more than 24 h. The divergence between male and female lipid production occurs at an early stage, when insects are less than one day old.

  10. SNTP program reactor design

    Science.gov (United States)

    Walton, Lewis A.; Sapyta, Joseph J.

    1993-06-01

    The Space Nuclear Thermal Propulsion (SNTP) program is evaluating the feasibility of a particle bed reactor for a high-performance nuclear thermal rocket engine. Reactors operating between 500 MW and 2,000 MW will produce engine thrusts ranging from 20,000 pounds to 80,000 pounds. The optimum reactor arrangement depends on the power level desired and the intended application. The key components of the reactor have been developed and are being tested. Flow-to-power matching considerations dominate the thermal-hydraulic design of the reactor. Optimal propellant management during decay heat cooling requires a three-pronged approach. Adequate computational methods exist to perform the neutronics analysis of the reactor core. These methods have been benchmarked to critical experiment data.

  11. Hybrid reactors. [Fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Moir, R.W.

    1980-09-09

    The rationale for hybrid fusion-fission reactors is the production of fissile fuel for fission reactors. A new class of reactor, the fission-suppressed hybrid promises unusually good safety features as well as the ability to support 25 light-water reactors of the same nuclear power rating, or even more high-conversion-ratio reactors such as the heavy-water type. One 4000-MW nuclear hybrid can produce 7200 kg of /sup 233/U per year. To obtain good economics, injector efficiency times plasma gain (eta/sub i/Q) should be greater than 2, the wall load should be greater than 1 MW.m/sup -2/, and the hybrid should cost less than 6 times the cost of a light-water reactor. Introduction rates for the fission-suppressed hybrid are usually rapid.

  12. Fast Spectrum Reactors

    CERN Document Server

    Todd, Donald; Tsvetkov, Pavel

    2012-01-01

    Fast Spectrum Reactors presents a detailed overview of world-wide technology contributing to the development of fast spectrum reactors. With a unique focus on the capabilities of fast spectrum reactors to address nuclear waste transmutation issues, in addition to the well-known capabilities of breeding new fuel, this volume describes how fast spectrum reactors contribute to the wide application of nuclear power systems to serve the global nuclear renaissance while minimizing nuclear proliferation concerns. Readers will find an introduction to the sustainable development of nuclear energy and the role of fast reactors, in addition to an economic analysis of nuclear reactors. A section devoted to neutronics offers the current trends in nuclear design, such as performance parameters and the optimization of advanced power systems. The latest findings on fuel management, partitioning and transmutation include the physics, efficiency and strategies of transmutation, homogeneous and heterogeneous recycling, in addit...

  13. Reactor for Photocatalytic Degradation of Chloroform

    DEFF Research Database (Denmark)

    Simonsen, Morten Enggrob; Søgaard, Erik Gydesen

    In the present study a new type of continuous photoreactor is developed in which the TiO2 catalyst is immobilized on the surface of quartz tubes surrounding the UV lamps and on the internal surface of the reactor walls. The study showed that an initial concentration chloroform of 7 mg/l was degra...

  14. International Guidelines on Sexuality Education and Their Relevance to a Contemporary Curriculum for Children Aged 5-8 Years

    Science.gov (United States)

    Goldman, Juliette D. G.

    2013-01-01

    This paper evaluates UNESCO's recommended sexuality educational framework for junior school students aged 5-8 years. It also compares it to an existing state-designed Health and Physical Education curriculum that includes sexual and reproductive health for the same cohort. Based on the universal values of respect and human rights, UNESCO's"…

  15. International Guidelines on Sexuality Education and Their Relevance to a Contemporary Curriculum for Children Aged 5-8 Years

    Science.gov (United States)

    Goldman, Juliette D. G.

    2013-01-01

    This paper evaluates UNESCO's recommended sexuality educational framework for junior school students aged 5-8 years. It also compares it to an existing state-designed Health and Physical Education curriculum that includes sexual and reproductive health for the same cohort. Based on the universal values of respect and human rights, UNESCO's"…

  16. LMFBR type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kanbe, Mitsuru

    1997-04-04

    An LMFBR type reactor comprises a plurality of reactor cores in a reactor container. Namely, a plurality of pot containing vessels are disposed in the reactor vessel and a plurality of reactor cores are formed in a state where an integrated-type fuel assembly is each inserted to a pot, and a coolant pipeline is connected to each of the pot containing-vessel to cool the reactor core respectively. When fuels are exchanged, the integrated-type fuel assembly is taken out together with the pot from the reactor vessel in a state where the integrated-type fuel assembly is immersed in the coolants in the pot as it is. Accordingly, coolants are supplied to each of the pot containing-vessel connected with the coolant pipeline and circulate while cooling the integrated-type fuel assembly for every pot. Then, when the fuels are exchanged, the integrated type fuel assembly is taken out to the outside of the reactor together with the pot by taking up the pot from the pot-containing vessel. Then, neutron economy is improved to thereby improve reactor power and the breeding ratio. (N.H.)

  17. INVAP's Research Reactor Designs

    Directory of Open Access Journals (Sweden)

    Eduardo Villarino

    2011-01-01

    Full Text Available INVAP, an Argentine company founded more than three decades ago, is today recognized as one of the leaders within the research reactor industry. INVAP has participated in several projects covering a wide range of facilities, designed in accordance with the requirements of our different clients. For complying with these requirements, INVAP developed special skills and capabilities to deal with different fuel assemblies, different core cooling systems, and different reactor layouts. This paper summarizes the general features and utilization of several INVAP research reactor designs, from subcritical and critical assemblies to high-power reactors.

  18. Multi purpose research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Raina, V.K. [Research Reactor Design and Projects Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)]. E-mail: vkrain@magnum.barc.ernet.in; Sasidharan, K. [Research Reactor Design and Projects Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Sengupta, Samiran [Research Reactor Design and Projects Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Singh, Tej [Research Reactor Services Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)

    2006-04-15

    At present Dhruva and Cirus reactors provide the majority of research reactor based facilities to cater to the various needs of a vast pool of researchers in the field of material sciences, physics, chemistry, bio sciences, research and development work for nuclear power plants and production of radio isotopes. With a view to further consolidate and expand the scope of research and development in nuclear and allied sciences, a new 20 MWt multi purpose research reactor is being designed. This paper describes some of the design features and safety aspects of this reactor.

  19. Reactor and process design in sustainable energy technology

    CERN Document Server

    Shi, Fan

    2014-01-01

    Reactor Process Design in Sustainable Energy Technology compiles and explains current developments in reactor and process design in sustainable energy technologies, including optimization and scale-up methodologies and numerical methods. Sustainable energy technologies that require more efficient means of converting and utilizing energy can help provide for burgeoning global energy demand while reducing anthropogenic carbon dioxide emissions associated with energy production. The book, contributed by an international team of academic and industry experts in the field, brings numerous reactor design cases to readers based on their valuable experience from lab R&D scale to industry levels. It is the first to emphasize reactor engineering in sustainable energy technology discussing design. It provides comprehensive tools and information to help engineers and energy professionals learn, design, and specify chemical reactors and processes confidently. Emphasis on reactor engineering in sustainable energy techn...

  20. A reference worldwide model for antineutrinos from reactors

    CERN Document Server

    Baldoncini, Marica; Fiorentini, Giovanni; Mantovani, Fabio; Ricci, Barbara; Strati, Virginia; Xhixha, Gerti

    2014-01-01

    Antineutrinos produced at nuclear reactors constitute a severe source of background for the detection of geoneutrinos, which bring to the Earth's surface information about natural radioactivity in the whole planet. In this framework we provide a reference worldwide model for antineutrinos from reactors, in view of reactors operational records yearly published by the International Atomic Energy Agency (IAEA). We evaluate the expected signal from commercial reactors for ongoing (KamLAND and Borexino), planned (SNO+) and proposed (Juno, RENO-50, LENA and Hanohano) experimental sites. Uncertainties related to reactor antineutrino production, propagation and detection processes are estimated using a Monte Carlo based approach, which provides an overall site dependent uncertainty on the signal in the geoneutrino energy window on the order of 3%. We also implement the off-equilibrium correction to the reference reactor spectra associated with the long-lived isotopes and we estimate a 2.4% increase of the unoscillate...

  1. Impact of the age of Biomphalaria alexandrina snails on Schistosoma mansoni transmission: modulation of the genetic outcome and the internal defence system of the snail

    Directory of Open Access Journals (Sweden)

    Iman Fathy Abou-El-Naga

    2015-08-01

    Full Text Available Of the approximately 34 identified Biomphalariaspecies,Biomphalaria alexandrinarepresents the intermediate host of Schistosoma mansoniin Egypt. Using parasitological and SOD1 enzyme assay, this study aimed to elucidate the impact of the age of B. alexandrinasnails on their genetic variability and internal defence against S. mansoniinfection. Susceptible and resistant snails were reared individually for self-reproduction; four subgroups of their progeny were used in experiment. The young susceptible subgroup showed the highest infection rate, the shortest pre-patent period, the highest total cercarial production, the highest mortality rate and the lowest SOD1 activity. Among the young and adult susceptible subgroups, 8% and 26% were found to be resistant, indicating the inheritance of resistance alleles from parents. The adult resistant subgroup, however, contained only resistant snails and showed the highest enzyme activity. The complex interaction between snail age, genetic background and internal defence resulted in great variability in compatibility patterns, with the highest significant difference between young susceptible and adult resistant snails. The results demonstrate that resistance alleles function to a greater degree in adults, with higher SOD1 activity and provide potential implications for Biomphalariacontrol. The identification of the most susceptible snail age enables determination of the best timing for applying molluscicides. Moreover, adult resistant snails could be beneficial in biological snail control.

  2. Sodium Reactor Experiment decommissioning. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Carroll, J.W.; Conners, C.C.; Harris, J.M.; Marzec, J.M.; Ureda, B.F.

    1983-08-15

    The Sodium Reactor Experiment (SRE) located at the Rockwell International Field Laboratories northwest of Los Angeles was developed to demonstrate a sodium-cooled, graphite-moderated reactor for civilian use. The reactor reached full power in May 1958 and provided 37 GWh to the Southern California Edison Company grid before it was shut down in 1967. Decommissioning of the SRE began in 1974 with the objective of removing all significant radioactivity from the site and releasing the facility for unrestricted use. Planning documentation was prepared to describe in detail the equipment and techniques development and the decommissioning work scope. A plasma-arc manipulator was developed for remotely dissecting the highly radioactive reactor vessels. Other important developments included techniques for using explosives to cut reactor vessel internal piping, clamps, and brackets; decontaminating porous concrete surfaces; and disposing of massive equipment and structures. The documentation defined the decommissioning in an SRE dismantling plan, in activity requirements for elements of the decommissioning work scope, and in detailed procedures for each major task.

  3. Mental health problems and resilience in international adoptees: Results from a population-based study of Norwegian adolescents aged 16-19 years.

    Science.gov (United States)

    Askeland, Kristin Gärtner; Hysing, Mari; Aarø, Leif Edvard; Tell, Grethe S; Sivertsen, Børge

    2015-10-01

    The aim of the study was to investigate mental health and resilience in adolescents who have been internationally adopted and their non-adopted peers and examine the potential interaction between adoption status and resilience on mental health problems. Data from the population based youth@hordaland-survey, conducted in Hordaland County, Norway, in 2012 was used. In all, 10 257 adolescents aged 16-19 years provided self-reported data on several mental health instruments. Of these, 45 adolescents were identified as internationally adopted. Adoptees reported more symptoms of depression, attention-deficit/hyperactivity disorder (ADHD), obsessive compulsive disorder (OCD) and perfectionism than non-adopted adolescents, but there were no differences regarding resilience. Adolescents with higher resilience scores reported fewer symptoms of mental health problems, however, no interaction effects were found for adoption status and total resilience score on measures of mental health problems. Our findings indicate that knowledge of resilience factors can form the basis for preventive interventions.

  4. Confessional Lutheran commitment in the International Lutheran Council – A conservative contribution of Lutheranism to the Ecumenical Age

    Directory of Open Access Journals (Sweden)

    Werner R.A. Klän

    2013-09-01

    Full Text Available The contribution of confessional Lutheran churches, especially those affiliated to the International Lutheran Council of the ecumenical movement was regarded more or less as marginal, compared to the mainstream Protestant churches. Rooted in the 16th century Reformation, relating to the confessional writings of the Lutheran Church as comprised in the Book of Concord (1580, these churches in the 19th century rediscovered what might be labelled ‘confessional identity’. Looking at the European scene as a paradigm of secularisation (in spite of necessary differentiations, it is observed how traditional faith, trying not to sever its biblical and confessional roots, approached and reacted to ‘modern’ developments in society and the church. A historical survey, combined with a systematic reflection on Lutheran identity in a post-Christian context, served to diagnose the problems of Christian responsibility in a globalising world. Through the changes and challenges that confront Christianity at the beginning of the 21st century, the confessional Lutheran churches – affiliated to the International Lutheran Council – came to face their ecumenical responsibility. The mission of the Church ought to be reconsidered in terms of its biblical foundation, its historical identity, its confessional self-understanding, and its ecumenical obligation.

  5. The inner containment of an EPR trademark pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ostermann, Dirk; Krumb, Christian; Wienand, Burkhard [AREVA GmbH, Offenbach (Germany)

    2014-08-15

    On February 12, 2014 the containment pressure and subsequent leak tightness tests on the containment of the Finnish Olkiluoto 3 EPR trademark reactor building were completed successfully. The containment of an EPR trademark pressurized water reactor consists of an outer containment to protect the reactor building against external hazards (such as airplane crash) and of an inner containment that is subjected to internal overpressure and high temperature in case of internal accidents. The current paper gives an overview of the containment structure, the design criteria, the validation by analyses and experiments and the containment pressure test.

  6. Characterization of the performances of an innovative heat-exchanger/reactor

    OpenAIRE

    Théron, Felicie; Anxionnaz-Minvielle, Zoé; Cabassud, Michel; Gourdon, Christophe; Tochon, Patrice

    2014-01-01

    International audience; The use of heat exchanger/reactors (HEX/reactors) is a promising way to overcome the barrier of poor heat transfer in batch reactors. However to reach residence time long enough to complete the chemistry,low Reynolds number has to be combined with both a plug flow behaviour and the intensification of heat and mass transfers. This work concerns the experimental approach used to characterize an innovative HEX/reactor. The pilot is made of three process plates sandwiched ...

  7. Pressure drop and axial dispersion in industrial millistructured heat exchange reactors

    OpenAIRE

    Moreau, Maxime; Di Miceli Raimondi, Nathalie; Le Sauze, Nathalie; Cabassud, Michel; Gourdon, Christophe

    2015-01-01

    International audience; Hydrodynamic characterization by means of pressure drop and residence time distribution (RTD)experiments is performed in three millistructured heat exchange reactors: two Corning reactors (further referred to as Corning HP and Corning RT) and a Chart reactor. Pressure drop is measured for different flow rates and fluids. Fanning friction factor is then calculated and its evolution versus Reynolds number is plotted for each reactor, showing the influence of the geometri...

  8. Investigation of Isfahan miniature neutron source reactor (MNSR) for boron neutron capture therapy by MCNP simulation

    OpenAIRE

    S. Z. Kalantari; H Tavakoli; Nami, M.

    2015-01-01

    One of the important neutron sources for Boron Neutron Capture Therapy (BNCT) is a nuclear reactor. It needs a high flux of epithermal neutrons. The optimum conditions of the neutron spectra for BNCT are provided by the International Atomic Energy Agency (IAEA). In this paper, Miniature Neutron Source Reactor (MNSR) as a neutron source for BNCT was investigated. For this purpose, we designed a Beam Shaping Assembly (BSA) for the reactor and the neutron transport from the core of the reactor t...

  9. Age-related patterns of vigorous-intensity physical activity in youth: The International Children's Accelerometry Database

    Directory of Open Access Journals (Sweden)

    Kirsten Corder

    2016-12-01

    Age-related declines in vigorous-intensity activity during youth appear relatively greater than those of moderate activity. However, due to a higher baseline, absolute moderate-intensity activity decreases more than vigorous. Overweight/obese individuals, girls, and North Americans appear especially in need of vigorous-intensity activity promotion due to low levels at 5.0–5.9 y and larger negative annual differences.

  10. Report form the 14th International ECHA conference: Re:thinking giftedness - giftedness in the digital age

    OpenAIRE

    Juriševič, Mojca

    2014-01-01

    Between the 14th and 20th September 2014, the city of Ljubljana, Slovenia, hosted the 14th ECHA Conference under the title Re:thinking Giftedness – giftedness in the digital age. The conference was organised by ECHA (i.e. European Council for High Ability) in cooperation with two local partners, the Faculty of Education of the University of Ljubljana and the company MIB d.o.o.

  11. Report form the 14th International ECHA conference: Re:thinking giftedness - giftedness in the digital age

    OpenAIRE

    Juriševič, Mojca

    2014-01-01

    Between the 14th and 20th September 2014, the city of Ljubljana, Slovenia, hosted the 14th ECHA Conference under the title Re:thinking Giftedness – giftedness in the digital age. The conference was organised by ECHA (i.e. European Council for High Ability) in cooperation with two local partners, the Faculty of Education of the University of Ljubljana and the company MIB d.o.o.

  12. A summary of the Proceedings of the Twelfth International Symposium on the Neurobiology and Neuroendocrinology of Aging, Bregenz, Austria July 27-August 1, 2014.

    Science.gov (United States)

    Brown-Borg, Holly M; Borg, Kurt E

    2015-08-01

    A summary of the Twelfth International Symposium on the Neurobiology and Neuroendocrinology of Aging that was held July 27-August 1, 2014 in Bregenz, Austria, is presented. Fifteen of the speakers that presented at the conference submitted review papers covering the topic of their presentation as well as an overview of their respective fields and are included in this special issue. The abstracts from each poster presentation as well as seven of the speakers' abstracts are also included at the end of the preface to the special issue.

  13. Review of Andrés Solimano, International Migration in the Age of Crisis and Globalisation. Historical and Recent Experiences, Cambridge University Press, 2010. 223 Pp.

    Directory of Open Access Journals (Sweden)

    Dirina Claudiu – Ciprian

    2015-12-01

    Full Text Available “International Migration in the Age of Crisis and Globalization. Historical and Recent Experiences” represents a work of major interest in the field of migration and globalization. Apparently two concepts that relate one to another, on a background of major population dynamics, the two notions are the two important pillars in what might perfectly be described as a complex analysis of migration, starting from the elements that have initiated it, and culminating with a fine comparison of positive and negative aspects of this phenomenon.

  14. International Atomic Energy Agency specialists meeting on experience in ageing, maintenance, and modernization of instrumentation and control systems for improving nuclear power plant availability

    Energy Technology Data Exchange (ETDEWEB)

    1993-10-01

    This report presents the proceedings of the Specialist`s Meeting on Experience in Aging, Maintenance and Modernization of Instrumentation and Control Systems for Improving Nuclear Power Plant Availability that was held at the Ramada Inn in Rockville, Maryland on May 5--7, 1993. The Meeting was presented in cooperation with the Electric Power Research Institute, Oak Ridge National Laboratory and the International Atomic Energy Agency. There were approximately 65 participants from 13 countries at the Meeting. Individual reports have been cataloged separately.

  15. International Inequality in the Age of Globalization: Japanese Economic Ascent and the Restructuring of the Capitalist World-Economy

    Directory of Open Access Journals (Sweden)

    Paul S. Ciccantell

    2015-08-01

    Full Text Available This paper shows how Japanese ?rms and the Japanese state constructed a development model based on the steel industry as a generative sector that drove Japan’s economic ascent in the world-historical context of U.S. hegemony. We make three arguments in this paper. First, there is a new model of capital accumulation that does create new forms of social inequality by redistributing costs and bene?ts in very di?erent ways than earlier models. Second, Japanese ?rms and the Japanese state created this new model of capital accumulation and social inequality via mechanisms including joint ventures, long term contracts, and other forms of international trade and investment, not U.S.-based transnational corporations, as is usually assumed. Third, world-systems theory reconstructed through the lens of the new historical materialism explains this restructuring of the capitalist world-economy as the outcome of Japan’s economic ascent over the last ?fty years. Further, we argue that this new model of capital accumulation has had similar impacts on redistributing the costs and bene?ts of development between core and peripheral regions of the capitalist world-economy in a wide range of global industries. These strategies created a tightly linked set of technological and organizational innovations to overcome the natural and social obstacles to Japanese development, dramatically increase Japan’s international economic competitiveness by lowering production costs in all sectors of the economy, turn Japan into the world’s largest exporter of manufactured products, restructure a range of global industries, and recreate the world-system hierarchy in support of Japanese development. In particular, organizational inno-vations in the use of long term contracts and joint ventures in raw materials industries to foster global excess capacity and lower rents to resource extracting ?rms and states reallocated the costs of providing the material building blocks of

  16. International research on the civic engagement of the youth and adolescents. Young citizens in the digital age

    Directory of Open Access Journals (Sweden)

    María Carmen Robles Vílchez

    2011-08-01

    Full Text Available This article presents a review of some of the research in the international context aimed at unveiling and valuing the experience of young people in experiences and beliefs regarding citizenship (Burke, 2007. Such research is generally intended to obtain information on students’ knowledge, attitudes and civic behavior in their own schools. We will highlight throughout the present paper that previous studies and the current state of scientific knowledge in civic education focuses on purely formal and structural aspects and lack a deep understanding and interpretation of the experiences of young citizens. We conclude, after our review, that it is necessary to include more comprehensive factors in the research and enquiry into the experiences and civic training of the youth.

  17. Reactor Materials Research

    Energy Technology Data Exchange (ETDEWEB)

    Van Walle, E

    2001-04-01

    The activities of the Reactor Materials Research Department of the Belgian Nuclear Research Centre SCK-CEN in 2000 are summarised. The programmes within the department are focussed on studies concerning (1) fusion, in particular mechanical testing; (2) Irradiation Assisted Stress Corrosion Cracking (IASCC); (3) nuclear fuel; and (4) Reactor Pressure Vessel Steel (RPVS)

  18. Space Nuclear Reactor Engineering

    Energy Technology Data Exchange (ETDEWEB)

    Poston, David Irvin [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-03-06

    We needed to find a space reactor concept that could be attractive to NASA for flight and proven with a rapid turnaround, low-cost nuclear test. Heat-pipe-cooled reactors coupled to Stirling engines long identified as the easiest path to near-term, low-cost concept.

  19. Light water reactor program

    Energy Technology Data Exchange (ETDEWEB)

    Franks, S.M.

    1994-12-31

    The US Department of Energy`s Light Water Reactor Program is outlined. The scope of the program consists of: design certification of evolutionary plants; design, development, and design certification of simplified passive plants; first-of-a-kind engineering to achieve commercial standardization; plant lifetime improvement; and advanced reactor severe accident program. These program activities of the Office of Nuclear Energy are discussed.

  20. Transportation and storage of foreign spent power reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    1979-09-30

    This report describes the generic actions to be taken by the Department of Energy, in cooperation with other US government agencies, foreign governments, and international organizations, in support of the implementation of Administration policies with respect to the following international spent fuel management activities: bilateral cooperation related to expansion of foreign national storage capacities; multilateral and international cooperation related to development of multinational and international spent fuel storage regimes; fee-based transfer of foreign spent power reactor fuel to the US for storage; and emergency transfer of foreign spent power reactor fuel to the US for storage.

  1. Status of French reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ballagny, A. [Commissariat a l`Energie Atomique, Saclay (France)

    1997-08-01

    The status of French reactors is reviewed. The ORPHEE and RHF reactors can not be operated with a LEU fuel which would be limited to 4.8 g U/cm{sup 3}. The OSIRIS reactor has already been converted to LEU. It will use U{sub 3}Si{sub 2} as soon as its present stock of UO{sub 2} fuel is used up, at the end of 1994. The decision to close down the SILOE reactor in the near future is not propitious for the start of a conversion process. The REX 2000 reactor, which is expected to be commissioned in 2005, will use LEU (except if the fast neutrons core option is selected). Concerning the end of the HEU fuel cycle, the best option is reprocessing followed by conversion of the reprocessed uranium to LEU.

  2. Nuclear reactor design

    CERN Document Server

    2014-01-01

    This book focuses on core design and methods for design and analysis. It is based on advances made in nuclear power utilization and computational methods over the past 40 years, covering core design of boiling water reactors and pressurized water reactors, as well as fast reactors and high-temperature gas-cooled reactors. The objectives of this book are to help graduate and advanced undergraduate students to understand core design and analysis, and to serve as a background reference for engineers actively working in light water reactors. Methodologies for core design and analysis, together with physical descriptions, are emphasized. The book also covers coupled thermal hydraulic core calculations, plant dynamics, and safety analysis, allowing readers to understand core design in relation to plant control and safety.

  3. Mirror reactor surface study

    Energy Technology Data Exchange (ETDEWEB)

    Hunt, A. L.; Damm, C. C.; Futch, A. H.; Hiskes, J. R.; Meisenheimer, R. G.; Moir, R. W.; Simonen, T. C.; Stallard, B. W.; Taylor, C. E.

    1976-09-01

    A general survey is presented of surface-related phenomena associated with the following mirror reactor elements: plasma first wall, ion sources, neutral beams, director converters, vacuum systems, and plasma diagnostics. A discussion of surface phenomena in possible abnormal reactor operation is included. Several studies which appear to merit immediate attention and which are essential to the development of mirror reactors are abstracted from the list of recommended areas for surface work. The appendix contains a discussion of the fundamentals of particle/surface interactions. The interactions surveyed are backscattering, thermal desorption, sputtering, diffusion, particle ranges in solids, and surface spectroscopic methods. A bibliography lists references in a number of categories pertinent to mirror reactors. Several complete published and unpublished reports on surface aspects of current mirror plasma experiments and reactor developments are also included.

  4. Current Abstracts Nuclear Reactors and Technology

    Energy Technology Data Exchange (ETDEWEB)

    Bales, J.D.; Hicks, S.C. [eds.

    1993-01-01

    This publication Nuclear Reactors and Technology (NRT) announces on a monthly basis the current worldwide information available from the open literature on nuclear reactors and technology, including all aspects of power reactors, components and accessories, fuel elements, control systems, and materials. This publication contains the abstracts of DOE reports, journal articles, conference papers, patents, theses, and monographs added to the Energy Science and Technology Database during the past month. Also included are US information obtained through acquisition programs or interagency agreements and international information obtained through acquisition programs or interagency agreements and international information obtained through the International Energy Agency`s Energy Technology Data Exchange or government-to-government agreements. The digests in NRT and other citations to information on nuclear reactors back to 1948 are available for online searching and retrieval on the Energy Science and Technology Database and Nuclear Science Abstracts (NSA) database. Current information, added daily to the Energy Science and Technology Database, is available to DOE and its contractors through the DOE Integrated Technical Information System. Customized profiles can be developed to provide current information to meet each user`s needs.

  5. Prevalence and Associated Factors of Secondhand Smoke Exposure among Internal Chinese Migrant Women of Reproductive Age: Evidence from China's Labor-Force Dynamic Survey.

    Science.gov (United States)

    Gong, Xiao; Luo, Xiaofeng; Ling, Li

    2016-04-01

    Secondhand smoke (SHS) is a major risk factor for poor health outcomes among women in China, where proportionately few women smoke. This is especially the case as it pertains to women's reproductive health, specifically migrant women who are exposed to SHS more than the population at large. There are several factors which may increase migrant women's risk of SHS exposure. This paper aims to investigate the prevalence and associated factors of SHS exposure among internal Chinese migrant women of reproductive age. The data used were derived from the 2014 Chinese Labor Dynamic Survey, a national representative panel survey. The age-adjusted rate of SHS exposure of women of reproductive age with migration experience was of 43.46% (95% CI: 40.73%-46.40%), higher than those without migration experience (35.28% (95% CI: 33.66%-36.97%)). Multivariate analysis showed that participants with a marital status of "Widowed" had statistically lower exposure rates, while those with a status of "Cohabitation" had statistically higher exposure. Those with an undergraduate degree or above had statistically lower SHS exposure. Those with increasing levels of social support, and those who currently smoke or drink alcohol, had statistically higher SHS exposure. Participants' different work-places had an effect on their SHS exposure, with outdoor workers statistically more exposed. Our findings suggest that urgent tobacco control measures should be taken to reduce smoking prevalence and SHS exposure. Specific attention should be paid to protecting migrant women of reproductive age from SHS.

  6. International Research Project on the Effects of Chemical Ageing of Polymers on Performance Properties: Chemical and Thermal Analysis

    Science.gov (United States)

    Bulluck, J. W.; Rushing, R. A.

    1996-01-01

    Work during the past six months has included significant research in several areas aimed at further clarification of the aging and chemical failure mechanism of thermoplastics (PVDF or Tefzel) pipes. Among the areas investigated were the crystallinity changes associated with both the Coflon and Tefzel after various simulated environmental exposures using X-ray diffraction analysis. We have found that significant changes in polymer crystallinity levels occur as a function of the exposures. These crystallinity changes may have important consequences on the fracture, fatigue, tensile, and chemical resistance of the materials. We have also noted small changes in the molecular weight distribution. Again these changes may result in variations in the mechanical and chemical properties in the material. We conducted numerous analytical studies with methods including X-ray Diffraction, Gel Permeation Chromatography, Fourier Transform Infrared Spectroscopy, Ultra- Violet Scanning Analysis, GC/Mass Spectrometry, Differential Scanning Calorimetry and Thermomechanical Analysis. In the ultra-violet analysis we noted the presence of an absorption band indicative of triene formation. We investigated a number of aged samples of both Tefzel and Coflon that were forwarded from MERL. We also cast films at SWT and subjected these films to a refluxing methanol 1% ethylene diamine solution. An updated literature search was conducted using Dialog and DROLLS to identify any new papers that may have been published in the open literature since the start of this project. The updated literature search and abstracts are contained in the Appendix section of this report.

  7. Participation frequency and perceived participation restrictions at older age: applying the International Classification of Functioning, Disability and Health (ICF) framework.

    Science.gov (United States)

    Arnadottir, Solveig A; Gunnarsdottir, Elin D; Stenlund, Hans; Lundin-Olsson, Lillemor

    2011-01-01

    To identify variables from different components of International Classification of Functioning, Disability and Health (ICF) associated with older people's participation frequency and perceived participation restrictions. Participants (N = 186) were community-living, 65-88 years old and 52% men. The dependent variables, participation frequency (linear regression) and perceived participation restrictions (logistic regression), were measured using The Late-Life Function and Disability Instrument. Independent variables were selected from various ICF components. Higher participation frequency was associated with living in urban rather than rural community (β = 2.8, p perceived participation restrictions (adjusted odds ratio [OR] = 5.5, p = 0.001; OR = 1.09, p perceived participation restriction decreased as depressive symptoms increased (OR = 0.8, p = 0.011). Our results highlight the importance of capturing and understanding both frequency and restriction aspects of older persons' participation. ICF may be a helpful reference to map factors associated with participation and to study further potentially modifiable influencing factors such as depressive symptoms and advanced lower extremity capacity.

  8. Literature study of the radiobiological parameters of Caesium-137 required for evaluating internal irradiation doses as a function of age; Etude bibliographique des parametres radiobiologiques du cesium-137 necessaires a l'evaluation des doses d'irradiation interne en fonction de l'age

    Energy Technology Data Exchange (ETDEWEB)

    Garnier, A. [Commissariat a l' Energie Atomique, 92 - Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1968-07-01

    This document reassembles information published in scientific literature on radiobiological parameters of Cs-137, necessary for the estimate of the internal irradiation dose of man according to his age (during growth). The data are completed by a commented review of the mathematical models, proposed in order to value the irradiation doses from ingested cesium and the biological parameters. (author) [French] Ce document rassemble les informations publiees dans la litterature scientifique, concernant les parametres radiobiologiqueo du cesium-137, necessaires a l'evaluation des doses d'irradiation interne de l'homme en fonction de l'age. Ces donnees sont completees par une revue commentee des modeles mathematiques proposes en vue de l'evaluation des doses d'irradiation a partir des quantites de cesium ingerees et des parametres biologiques. (auteur)

  9. Multiplicity features of adiabatic autothermal reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lovo, M.; Balakotaiah, V. (Houston Univ., TX (United States). Dept. of Chemical Engineering)

    1992-01-01

    In this paper singularity theory, large activation energy asymptotic, and numerical methods are used to present a comprehensive study of the steady-state multiplicity features of three classical adiabatic autothermal reactor models: tubular reactor with internal heat exchange, tubular reactor with external heat exchange, and the CSTR with external heat exchange. Specifically, the authors derive the exact uniqueness-multiplicity boundary, determine typical cross-sections of the bifurcation set, and classify the different types of bifurcation diagrams of conversion vs. residence time. Asymptotic (limiting) models are used to determine analytical expressions for the uniqueness boundary and the ignition and extinction points. The analytical results are used to present simple, explicit and accurate expressions defining the boundary of the region of autothermal operation in the physical parameter space.

  10. Slurry reactor design studies

    Energy Technology Data Exchange (ETDEWEB)

    Fox, J.M.; Degen, B.D.; Cady, G.; Deslate, F.D.; Summers, R.L. (Bechtel Group, Inc., San Francisco, CA (USA)); Akgerman, A. (Texas A and M Univ., College Station, TX (USA)); Smith, J.M. (California Univ., Davis, CA (USA))

    1990-06-01

    The objective of these studies was to perform a realistic evaluation of the relative costs of tublar-fixed-bed and slurry reactors for methanol, mixed alcohols and Fischer-Tropsch syntheses under conditions where they would realistically be expected to operate. The slurry Fischer-Tropsch reactor was, therefore, operated at low H{sub 2}/CO ratio on gas directly from a Shell gasifier. The fixed-bed reactor was operated on 2.0 H{sub 2}/CO ratio gas after adjustment by shift and CO{sub 2} removal. Every attempt was made to give each reactor the benefit of its optimum design condition and correlations were developed to extend the models beyond the range of the experimental pilot plant data. For the methanol design, comparisons were made for a recycle plant with high methanol yield, this being the standard design condition. It is recognized that this is not necessarily the optimum application for the slurry reactor, which is being proposed for a once-through operation, coproducing methanol and power. Consideration is also given to the applicability of the slurry reactor to mixed alcohols, based on conditions provided by Lurgi for an Octamix{trademark} plant using their standard tubular-fixed reactor technology. 7 figs., 26 tabs.

  11. Transit time distributions and StorAge Selection functions in a sloping soil lysimeter with time-varying flow paths: Direct observation of internal and external transport variability

    Science.gov (United States)

    Kim, Minseok; Pangle, Luke A.; Cardoso, Charléne; Lora, Marco; Volkmann, Till H. M.; Wang, Yadi; Harman, Ciaran J.; Troch, Peter A.

    2016-09-01

    Transit times through hydrologic systems vary in time, but the nature of that variability is not well understood. Transit times variability was investigated in a 1 m3 sloping lysimeter, representing a simplified model of a hillslope receiving periodic rainfall events for 28 days. Tracer tests were conducted using an experimental protocol that allows time-variable transit time distributions (TTDs) to be calculated from data. Observed TTDs varied with the storage state of the system, and the history of inflows and outflows. We propose that the observed time variability of the TTDs can be decomposed into two parts: "internal" variability associated with changes in the arrangement of, and partitioning between, flow pathways; and "external" variability driven by fluctuations in the flow rate along all flow pathways. These concepts can be defined quantitatively in terms of rank StorAge Selection (rSAS) functions, which is a theory describing lumped transport dynamics. Internal variability is associated with temporal variability in the rSAS function, while external is not. The rSAS function variability was characterized by an "inverse storage effect," whereby younger water is released in greater proportion under wetter conditions than drier. We hypothesize that this effect is caused by the rapid mobilization of water in the unsaturated zone by the rising water table. Common approximations used to model transport dynamics that neglect internal variability were unable to reproduce the observed breakthrough curves accurately. This suggests that internal variability can play an important role in hydrologic transport dynamics, with implications for field data interpretation and modeling.

  12. Does the impact of osteoarthritis vary by age, gender and social deprivation? A community study using the International Classification of Functioning, Disability and Health.

    Science.gov (United States)

    Pollard, Beth; Dixon, Diane; Johnston, Marie

    2014-01-01

    Abstract Purpose: The aim of the study was to explore if the impact of osteoarthritis varies with respect to age, gender and social deprivation. Impact was defined as impairment, activity limitations and participation restriction (International Classification of Functioning, Disability and Health (ICF)). Investigating the functioning of the ICF model for subgroups is important both practically and theoretically. The sample comprised a community sample of 763 people diagnosed with osteoarthritis. Uncontaminated measures of the ICF constructs were developed using discriminant content validity from a pool of 134 items, including the WOMAC and SF-36. Multigroup Structural Equation Modelling was used to explore if the same pathways exist for subgroups of gender, age and social deprivation. Different significant paths were found for gender and social deprivation: impairment did not predict participation restriction for women and those most deprived, whereas these paths were significant for men and those less deprived. No difference in the paths was found for age. The impact of osteoarthritis appears to vary with respect to gender and social deprivation but not age. This suggests both that osteoarthritis per se does not adequately explain the health outcomes observed and that different clinical approaches may be appropriate for people of different gender and levels of deprivation. Implications of Rehabilitation The ICF model appears to vary with respect to gender and social deprivation for people with osteoarthritis. The ICF model did not appear to vary with respect to age for people with osteoarthritis. Different treatments and interventions for osteoarthritis may need to be targeted for specific gender and social deprivation groups.

  13. Further evidence for an association of ABCR alleles with age-related macular degeneration. The International ABCR Screening Consortium.

    Science.gov (United States)

    Allikmets, R

    2000-08-01

    Age-related macular degeneration (AMD) accounts for >50% of the registered visual disability among North American and Western European populations and has been associated both with environmental factors, such as smoking, and with genetic factors. Previously we have reported disease-associated variants in the ABCR (also called ABCA4) gene in a subset of patients affected with this complex disorder. We have now tested our original hypothesis, that ABCR is a dominant susceptibility locus for AMD, by screening 1,218 unrelated AMD patients of North American and Western European origin and 1,258 comparison individuals from 15 centers in North America and Europe for the two most frequent AMD-associated variants found in ABCR. These two sequence changes, G1961E and D2177N, were found in one allele of ABCR in 40 patients ( approximately 3.4%), and in 13 control subjects ( approximately 0.95%). Fisher's two-sided exact test confirmed that these two variants are associated with AMD at a statistically significant level (PAMD is elevated approximately threefold in D2177N carriers and approximately fivefold in G1961E carriers. The identification of a gene that confers risk of AMD is an important step in unraveling this complex disorder.

  14. Climate change threatens archaeologically significant ice patches: insights into their age, internal structure, mass balance and climate sensitivity

    Science.gov (United States)

    Strand Ødegård, Rune; Nesje, Atle; Isaksen, Ketil; Andreassen, Liss Marie; Eiken, Trond; Schwikowski, Margit; Uglietti, Chiara

    2017-01-01

    Despite numerous spectacular archaeological discoveries worldwide related to melting ice patches and the emerging field of glacial archaeology, governing processes related to ice patch development during the Holocene and their sensitivity to climate change are still largely unexplored. Here we present new results from an extensive 6-year (2009-2015) field experiment at the Juvfonne ice patch in Jotunheimen in central southern Norway. Our results show that the ice patch has existed continuously since the late Mesolithic period. Organic-rich layers and carbonaceous aerosols embedded in clear ice show ages spanning from modern at the surface to ca. 7600 cal years BP at the bottom. This is the oldest dating of ice in mainland Norway. The expanding ice patch covered moss mats appearing along the margin of Juvfonne about 2000 years ago. During the study period, the mass balance record showed a strong negative balance, and the annual balance is highly asymmetric over short distances. Snow accumulation is poorly correlated with estimated winter precipitation, and single storm events may contribute significantly to the total winter balance. Snow accumulation is approx. 20 % higher in the frontal area compared to the upper central part of the ice patch. There is sufficient meltwater to bring the permeable snowpack to an isothermal state within a few weeks in early summer. Below the seasonal snowpack, ice temperatures are between -2 and -4 °C. Juvfonne has clear ice stratification of isochronic origin.

  15. Gas cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1972-06-01

    Although most of the development work on fast breeder reactors has been devoted to the use of liquid metal cooling, interest has been expressed for a number of years in alternative breeder concepts using other coolants. One of a number of concepts in which interest has been retained is the Gas-Cooled Fast Reactor (GCFR). As presently envisioned, it would operate on the uranium-plutonium mixed oxide fuel cycle, similar to that used in the Liquid Metal Fast Breeder Reactor (LMFBR), and would use helium gas as the coolant.

  16. Microfluidic electrochemical reactors

    Science.gov (United States)

    Nuzzo, Ralph G [Champaign, IL; Mitrovski, Svetlana M [Urbana, IL

    2011-03-22

    A microfluidic electrochemical reactor includes an electrode and one or more microfluidic channels on the electrode, where the microfluidic channels are covered with a membrane containing a gas permeable polymer. The distance between the electrode and the membrane is less than 500 micrometers. The microfluidic electrochemical reactor can provide for increased reaction rates in electrochemical reactions using a gaseous reactant, as compared to conventional electrochemical cells. Microfluidic electrochemical reactors can be incorporated into devices for applications such as fuel cells, electrochemical analysis, microfluidic actuation, pH gradient formation.

  17. Nonpharmacological Interventions Targeted at Delirium Risk Factors, Delivered by Trained Volunteers (Medical and Psychology Students), Reduced Need for Antipsychotic Medications and the Length of Hospital Stay in Aged Patients Admitted to an Acute Internal Medicine Ward: Pilot Study.

    Science.gov (United States)

    Gorski, Stanislaw; Piotrowicz, Karolina; Rewiuk, Krzysztof; Halicka, Monika; Kalwak, Weronika; Rybak, Paulina; Grodzicki, Tomasz

    2017-01-01

    Purpose. Effectiveness of nonpharmacological multicomponent prevention delivered by trained volunteers (medical and psychology students), targeted at delirium risk factors in geriatric inpatients, was assessed at an internal medicine ward in Poland. Patients and Methods. Participants were recruited to intervention and control groups at the internal medicine ward (inclusion criteria: age ≥ 75, acute medical condition, basic orientation, and logical contact on admission; exclusion criteria: life expectancy internal medicine ward.

  18. Socio-demographic characteristics and factors associated with hospitalization in psychiatry of old age patients: an international comparison between Ireland and Turkey.

    Science.gov (United States)

    Carpar, Elif; McCarthy, Geraldine; Adamis, Dimitrios; Donmezler, Gizem; Cesur, Ender; Fistikci, Nurhan

    2017-08-14

    Taking predictors of hospitalization characteristics into consideration internationally would broaden our understanding of this population on a local basis. We aimed to examine and compare socio-demographic profiles along with hospitalization characteristics including length of hospital stay (LOS), reasons for admission and diagnoses among older adult inpatients hospitalized in Ireland and Turkey, and to assess factors predicting these features. The admission charts of 356 psychiatric inpatients over 65 years of age who were admitted to two different acute psychiatric hospitals (Sligo/Ireland and Istanbul/Turkey) were analysed by means of descriptive modalities and logistic regression. There were significant differences in several domains of socio-demographics, reasons of admission and diagnoses. LOS was significantly longer in Ireland. Living alone was the only significant predictor for longer LOS in both countries, whereas in addition to living alone, younger age was also a contributor for longer LOS in Turkey. Given that the only factor predicting LOS both in Turkey and Ireland was living alone, helping to identify more acceptable ways of providing social support for living arrangements constitutes an important service to shorten LOS in old age psychiatric population. It is possible to infer that independent from the cultural diversities, living arrangement is a consistent entity to influence length of hospital stay in older adult population.

  19. Radiation Damage In Reactor Cavity Concrete

    Energy Technology Data Exchange (ETDEWEB)

    Field, Kevin G [ORNL; Le Pape, Yann [ORNL; Naus, Dan J [ORNL; Remec, Igor [ORNL; Busby, Jeremy T [ORNL; Rosseel, Thomas M [ORNL; Wall, Dr. James Joseph [Electric Power Research Institute (EPRI)

    2015-01-01

    License renewal up to 60 years and the possibility of subsequent license renewal to 80 years has established a renewed focus on long-term aging of nuclear generating stations materials, and recently, on concrete. Large irreplaceable sections of most nuclear generating stations include concrete. The Expanded Materials Degradation Analysis (EMDA), jointly performed by the Department of Energy, the Nuclear Regulatory Commission and Industry, identified the urgent need to develop a consistent knowledge base on irradiation effects in concrete. Much of the historical mechanical performance data of irradiated concrete does not accurately reflect typical radiation conditions in NPPs or conditions out to 60 or 80 years of radiation exposure. To address these potential gaps in the knowledge base, The Electric Power Research Institute and Oak Ridge National Laboratory are working to disposition radiation damage as a degradation mechanism. This paper outlines the research program within this pathway including: (i) defining the upper bound of the neutron and gamma dose levels expected in the biological shield concrete for extended operation (80 years of operation and beyond), (ii) determining the effects of neutron and gamma irradiation as well as extended time at temperature on concrete, (iii) evaluating opportunities to irradiate prototypical concrete under accelerated neutron and gamma dose levels to establish a conservative bound and share data obtained from different flux, temperature, and fluence levels, (iv) evaluating opportunities to harvest and test irradiated concrete from international NPPs, (v) developing cooperative test programs to improve confidence in the results from the various concretes and research reactors, (vi) furthering the understanding of the effects of radiation on concrete (see companion paper) and (vii) establishing an international collaborative research and information exchange effort to leverage capabilities and knowledge.

  20. Preparation Before Signature of Upgrade of Algeria Heavy Water Research Reactor Contract

    Institute of Scientific and Technical Information of China (English)

    LI; Song; ZAN; Huai-qi; XU; Qi-guo; JIA; Yu-wen

    2012-01-01

    <正>Algeria heavy water research reactor (Birine) is a multiple-purpose research reactor, which was constructed with the help of China more than 20 years ago. By request of Algeria, China will upgrade the research reactor; so as to improve the status of current reactor such as equipment ageing, shortage of spare parts, several systems do not meet requirements of current standards and criteria etc.

  1. Preliminary hazards review overboring Hanford reactors

    Energy Technology Data Exchange (ETDEWEB)

    Nilson, R.; Carlson, P.A.

    1962-07-25

    The General Electric Company, as prime contractor to the AEC at Hanford, is proposing to modify the lattice characteristics of the 8 3/8-inch lattice reactors for the purposes of improving the conversion ratio of these reactors. The proposed overbore modification of the reactors would remove the existing aluminum process tubes, enlarge the diameters of the graphite channels by about one-half inch, insert smooth-bore Zircaloy-2 process tubes and refuel the reactor with larger size, self-supported fuel elements. The overbore fuel will remain the internally-and-externally-cooled cylindrical type, but the weight per foot will be about twice that of the present fuel element. The removal of the inlet and outlet piping connections which would be required in the overboring process will permit the replacement of the existing fittings with ones of improved design. Furthermore, new orifices and venturis which are compatible with the hydraulic characteristics of the overbore tube and fuel geometry and the pumping system will be installed. No basic changes are proposed in the pumping system though the reactor flaw rate may be increased 5--10 percent by changes in hydraulic characteristics depending on the water plant flow capacity.

  2. New reactor type proposed

    CERN Multimedia

    2003-01-01

    "Russian scientists at the Research Institute of Nuclear Power Engineering in Moscow are hoping to develop a new reactor that will use lead and bismuth as fuel instead of uranium and plutonium" (1/2 page).

  3. Reactor Neutrino Spectra

    CERN Document Server

    Hayes, A C

    2016-01-01

    We present a review of the antineutrino spectra emitted from reactors. Knowledge of these and their associated uncertainties are crucial for neutrino oscillation studies. The spectra used to-date have been determined by either conversion of measured electron spectra to antineutrino spectra or by summing over all of the thousands of transitions that makeup the spectra using modern databases as input. The uncertainties in the subdominant corrections to beta-decay plague both methods, and we provide estimates of these uncertainties. Improving on current knowledge of the antineutrino spectra from reactors will require new experiments. Such experiments would also address the so-called reactor neutrino anomaly and the possible origin of the shoulder observed in the antineutrino spectra measured in recent high-statistics reactor neutrino experiments.

  4. International kernekraftstatus 1997

    DEFF Research Database (Denmark)

    Højerup, C.F.; Majborn, Benny; Ølgaard, Povl Lebeck

    1998-01-01

    This report is the fourth in a series of annual reports on the international development of nuclear power with special emphasis on reactor safety. For 1997, the report contains: -General trends in the development of nuclear power -A review of what can bedone with the plutonium stocks of the world...

  5. High-intensity power-resolved radiation imaging of an operational nuclear reactor

    Science.gov (United States)

    Beaumont, Jonathan S.; Mellor, Matthew P.; Villa, Mario; Joyce, Malcolm J.

    2015-01-01

    Knowledge of the neutron distribution in a nuclear reactor is necessary to ensure the safe and efficient burnup of reactor fuel. Currently these measurements are performed by in-core systems in what are extremely hostile environments and in most reactor accident scenarios it is likely that these systems would be damaged. Here we present a compact and portable radiation imaging system with the ability to image high-intensity fast-neutron and gamma-ray fields simultaneously. This system has been deployed to image radiation fields emitted during the operation of a TRIGA test reactor allowing a spatial visualization of the internal reactor conditions to be obtained. The imaged flux in each case is found to scale linearly with reactor power indicating that this method may be used for power-resolved reactor monitoring and for the assay of ongoing nuclear criticalities in damaged nuclear reactors. PMID:26450669

  6. High-intensity power-resolved radiation imaging of an operational nuclear reactor

    Science.gov (United States)

    Beaumont, Jonathan S.; Mellor, Matthew P.; Villa, Mario; Joyce, Malcolm J.

    2015-10-01

    Knowledge of the neutron distribution in a nuclear reactor is necessary to ensure the safe and efficient burnup of reactor fuel. Currently these measurements are performed by in-core systems in what are extremely hostile environments and in most reactor accident scenarios it is likely that these systems would be damaged. Here we present a compact and portable radiation imaging system with the ability to image high-intensity fast-neutron and gamma-ray fields simultaneously. This system has been deployed to image radiation fields emitted during the operation of a TRIGA test reactor allowing a spatial visualization of the internal reactor conditions to be obtained. The imaged flux in each case is found to scale linearly with reactor power indicating that this method may be used for power-resolved reactor monitoring and for the assay of ongoing nuclear criticalities in damaged nuclear reactors.

  7. Status and problems of fusion reactor development.

    Science.gov (United States)

    Schumacher, U

    2001-03-01

    Thermonuclear fusion of deuterium and tritium constitutes an enormous potential for a safe, environmentally compatible and sustainable energy supply. The fuel source is practically inexhaustible. Further, the safety prospects of a fusion reactor are quite favourable due to the inherently self-limiting fusion process, the limited radiologic toxicity and the passive cooling property. Among a small number of approaches, the concept of toroidal magnetic confinement of fusion plasmas has achieved most impressive scientific and technical progress towards energy release by thermonuclear burn of deuterium-tritium fuels. The status of thermonuclear fusion research activity world-wide is reviewed and present solutions to the complicated physical and technological problems are presented. These problems comprise plasma heating, confinement and exhaust of energy and particles, plasma stability, alpha particle heating, fusion reactor materials, reactor safety and environmental compatibility. The results and the high scientific level of this international research activity provide a sound basis for the realisation of the International Thermonuclear Experimental Reactor (ITER), whose goal is to demonstrate the scientific and technological feasibility of a fusion energy source for peaceful purposes.

  8. Future Reactor Experiments

    OpenAIRE

    He, Miao

    2013-01-01

    The measurement of the neutrino mixing angle $\\theta_{13}$ opens a gateway for the next generation experiments to measure the neutrino mass hierarchy and the leptonic CP-violating phase. Future reactor experiments will focus on mass hierarchy determination and the precision measurement of mixing parameters. Mass hierarchy can be determined from the disappearance of reactor electron antineutrinos based on the interference effect of two separated oscillation modes. Relative and absolute measure...

  9. Reactor Neutrino Experiments

    OpenAIRE

    Cao, Jun

    2007-01-01

    Precisely measuring $\\theta_{13}$ is one of the highest priority in neutrino oscillation study. Reactor experiments can cleanly determine $\\theta_{13}$. Past reactor neutrino experiments are reviewed and status of next precision $\\theta_{13}$ experiments are presented. Daya Bay is designed to measure $\\sin^22\\theta_{13}$ to better than 0.01 and Double Chooz and RENO are designed to measure it to 0.02-0.03. All are heading to full operation in 2010. Recent improvements in neutrino moment measu...

  10. Department of Reactor Technology

    DEFF Research Database (Denmark)

    Risø National Laboratory, Roskilde

    The general development of the Department of Reactor Technology at Risø during 1981 is presented, and the activities within the major subject fields are described in some detail. Lists of staff, publications, and computer programs are included.......The general development of the Department of Reactor Technology at Risø during 1981 is presented, and the activities within the major subject fields are described in some detail. Lists of staff, publications, and computer programs are included....

  11. Helias reactor studies

    Energy Technology Data Exchange (ETDEWEB)

    Beidler, C.D. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Grieger, G. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Harmeyer, E. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Kisslinger, J. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Karulin, N. [Nuclear Fusion Institute, Moscow (Russian Federation); Maurer, W. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany); Nuehrenberg, J. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Rau, F. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Sapper, J. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Wobig, H. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany)

    1995-10-01

    The present status of Helias reactor studies is characterised by the identification and investigation of specific issues which result from the particular properties of this type of stellarator. On the technical side these are issues related to the coil system, while physics studies have concentrated on confinement, alpha-particle behaviour and ignition conditions. The usual assumptions have been made in those fields which are common to all toroidal fusion reactors: blanket and shield, refuelling and exhaust, safety and economic aspects. For blanket and shield sufficient space has been provided, a detailed concept will be developed in future. To date more emphasis has been placed on scoping and parameter studies as opposed to fixing a specific set of parameters and providing a detailed point study. One result of the Helias reactor studies is that physical dimensions are on the same order as those of tokamak reactors. However, it should be noticed that this comparison is difficult in view of the large spectrum of tokamak reactors ranging from a small reactor like Aries, to a large device such as SEAFP. The notion that the large aspect ratio of 10 or more in Helias configurations also leads to large reactors is misleading, since the large major radius of 22 m is compensated by the average plasma radius of 1.8 m and the average coil radius of 5 m. The plasma volume of 1400 m{sup 3} is about the same as the ITER reactor and the magnetic energy of the coil system is about the same or even slightly smaller than envisaged in ITER. (orig.)

  12. INVAP's Research Reactor Designs

    OpenAIRE

    Eduardo Villarino; Alicia Doval

    2011-01-01

    INVAP, an Argentine company founded more than three decades ago, is today recognized as one of the leaders within the research reactor industry. INVAP has participated in several projects covering a wide range of facilities, designed in accordance with the requirements of our different clients. For complying with these requirements, INVAP developed special skills and capabilities to deal with different fuel assemblies, different core cooling systems, and different reactor layouts. This paper ...

  13. High-temperature reactor developments in the Netherlands

    Energy Technology Data Exchange (ETDEWEB)

    Schram, R.P.C.; Cordfunke, E.H.P.; Heek, A.I. van

    1996-01-01

    The high-temperature reactor development in the Netherland is embedded in the WHITE reactor program, in which several Dutch research institutes and engineering companies participate. The activities within the WHITE program are focused on the development of a small scale HTS for combined heat and power generation. In 1995, design choices for a pebble bed reactor were made at ECN. The first concept HTR will gave a closed cycle helium turbine and a power level of 40 MWth. It is intended to make the market introduction of a commercially competitive HTR feasible. The design will be an optimization of the Peu-a-Peu (PAP) concept of KFA Juelich. Computer codes necessary for the evaluation of reactor physics aspects of this reactor are developed in cooperation with international partners. An evaluation of a 20 MWth PAP concept showed that the maximum fuel termmperature after depressurization does not exceed 1300 C. (orig.).

  14. The reactor antineutrino anomalies

    Energy Technology Data Exchange (ETDEWEB)

    Haser, Julia; Buck, Christian; Lindner, Manfred [Max-Planck-Institut fuer Kernphysik, Heidelberg (Germany)

    2016-07-01

    Major discoveries were made in the past few years in the field of neutrino flavour oscillation. Nuclear reactors produce a clean and intense flux of electron antineutrinos and are thus an essential neutrino source for the determination of oscillation parameters. Most currently the reactor antineutrino experiments Double Chooz, Daya Bay and RENO have accomplished to measure θ{sub 13}, the smallest of the three-flavour mixing angles. In the course of these experiments two anomalies emerged: (1) the reanalysis of the reactor predictions revealed a deficit in experimentally observed antineutrino flux, known as the ''reactor antineutrino anomaly''. (2) The high precision of the latest generation of neutrino experiments resolved a spectral shape distortion relative to the expected energy spectra. Both puzzles are yet to be solved and triggered new experimental as well as theoretical studies, with the search for light sterile neutrinos as most popular explanation for the flux anomaly. This talk outlines the two reactor antineutrino anomalies. Discussing possible explanations for their occurrence, recent and upcoming efforts to solve the reactor puzzles are highlighted.

  15. Moon base reactor system

    Science.gov (United States)

    Chavez, H.; Flores, J.; Nguyen, M.; Carsen, K.

    1989-01-01

    The objective of our reactor design is to supply a lunar-based research facility with 20 MW(e). The fundamental layout of this lunar-based system includes the reactor, power conversion devices, and a radiator. The additional aim of this reactor is a longevity of 12 to 15 years. The reactor is a liquid metal fast breeder that has a breeding ratio very close to 1.0. The geometry of the core is cylindrical. The metallic fuel rods are of beryllium oxide enriched with varying degrees of uranium, with a beryllium core reflector. The liquid metal coolant chosen was natural lithium. After the liquid metal coolant leaves the reactor, it goes directly into the power conversion devices. The power conversion devices are Stirling engines. The heated coolant acts as a hot reservoir to the device. It then enters the radiator to be cooled and reenters the Stirling engine acting as a cold reservoir. The engines' operating fluid is helium, a highly conductive gas. These Stirling engines are hermetically sealed. Although natural lithium produces a lower breeding ratio, it does have a larger temperature range than sodium. It is also corrosive to steel. This is why the container material must be carefully chosen. One option is to use an expensive alloy of cerbium and zirconium. The radiator must be made of a highly conductive material whose melting point temperature is not exceeded in the reactor and whose structural strength can withstand meteor showers.

  16. ANAEROBIC DIGESTION AND THE DENITRIFICATION IN UASB REACTOR

    Directory of Open Access Journals (Sweden)

    José Tavares de Sousa

    2008-01-01

    Full Text Available The environmental conditions in Brazil have been contributing to the development of anaerobic systems in the treatment of wastewaters, especially UASB - Upflow Anaerobic Sludge Blanket reactors. The classic biological process for removal of nutrients uses three reactors - Bardenpho System, therefore, this work intends an alternative system, where the anaerobic digestion and the denitrification happen in the same reactor reducing the number of reactors for two. The experimental system was constituted by two units: first one was a nitrification reactor with 35 L volume and 15 d of sludge age. This system was fed with raw sanitary waste. Second unit was an UASB, with 7.8 L and 6 h of hydraulic detention time, fed with ¾ of effluent nitrification reactor and ¼ of raw sanitary waste. This work had as objective to evaluate the performance of the UASB reactor. In terms of removal efficiency, of bath COD and nitrogen, it was verified that the anaerobic digestion process was not affected. The removal efficiency of organic material expressed in COD was 71%, performance already expected for a reactor of this type. It was also observed that the denitrification process happened; the removal nitrate efficiency was 90%. Therefore, the denitrification process in reactor UASB is viable.

  17. Reduced enrichment for research and test reactors: Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    1993-07-01

    The 15th annual Reduced Enrichment for Research and Test Reactors (RERTR) international meeting was organized by Ris{o} National Laboratory in cooperation with the International Atomic Energy Agency and Argonne National Laboratory. The topics of the meeting were the following: National Programs, Fuel Fabrication, Licensing Aspects, States of Conversion, Fuel Testing, and Fuel Cycle. Individual papers have been cataloged separately.

  18. Assessing the Validity of Self-Rated Health with the Short Physical Performance Battery: A Cross-Sectional Analysis of the International Mobility in Aging Study.

    Directory of Open Access Journals (Sweden)

    Mario U Pérez-Zepeda

    Full Text Available The aim of this study was to explore the validity of self-rated health across different populations of older adults, when compared to the Short Physical Performance Battery.Cross-sectional analysis of the International Mobility in Aging Study.Five locations: Saint-Hyacinthe and Kingston (Canada, Tirana (Albania, Manizales (Colombia, and Natal (Brazil.Older adults between 65 and 74 years old (n = 1,995.The Short Physical Performance Battery (SPPB was used to measure physical performance. Self-rated health was assessed with one single five-point question. Linear trends between SPPB scores and self-rated health were tested separately for men and women at each of the five international study sites. Poor physical performance (independent variable (SPPB less than 8 was used in logistic regression models of self-rated health (dependent variable, adjusting for potential covariates. All analyses were stratified by gender and site of origin.A significant linear association was found between the mean scores of the Short Physical Performance Battery and ordinal categories of self-rated health across research sites and gender groups. After extensive control for objective physical and mental health indicators and socio-demographic variables, these graded associations became non-significant in some research sites.These findings further confirm the validity of SRH as a measure of overall health status in older adults.

  19. Happy Aging

    Institute of Scientific and Technical Information of China (English)

    梁秉中

    2009-01-01

    Aging is a normal physiological process in human life.The decline in the ability to repair and regenerate predisposes the aging person to develop disabling problems in the cardiovascular and skeletal systems.Full awareness of aging problems and advocations on the means to prevent their occurrence are mounting.European and US groups rely on scientific,target-oriented means to treat aging manifestations. Oriental medicine aims at prevention,using nutrition and exercise to maintain internal harmony.

  20. Proliferation resistance assessment of high temperature gas reactors

    Energy Technology Data Exchange (ETDEWEB)

    Chikamatsu N, M. A. [Instituto Tecnologico y de Estudios Superiores de Monterrey, Campus Santa Fe, Av. Carlos Lazo No. 100, Santa Fe, 01389 Mexico D. F. (Mexico); Puente E, F., E-mail: midori.chika@gmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2014-10-15

    The Generation IV International Forum has established different objectives for the new generation of reactors to accomplish. These objectives are focused on sustain ability, safety, economics and proliferation resistance. This paper is focused on how the proliferation resistance of the High Temperature Gas Reactors (HTGR) is assessed and the advantages that these reactors present currently. In this paper, the focus will be on explaining why such reactors, HTGR, can achieve the goals established by the GIF and can present a viable option in terms of proliferation resistance, which is an issue of great importance in the field of nuclear energy generation. The reason why the HTGR are being targeted in this writing is that these reactors are versatile, and present different options from modular reactors to reactors with the same size as the ones that are being operated today. Besides their versatility, the HTGR has designed features that might improve on the overall sustain ability of the nuclear reactors. This is because the type of safety features and materials that are used open up options for industrial processes to be carried out; cogeneration for instance. There is a small section that mentions how HTGR s are being developed in the international sector in order to present the current world view in this type of technology and the further developments that are being sought. For the proliferation resistance section, the focus is on both the intrinsic and the extrinsic features of the nuclear systems. The paper presents a comparison between the features of Light Water Reactors (LWR) and the HTGR in order to be able to properly compare the most used technology today and one that is gaining international interest. (Author)

  1. Irradiation rigs in material testing reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rozenblum, F.; Gonnier, C.; Bignan, G. [CEA, Research Centers of Saclay and Cadarache (France)

    2011-07-01

    Osiris is a research reactor with a thermal power of 70 MW. It is a light-water reactor, open-core pool type, the principal aim of which is to carry out tests and irradiate structural materials and fuel elements of nuclear power plants under a high flux of neutrons, and to produce radioisotopes. Osiris operates around 200 days a year, in cycles of varying lengths from 3 to 4 weeks. A shutdown of about 10 days between two cycles allows reloading the core with fuel. Mainly 2 types of irradiation device are present: capsules for materials irradiation (CHOUCA and IRMA devices) and fuels irradiation loops (GRIFFONOS and ISABELLE). Although Osiris is still providing experiments of very good quality, it is facing obsolescence due to its ageing. Osiris is planned to be shut down during next decade. Consequently, it has been decided to launch the construction of the Jules Horowitz Reactor (JHR) in Cadarache. JHR is a water cooled reactor which provides the necessary flexibility and accessibility to manage several highly instrumented experiments, reproducing different reactor environments (water, gas or liquid metal loops), generating transient regimes (key for safety). The JHR facility includes the reactor building, including core, cooling system and the experimental bunkers connected to the core through pool wall penetrations and the auxiliary building, including pools and hot cells necessary for the experimental irradiation process. JHR core is optimised to produce high fast neutron flux to study structural material ageing and high thermal neutrons flux for fuel experiments. The conception of this first fleet of devices integrates the operational experience accumulated by the existing MTR and specifically the Osiris one

  2. Reactor Safety Planning for Prometheus Project, for Naval Reactors Information

    Energy Technology Data Exchange (ETDEWEB)

    P. Delmolino

    2005-05-06

    The purpose of this letter is to submit to Naval Reactors the initial plan for the Prometheus project Reactor Safety work. The Prometheus project is currently developing plans for cold physics experiments and reactor prototype tests. These tests and facilities may require safety analysis and siting support. In addition to the ground facilities, the flight reactor units will require unique analyses to evaluate the risk to the public from normal operations and credible accident conditions. This letter outlines major safety documents that will be submitted with estimated deliverable dates. Included in this planning is the reactor servicing documentation and shipping analysis that will be submitted to Naval Reactors.

  3. [Our viewpoints on Deqi in the later ages after birth of classical works "The Yellow Emperor's Internal Classic" and "Canon of Difficult Medical Problems"].

    Science.gov (United States)

    Hao, Jie; Zhu, Jiang; Zhang, Peng; Xin, Si-Yuan; Qi, Dan-Dan; Hu, Ni-Juan; Lin, Chi; Wang, Pei; Zhao, Min-Yi; Hu, Shang-Qing; Wu, Gui-Wen

    2015-04-01

    In our previous paper, we analyzed "Deqi" in book Huangdi Neijing (The Yellow Emperor's Internal Classic) and Nanjing (Canon of Difficult Medical Problems) from "Zhishen"(Treating mentality) and Tiaoqi (Regulating qi). In the present paper, the authors discuss the connotations of "Deqi" and related events in the later ages of the abovementioned two classic books to the later stage of the Qing Dynasty when involves about 20 classical works as Zhenjiu Dacheng ( The Great Compendium of Acupuncture and Moxibustion), Zhenjing Zhinan (Guide to the classics of Acupuncture), Zhenjiu Daquan (A Complete Works of Acupuncture and Moxibustion) etc. from 1) close association between "Deqi" and patients' mental activity; 2) how to wait for arrival of qi if the needling does not induce "Deqi" for the time being; 3) how to identify "qi-arrival" and then, performing suitable manipulations; 4) Deqi and shallow- or deep-needling; 5) putting more emphasis on patients' feeling and reactions, rather than the practitioners perception beneath the needle which is described in book Huangdi Neijing; and 6) not withdrawing the acupuncture needles if qi does not arrive. Generally, in the later ages, the connotations of Deqi are enriched greatly.

  4. Fuel development for gas-cooled fast reactors

    Science.gov (United States)

    Meyer, M. K.; Fielding, R.; Gan, J.

    2007-09-01

    The Generation IV Gas-cooled Fast Reactor (GFR) concept is proposed to combine the advantages of high-temperature gas-cooled reactors (such as efficient direct conversion with a gas turbine and the potential for application of high-temperature process heat), with the sustainability advantages that are possible with a fast-spectrum reactor. The latter include the ability to fission all transuranics and the potential for breeding. The GFR is part of a consistent set of gas-cooled reactors that includes a medium-term Pebble Bed Modular Reactor (PBMR)-like concept, or concepts based on the Gas Turbine Modular Helium Reactor (GT-MHR), and specialized concepts such as the Very High-Temperature Reactor (VHTR), as well as actinide burning concepts [A Technology Roadmap for Generation IV Nuclear Energy Systems, US DOE Nuclear Energy Research Advisory Committee and the Generation IV International Forum, December 2002]. To achieve the necessary high power density and the ability to retain fission gas at high temperature, the primary fuel concept proposed for testing in the United States is dispersion coated fuel particles in a ceramic matrix. Alternative fuel concepts considered in the US and internationally include coated particle beds, ceramic clad fuel pins, and novel ceramic 'honeycomb' structures. Both mixed carbide and mixed nitride-based solid solutions are considered as fuel phases.

  5. REACTOR GROUT THERMAL PROPERTIES

    Energy Technology Data Exchange (ETDEWEB)

    Steimke, J.; Qureshi, Z.; Restivo, M.; Guerrero, H.

    2011-01-28

    Savannah River Site has five dormant nuclear production reactors. Long term disposition will require filling some reactor buildings with grout up to ground level. Portland cement based grout will be used to fill the buildings with the exception of some reactor tanks. Some reactor tanks contain significant quantities of aluminum which could react with Portland cement based grout to form hydrogen. Hydrogen production is a safety concern and gas generation could also compromise the structural integrity of the grout pour. Therefore, it was necessary to develop a non-Portland cement grout to fill reactors that contain significant quantities of aluminum. Grouts generate heat when they set, so the potential exists for large temperature increases in a large pour, which could compromise the integrity of the pour. The primary purpose of the testing reported here was to measure heat of hydration, specific heat, thermal conductivity and density of various reactor grouts under consideration so that these properties could be used to model transient heat transfer for different pouring strategies. A secondary purpose was to make qualitative judgments of grout pourability and hardened strength. Some reactor grout formulations were unacceptable because they generated too much heat, or started setting too fast, or required too long to harden or were too weak. The formulation called 102H had the best combination of characteristics. It is a Calcium Alumino-Sulfate grout that contains Ciment Fondu (calcium aluminate cement), Plaster of Paris (calcium sulfate hemihydrate), sand, Class F fly ash, boric acid and small quantities of additives. This composition afforded about ten hours of working time. Heat release began at 12 hours and was complete by 24 hours. The adiabatic temperature rise was 54 C which was within specification. The final product was hard and displayed no visible segregation. The density and maximum particle size were within specification.

  6. Scaleable, High Efficiency Microchannel Sabatier Reactor Project

    Data.gov (United States)

    National Aeronautics and Space Administration — A Microchannel Sabatier Reactor System (MSRS) consisting of cross connected arrays of isothermal or graded temperature reactors is proposed. The reactor array...

  7. Use of national and international growth charts for studying height in European children: development of up-to-date European height-for-age charts.

    Directory of Open Access Journals (Sweden)

    Marjolein Bonthuis

    Full Text Available BACKGROUND: Growth charts based on data collected in different populations and time periods are key tools to assess children's linear growth. We analyzed the impact of geographic factors and the secular trend on height-for-age charts currently used in European populations, developed up-to-date European growth charts, and studied the effect of using different charts in a sample of growth retarded children. METHODS AND FINDINGS: In an international survey we obtained 18 unique national height-for-age charts from 28 European countries and compared them with charts from the World Health Organization (WHO, Euro-Growth reference, and Centers of Disease Control and Prevention (CDC. As an example, we obtained height data from 3,534 children with end-stage renal disease (ESRD from 13 countries via the ESPN/ERA-EDTA registry, a patient group generally suffering from growth retardation. National growth charts showed a clear secular trend in height (mean height increased on average 0.6 cm/decade and a North-South height gradient in Europe. For countries without a recent (>1990 national growth chart novel European growth charts were constructed from Northern and Southern European reference populations, reflecting geographic height differences in mean final height of 3.9 cm in boys and 3.8 cm in girls. Mean height SDS of 2- to 17-year-old ESRD patients calculated from recent national or derived European growth charts (-1.91, 95% CI: -1.97 to -1.85 was significantly lower than when using CDC or WHO growth charts (-1.55, 95% CI: -1.61 to -1.49 (P<0.0001. CONCLUSION: Differences between height-for-age charts may reflect true population differences, but are also strongly affected by the secular trend in height. The choice of reference charts substantially affects the clinical decision whether a child is considered short-for-age. Therefore, we advocate using recent national or European height-for-age charts derived from recent national data when monitoring growth

  8. Health hazards and medical treatment of volunteers aged 18-30 years working in international social projects of non-governmental organizations (NGO).

    Science.gov (United States)

    Küpper, T; Rieke, B; Neppach, K; Morrison, A; Martin, J

    2014-01-01

    The specific health risk profile and diversity of treatments sought by young volunteers participating in international social projects should differ from those of their older colleagues. In the absence of any data to identify whether this was correct, a retrospective analysis was performed using a standardized questionnaire. Questions included what diseases occurred, and details of the frequency and types of treatment sought during their stay - (e.g. self-treatment, medical/dental intervention, or local healer). The 153 participants were aged 18-30 years and worked in a non-governmental organization for >6 months. The participants were: 53% female, mean age 20 years, and mean duration of stay was 11.2 months. Their NGO placement abroad was in Latin America 65.4%, 14.4% in Africa, and 9.8% in Asia. 83% of the young volunteers had received some advice regarding travel medicine before their departure. However, they suffered from more injuries compared to private travellers, and febrile infections were more common when compared to older studies. 21.2% suffered from dental problems and 50% of them sought medical treatment. This study highlights a previously unreported higher risk profile of specific health problems occurring in young NGO volunteers, including some potentially life-threatening diagnoses that differed from their older colleagues and normal travellers. It is recommended that young volunteers should receive age specific, comprehensive pre-departure training in health and safety, first aid, and management of common health problems. A medical check-up upon returning home should be mandatory. The provision of a basic first aid kit to each volunteer before departure is also recommended.

  9. Sodium fast reactor safety and licensing research plan. Volume II.

    Energy Technology Data Exchange (ETDEWEB)

    Ludewig, H. (Brokhaven National Laboratory, Upton, NY); Powers, D. A.; Hewson, John C.; LaChance, Jeffrey L.; Wright, A. (Argonne National Laboratory, Argonne, IL); Phillips, J.; Zeyen, R. (Institute for Energy Petten, Saint-Paul-lez-Durance, France); Clement, B. (IRSN/DPAM.SEMIC Bt 702, Saint-Paul-lez-Durance, France); Garner, Frank (Radiation Effects Consulting, Richland, WA); Walters, Leon (Advanced Reactor Concepts, Los Alamos, NM); Wright, Steve; Ott, Larry J. (Oak Ridge National Laboratory, Oak Ridge, TN); Suo-Anttila, Ahti Jorma; Denning, Richard (Ohio State University, Columbus, OH); Ohshima, Hiroyuki (Japan Atomic Energy Agency, Ibaraki, Japan); Ohno, S. (Japan Atomic Energy Agency, Ibaraki, Japan); Miyhara, S. (Japan Atomic Energy Agency, Ibaraki, Japan); Yacout, Abdellatif (Argonne National Laboratory, Argonne, IL); Farmer, M. (Argonne National Laboratory, Argonne, IL); Wade, D. (Argonne National Laboratory, Argonne, IL); Grandy, C. (Argonne National Laboratory, Argonne, IL); Schmidt, R.; Cahalen, J. (Argonne National Laboratory, Argonne, IL); Olivier, Tara Jean; Budnitz, R. (Lawrence Berkeley National Laboratory, Berkeley, CA); Tobita, Yoshiharu (Japan Atomic Energy Agency, Ibaraki, Japan); Serre, Frederic (Centre d' %C3%94etudes nucl%C3%94eaires de Cadarache, Cea, France); Natesan, Ken (Argonne National Laboratory, Argonne, IL); Carbajo, Juan J. (Oak Ridge National Laboratory, Oak Ridge, TN); Jeong, Hae-Yong (Korea Atomic Energy Research Institute, Daejeon, Korea); Wigeland, Roald (Idaho National Laboratory, Idaho Falls, ID); Corradini, Michael (University of Wisconsin-Madison, Madison, WI); Thomas, Justin (Argonne National Laboratory, Argonne, IL); Wei, Tom (Argonne National Laboratory, Argonne, IL); Sofu, Tanju (Argonne National Laboratory, Argonne, IL); Flanagan, George F. (Oak Ridge National Laboratory, Oak Ridge, TN); Bari, R. (Brokhaven National Laboratory, Upton, NY); Porter D. (Idaho National Laboratory, Idaho Falls, ID); Lambert, J. (Argonne National Laboratory, Argonne, IL); Hayes, S. (Idaho National Laboratory, Idaho Falls, ID); Sackett, J. (Idaho National Laboratory, Idaho Falls, ID); Denman, Matthew R.

    2012-05-01

    Expert panels comprised of subject matter experts identified at the U.S. National Laboratories (SNL, ANL, INL, ORNL, LBL, and BNL), universities (University of Wisconsin and Ohio State University), international agencies (IRSN, CEA, JAEA, KAERI, and JRC-IE) and private consultation companies (Radiation Effects Consulting) were assembled to perform a gap analysis for sodium fast reactor licensing. Expert-opinion elicitation was performed to qualitatively assess the current state of sodium fast reactor technologies. Five independent gap analyses were performed resulting in the following topical reports: (1) Accident Initiators and Sequences (i.e., Initiators/Sequences Technology Gap Analysis), (2) Sodium Technology Phenomena (i.e., Advanced Burner Reactor Sodium Technology Gap Analysis), (3) Fuels and Materials (i.e., Sodium Fast Reactor Fuels and Materials: Research Needs), (4) Source Term Characterization (i.e., Advanced Sodium Fast Reactor Accident Source Terms: Research Needs), and (5) Computer Codes and Models (i.e., Sodium Fast Reactor Gaps Analysis of Computer Codes and Models for Accident Analysis and Reactor Safety). Volume II of the Sodium Research Plan consolidates the five gap analysis reports produced by each expert panel, wherein the importance of the identified phenomena and necessities of further experimental research and code development were addressed. The findings from these five reports comprised the basis for the analysis in Sodium Fast Reactor Research Plan Volume I.

  10. Dismantling design for the loop rooms on the MR reactor

    Energy Technology Data Exchange (ETDEWEB)

    Craig, D.; Fecitt, L. [NUKEM Limited, Dounreay (United Kingdom); Gorlinsky, Yu.E. [RRC Kurchatov Institute, Moscow (Russian Federation); Harman, N.F.; Jackson, R. [Serco Technical and Assurance Services, Warrington (United Kingdom); Kolyadin, V.I. [RRC Kurchatov Institute, Moscow (Russian Federation); Lobach, Yu.N., E-mail: lobach@kinr.kiev.u [Institute for Nuclear Research of NASU, pr.Nauki, 47, 03680 Kiev (Ukraine); Pavlenko, V.I. [RRC Kurchatov Institute, Moscow (Russian Federation)

    2009-12-15

    The recently completed international co-operation project was aimed at planning for decommissioning the MR reactor identified as a pilot plant for the decommissioning of the other shutdown reactors on the site. The MR reactor was a pool-type, materials testing reactor with the total thermal power of 50 MW which incorporated pressure tubes containing fuel under test. The MR facility includes the reactor with its nine loop rig rooms containing pumps, heat exchangers and experimental equipment as well as systems and equipment located in other buildings in the complex. The objective of the MR reactor decommissioning project was to identify dismantling equipment and the decommissioning methodology for the reactor, loop rooms and redundant services to permit the refit and re-use of the building for a different nuclear related purpose. The dismantling design comprises two separate, but combined, tasks, namely, the dismantling of reactor installation itself and dismantling of experimental loops. The techniques proposed to undertake the dismantling operations within the loop rooms are described. Two options have been developed for removing contaminated equipment from the high radiation field loop rooms and packaging the waste into approved waste containers. The benefits and detriments of both methods have been identified, which allows implementing the safe, timely and cost-effective decommissioning.

  11. Present status and future perspectives of research and test reactor in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Kaneko, Yoshihiko [Atomic Energy Research Laboratory, Musashi Institute of Technology, Kawasaki, Kanagawa (Japan); Kaieda, Keisuke [Department of Research Reactor, Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan)

    2000-10-01

    Since 1957, Japan Atomic Energy Research Institute (JAERI) has constructed several research and test reactors to fulfill a major role in the study of nuclear energy and fundamental research. At present four reactors, the Japan Research Reactor No. 3 and No. 4 (JRR-3M and JRR-4 respectively), the Japan Materials Testing Reactor (JMTR) and the Nuclear Safety Research Reactor (NSRR) are in operation, and a new High Temperature Engineering Test Reactor (HTTR) has recently reached first criticality and now in the power up test. In 1966, the Kyoto University built the Kyoto University Reactor (KUR) and started its operation for joint use program of the Japanese universities. This paper introduces these reactors and describes their present operational status and also efforts for aging management. The recent tendency of utilization and future perspectives is also reported. (author)

  12. LMFBR type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shimizu, Takeshi; Iida, Masaaki; Moriki, Yasuyuki

    1994-10-18

    A reactor core is divided into a plurality of coolants flowrate regions, and electromagnetic pumps exclusively used for each of the flowrate regions are disposed to distribute coolants flowrates in the reactor core. Further, the flowrate of each of the electromagnetic pumps is automatically controlled depending on signals from a temperature detector disposed at the exit of the reactor core, so that the flowrate of the region can be controlled optimally depending on the burning of reactor core fuels. Then, the electromagnetic pumps disposed for every divided region are controlled respectively, so that the coolants flowrate distribution suitable to each of the regions can be attained. Margin for fuel design is decreased, fuels are used effectively, as well as an operation efficiency can be improved. Moreover, since the electromagnetic pump has less flow resistance compared with a mechanical type pump, and flow resistance of the reactor core flowrate control mechanism is eliminated, greater circulating flowrate can be ensured after occurrence of accident in a natural convection using a buoyancy of coolants utilizable for after-heat removal as a driving force. (N.H.).

  13. Worldwide advanced nuclear power reactors with passive and inherent safety: What, why, how, and who

    Energy Technology Data Exchange (ETDEWEB)

    Forsberg, C.W.; Reich, W.J.

    1991-09-01

    The political controversy over nuclear power, the accidents at Three Mile Island (TMI) and Chernobyl, international competition, concerns about the carbon dioxide greenhouse effect and technical breakthroughs have resulted in a segment of the nuclear industry examining power reactor concepts with PRIME safety characteristics. PRIME is an acronym for Passive safety, Resilience, Inherent safety, Malevolence resistance, and Extended time after initiation of an accident for external help. The basic ideal of PRIME is to develop power reactors in which operator error, internal sabotage, or external assault do not cause a significant release of radioactivity to the environment. Several PRIME reactor concepts are being considered. In each case, an existing, proven power reactor technology is combined with radical innovations in selected plant components and in the safety philosophy. The Process Inherent Ultimate Safety (PIUS) reactor is a modified pressurized-water reactor, the Modular High Temperature Gas-Cooled Reactor (MHTGR) is a modified gas-cooled reactor, and the Advanced CANDU Project is a modified heavy-water reactor. In addition to the reactor concepts, there is parallel work on super containments. The objective is the development of a passive box'' that can contain radioactivity in the event of any type of accident. This report briefly examines: why a segment of the nuclear power community is taking this new direction, how it differs from earlier directions, and what technical options are being considered. A more detailed description of which countries and reactor vendors have undertaken activities follows. 41 refs.

  14. Reactor Structural Materials: Reactor Pressure Vessel Steels

    Energy Technology Data Exchange (ETDEWEB)

    Chaouadi, R

    2000-07-01

    The objectives of SCK-CEN's R and D programme on Rector Pressure Vessel (RPV) Steels are:(1) to complete the fracture toughness data bank of various reactor pressure vessel steels by using precracked Charpy specimens that were tested statically as well as dynamically; (2) to implement the enhanced surveillance approach in a user-friendly software; (3) to improve the existing reconstitution technology by reducing the input energy (short cycle welding) and modifying the stud geometry. Progress and achievements in 1999 are reported.

  15. Improvement of system code importing evaluation of Life Cycle Analysis of tokamak fusion power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kobori, Hikaru [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Kasada, Ryuta, E-mail: r-kasada@iae.kyoto-u.ac.jp [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Hiwatari, Ryoji [Central Research Institute of Electric Power Industry, Tokyo (Japan); Konishi, Satoshi [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan)

    2016-11-01

    Highlights: • We incorporated the Life Cycle Analysis (LCA) of tokamak type DEMO reactor and following commercial reactors as an extension of a system code. • We calculated CO{sub 2} emissions from reactor construction, operation and decommissioning that is considered as a major environmental cost. • We found that the objective of conceptual design of the tokamak fusion power reactor is moved by changing evaluation index. • The tokamak fusion reactor can reduce CO{sub 2} emissions in the life cycle effectively by reduction of the amount involved in the replacement of internal components. • The tokamak fusion reactor achieves under 0.174$/kWh electricity cost, the tokamak fusion reactor is contestable with 1500 degrees-class LNG-fired combined cycle power plant. - Abstract: This study incorporate the Life Cycle Analysis (LCA) of tokamak type DEMO reactor and following commercial reactors as an extension of a system code to calculate CO{sub 2} emissions from reactor construction, operation and decommissioning that is considered as a major environmental cost. Competitiveness of tokamak fusion power reactors is expected to be evaluated by the cost and environmental impact represented by the CO{sub 2} emissions, compared with present and future power generating systems such as fossil, nuclear and renewables. Result indicated that (1) The objective of conceptual design of the tokamak fusion power reactor is moved by changing evaluation index. (2) The tokamak fusion reactor can reduce CO{sub 2} emissions in the life cycle effectively by reduction of the amount involved in the replacement of internal components. (3) The tokamak fusion reactor achieves under 0.174$/kWh electricity cost, the tokamak fusion reactor is contestable with 1500 degrees-class LNG-fired combined cycle power plant.

  16. Operation of Reactor

    Institute of Scientific and Technical Information of China (English)

    1996-01-01

    3.1 Annual Report of SPR Operation Chu Shaochu Having overseen by National Nuclear Safety Administration and specialists, the reactor restarted up successfully after Safety renovation on April 16, 1996. In August 1996 the normal operation of SPR was approved by the authorities of Naitonal Nuclear Safety Administration. 1 Operation status In 1996, the reactor operated safely for 40 d and the energy released was about 137.3 MW·d. The operation status of SPR is shown in table 1. The reactor started up to higher power (power more than 1 MW) and lower power (for physics experiments) 4 times and 14 times respectively. Measurement of control rod efficiency and other measurement tasks were 2 times and 5 times respectively.

  17. Nuclear Rocket Engine Reactor

    CERN Document Server

    Lanin, Anatoly

    2013-01-01

    The development of a nuclear rocket engine reactor (NRER ) is presented in this book. The working capacity of an active zone NRER under mechanical and thermal load, intensive neutron fluxes, high energy generation (up to 30 MBT/l) in a working medium (hydrogen) at temperatures up to 3100 K is displayed. Design principles and bearing capacity of reactors area discussed on the basis of simulation experiments and test data of a prototype reactor. Property data of dense constructional, porous thermal insulating and fuel materials like carbide and uranium carbide compounds in the temperatures interval 300 - 3000 K are presented. Technological aspects of strength and thermal strength resistance of materials are considered. The design procedure of possible emergency processes in the NRER is developed and risks for their origination are evaluated. Prospects of the NRER development for pilotless space devices and piloted interplanetary ships are viewed.

  18. Thermionic Reactor Design Studies

    Energy Technology Data Exchange (ETDEWEB)

    Schock, Alfred

    1994-08-01

    Paper presented at the 29th IECEC in Monterey, CA in August 1994. The present paper describes some of the author's conceptual designs and their rationale, and the special analytical techniques developed to analyze their (thermionic reactor) performance. The basic designs, first published in 1963, are based on single-cell converters, either double-ended diodes extending over the full height of the reactor core or single-ended diodes extending over half the core height. In that respect they are similar to the thermionic fuel elements employed in the Topaz-2 reactor subsequently developed in the Soviet Union, copies of which were recently imported by the U.S. As in the Topaz-2 case, electrically heated steady-state performance tests of the converters are possible before fueling.

  19. An Overview of Reactor Concepts, a Survey of Reactor Designs.

    Science.gov (United States)

    1985-02-01

    Public Affairs Office and is releasaole to the National Technical Information Services (NTIS). At NTIS, it will be available to the general public...Reactors that use deu- terium (heavy water) as a coolant can use natural uranium as a fuel. The * Canadian reactor, CANDU , utilizes this concept...reactor core at the top and discharged at the Dotton while the reactor is in operation. The discharged fuel can then b inspected to see if it can De used

  20. Oscillatory flow chemical reactors

    Directory of Open Access Journals (Sweden)

    Slavnić Danijela S.

    2014-01-01

    Full Text Available Global market competition, increase in energy and other production costs, demands for high quality products and reduction of waste are forcing pharmaceutical, fine chemicals and biochemical industries, to search for radical solutions. One of the most effective ways to improve the overall production (cost reduction and better control of reactions is a transition from batch to continuous processes. However, the reactions of interests for the mentioned industry sectors are often slow, thus continuous tubular reactors would be impractically long for flow regimes which provide sufficient heat and mass transfer and narrow residence time distribution. The oscillatory flow reactors (OFR are newer type of tube reactors which can offer solution by providing continuous operation with approximately plug flow pattern, low shear stress rates and enhanced mass and heat transfer. These benefits are the result of very good mixing in OFR achieved by vortex generation. OFR consists of cylindrical tube containing equally spaced orifice baffles. Fluid oscillations are superimposed on a net (laminar flow. Eddies are generated when oscillating fluid collides with baffles and passes through orifices. Generation and propagation of vortices create uniform mixing in each reactor cavity (between baffles, providing an overall flow pattern which is close to plug flow. Oscillations can be created by direct action of a piston or a diaphragm on fluid (or alternatively on baffles. This article provides an overview of oscillatory flow reactor technology, its operating principles and basic design and scale - up characteristics. Further, the article reviews the key research findings in heat and mass transfer, shear stress, residence time distribution in OFR, presenting their advantages over the conventional reactors. Finally, relevant process intensification examples from pharmaceutical, polymer and biofuels industries are presented.

  1. Perspectives on reactor safety

    Energy Technology Data Exchange (ETDEWEB)

    Haskin, F.E. [New Mexico Univ., Albuquerque, NM (United States). Dept. of Chemical and Nuclear Engineering; Camp, A.L. [Sandia National Labs., Albuquerque, NM (United States)

    1994-03-01

    The US Nuclear Regulatory Commission (NRC) maintains a technical training center at Chattanooga, Tennessee to provide appropriate training to both new and experienced NRC employees. This document describes a one-week course in reactor, safety concepts. The course consists of five modules: (1) historical perspective; (2) accident sequences; (3) accident progression in the reactor vessel; (4) containment characteristics and design bases; and (5) source terms and offsite consequences. The course text is accompanied by slides and videos during the actual presentation of the course.

  2. Reactor Materials Research

    Energy Technology Data Exchange (ETDEWEB)

    Van Walle, E

    2002-04-01

    The activities of SCK-CEN's Reactor Materials Research Department for 2001 are summarised. The objectives of the department are: (1) to evaluate the integrity and behaviour of structural materials used in nuclear power industry; (2) to conduct research to unravel and understand the parameters that determine the material behaviour under or after irradiation; (3) to contribute to the interpretation, the modelling of the material behaviour and to develop and assess strategies for optimum life management of nuclear power plant components. The programmes within the department are focussed on studies concerning (1) Irradiation Assisted Stress Corrosion Cracking (IASCC); (2) nuclear fuel; and (3) Reactor Pressure Vessel Steel.

  3. Fusion Reactor Materials

    Energy Technology Data Exchange (ETDEWEB)

    Decreton, M

    2000-07-01

    SCK-CEN's research and development programme on fusion reactor materials includes: (1) the study of the mechanical behaviour of structural materials under neutron irradiation (including steels, inconel, molybdenum, chromium); (2) the determination and modelling of the characteristics of irradiated first wall materials such as beryllium; (3) the detection of abrupt electrical degradation of insulating ceramics under high temperature and neutron irradiation; (4) the study of the dismantling and waste disposal strategy for fusion reactors.; (5) a feasibility study for the testing of blanket modules under neutron radiation. Main achievements in these topical areas in the year 1999 are summarised.

  4. Benchmark Evaluation of the NRAD Reactor LEU Core Startup Measurements

    Energy Technology Data Exchange (ETDEWEB)

    J. D. Bess; T. L. Maddock; M. A. Marshall

    2011-09-01

    The Neutron Radiography (NRAD) reactor is a 250-kW TRIGA-(Training, Research, Isotope Production, General Atomics)-conversion-type reactor at the Idaho National Laboratory; it is primarily used for neutron radiography analysis of irradiated and unirradiated fuels and materials. The NRAD reactor was converted from HEU to LEU fuel with 60 fuel elements and brought critical on March 31, 2010. This configuration of the NRAD reactor has been evaluated as an acceptable benchmark experiment and is available in the 2011 editions of the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP Handbook) and the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook). Significant effort went into precisely characterizing all aspects of the reactor core dimensions and material properties; detailed analyses of reactor parameters minimized experimental uncertainties. The largest contributors to the total benchmark uncertainty were the 234U, 236U, Er, and Hf content in the fuel; the manganese content in the stainless steel cladding; and the unknown level of water saturation in the graphite reflector blocks. A simplified benchmark model of the NRAD reactor was prepared with a keff of 1.0012 {+-} 0.0029 (1s). Monte Carlo calculations with MCNP5 and KENO-VI and various neutron cross section libraries were performed and compared with the benchmark eigenvalue for the 60-fuel-element core configuration; all calculated eigenvalues are between 0.3 and 0.8% greater than the benchmark value. Benchmark evaluations of the NRAD reactor are beneficial in understanding biases and uncertainties affecting criticality safety analyses of storage, handling, or transportation applications with LEU-Er-Zr-H fuel.

  5. High Flux Isotope Reactor (HFIR)

    Data.gov (United States)

    Federal Laboratory Consortium — The HFIR at Oak Ridge National Laboratory is a light-water cooled and moderated reactor that is the United States’ highest flux reactor-based neutron source. HFIR...

  6. Reactor operation environmental information document

    Energy Technology Data Exchange (ETDEWEB)

    Haselow, J.S.; Price, V.; Stephenson, D.E.; Bledsoe, H.W.; Looney, B.B.

    1989-12-01

    The Savannah River Site (SRS) produces nuclear materials, primarily plutonium and tritium, to meet the requirements of the Department of Defense. These products have been formed in nuclear reactors that were built during 1950--1955 at the SRS. K, L, and P reactors are three of five reactors that have been used in the past to produce the nuclear materials. All three of these reactors discontinued operation in 1988. Currently, intense efforts are being extended to prepare these three reactors for restart in a manner that protects human health and the environment. To document that restarting the reactors will have minimal impacts to human health and the environment, a three-volume Reactor Operations Environmental Impact Document has been prepared. The document focuses on the impacts of restarting the K, L, and P reactors on both the SRS and surrounding areas. This volume discusses the geology, seismology, and subsurface hydrology. 195 refs., 101 figs., 16 tabs.

  7. Evidence for Archean inheritance in the pre-Panafrican crust of Central Cameroon: Insight from zircon internal structure and LA-MC-ICP-MS Usbnd Pb ages

    Science.gov (United States)

    Ganwa, Alembert Alexandre; Klötzli, Urs Stephan; Hauzenberger, Christoph

    2016-08-01

    The main geological feature of Central Cameroon is the wide spread occurrence of granitoids emplaced in close association with transcurrent regional shear zones. The basement of this vast domain is a Paleoproterozoic ortho-and para-derivative formation, which has been intensely reworked, together with subsequent intrusions and sediments, during the Panafrican orogenesis in the Neoproterozoic. As consequence, the area underwent pervasive metamorphism and intense deformation. This makes it difficult to distinguish between Panafrican metasediments or syntectonic plutonites and their respective basement. Our study presents zircon features (CL-BSE-SE) and in-situ U-Th-Pb LA-MC-ICP-MS geochronology of a meta-sedimentary pyroxene-amphibole-bearing gneiss of the Méiganga area in Central Cameroon. Based on the Internal structures of the zircon four characteristic zonation patterns can be deciphered: 1) cores with magmatic oscillatory zonation 2) zircons with oscillatory or sector zonation, 3) zircons with sector zoning or blurred zoning, and 4) narrow bright un-zoned rims. These groups suggest that the rock experienced a number of geological events. Considering this zircon characteristic, the U-Th-Pb data allow to distinguish four ages: 2116 ± 57 Ma, consistent with ages from the Paleoproterozoic West Central African Belt; 2551 ± 33 Ma which marks a late Neoarchean magmatic event; 2721 ± 27 Ma related to a Neoarchean magmatic even in Central Cameroon, similar to one found in the Congo Craton. A zircon core gives ages around 2925 Ma which provides some evidence of the presence of the Mesoarchean basement prior to the Neoarchean magmatism. A weighted average of lower intercepts ages gives a value of 821 ± 50 Ma, representing the age of later metamorphism event. The various characteristic group and related ages reflect not only the complexity of the history of the pyroxene amphibole gneiss, but also show that the meta-sediment has at least three zircon contributing

  8. Reactor operation safety information document

    Energy Technology Data Exchange (ETDEWEB)

    1990-01-01

    The report contains a reactor facility description which includes K, P, and L reactor sites, structures, operating systems, engineered safety systems, support systems, and process and effluent monitoring systems; an accident analysis section which includes cooling system anomalies, radioactive materials releases, and anticipated transients without scram; a summary of onsite doses from design basis accidents; severe accident analysis (reactor core disruption); a description of operating contractor organization and emergency planning; and a summary of reactor safety evolution. (MB)

  9. Joint KAERI/VAEC pre-possibility study on a new research reactor for Vietnam

    Energy Technology Data Exchange (ETDEWEB)

    Park, Cheol; Lee, B. C.; Chae, H. T.; Kim, H.; Lee, C. S.; Choi, C. O.; Jun, B. J. [KAERI, Taejon (Korea, Republic of); Vien, Luong Ba; Dien, Nguyen Nhi [Vietnam Atomic Energy Commission, Hanoi (Viet Nam)

    2004-05-01

    Based on the agreement on the technical cooperation for nuclear technology between Korea and Vietnam, a KAERI/VAEC joint study on the pre-possibility of a new research reactor for Vietnam has been carried out in the research reactor area from Nov. 2003 to May 2004. In this report, the results of the pre-possibility study on a new research reactor are described. The report presents the necessity of a new research reactor in Vietnam, and the desired performance requirements of the new research reactor if necessary. The major design characteristics of some existing research reactors and those under planning were also reviewed and the main characteristics which should be considered in selecting a new multipurpose research reactor for Vietnam were drawn. Some recommendations on the considerations for the next step of the feasibility study such as the project formulation, manpower requirements and international co-operation were also briefly touched upon.

  10. Irradiation response of delta ferrite in as-cast and thermally aged cast stainless steel

    Energy Technology Data Exchange (ETDEWEB)

    Li, Zhangbo; Lo, Wei-Yang [Department of Materials Science and Engineering, Nuclear Engineering Program, University of Florida, Gainesville, FL 32611 (United States); Chen, Yiren [Nuclear Engineering Division, Argonne National Laboratory, Lemont, IL 60439 (United States); Pakarinen, Janne [Belgian Nuclear Research Center (SCK-CEN), Boeretang 200, B-2400 Mol (Belgium); Wu, Yaqiao [Department of Materials Science and Engineering, Boise State University, Boise, ID 83715 (United States); Center for Advanced Energy Studies, Idaho Falls, ID 83401 (United States); Allen, Todd [Engineering Physics Department, University of Wisconsin, Madison, WI 53706 (United States); Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Yang, Yong, E-mail: yongyang@ufl.edu [Department of Materials Science and Engineering, Nuclear Engineering Program, University of Florida, Gainesville, FL 32611 (United States)

    2015-11-15

    To enable the life extension of Light Water Reactors (LWRs) beyond 60 years, it is critical to gain adequate knowledge for making conclusive predictions to assure the integrity of duplex stainless steel reactor components, e.g. primary pressure boundary and reactor vessel internal. Microstructural changes in the ferrite of thermally aged, neutron irradiated only, and neutron irradiated after being thermally aged cast austenitic stainless steels (CASS) were investigated using atom probe tomography. The thermal aging was performed at 400 °C for 10,000 h and the irradiation was conducted in the Halden reactor at ∼315 °C to 0.08 dpa (5.6 × 10{sup 19} n/cm{sup 2}, E > 1 MeV). Low dose neutron irradiation at a dose rate of 5 × 10{sup −9} dpa/s was found to induce spinodal decomposition in the ferrite of as-cast microstructure, and further to enhance the spinodal decomposition in the thermally aged cast alloys. Regarding the G-phase precipitates, the neutron irradiation dramatically increases the precipitate size, and alters the composition of the precipitates with increased, Mn, Ni, Si and Mo and reduced Fe and Cr contents. The results have shown that low dose neutron irradiation can further accelerate the degradation of ferrite in a duplex stainless steel at the LWR relevant condition.

  11. Irradiation response of delta ferrite in as-cast and thermally aged cast stainless steel

    Science.gov (United States)

    Li, Zhangbo; Lo, Wei-Yang; Chen, Yiren; Pakarinen, Janne; Wu, Yaqiao; Allen, Todd; Yang, Yong

    2015-11-01

    To enable the life extension of Light Water Reactors (LWRs) beyond 60 years, it is critical to gain adequate knowledge for making conclusive predictions to assure the integrity of duplex stainless steel reactor components, e.g. primary pressure boundary and reactor vessel internal. Microstructural changes in the ferrite of thermally aged, neutron irradiated only, and neutron irradiated after being thermally aged cast austenitic stainless steels (CASS) were investigated using atom probe tomography. The thermal aging was performed at 400 °C for 10,000 h and the irradiation was conducted in the Halden reactor at ∼315 °C to 0.08 dpa (5.6 × 1019 n/cm2, E > 1 MeV). Low dose neutron irradiation at a dose rate of 5 × 10-9 dpa/s was found to induce spinodal decomposition in the ferrite of as-cast microstructure, and further to enhance the spinodal decomposition in the thermally aged cast alloys. Regarding the G-phase precipitates, the neutron irradiation dramatically increases the precipitate size, and alters the composition of the precipitates with increased, Mn, Ni, Si and Mo and reduced Fe and Cr contents. The results have shown that low dose neutron irradiation can further accelerate the degradation of ferrite in a duplex stainless steel at the LWR relevant condition.

  12. REACTOR FUEL ELEMENTS TESTING CONTAINER

    Science.gov (United States)

    Whitham, G.K.; Smith, R.R.

    1963-01-15

    This patent shows a method for detecting leaks in jacketed fuel elements. The element is placed in a sealed tank within a nuclear reactor, and, while the reactor operates, the element is sparged with gas. The gas is then led outside the reactor and monitored for radioactive Xe or Kr. (AEC)

  13. RTNDT drop weight test of no ductility transition temperature for 12Cr2Mo1 alloy steel forgings used for the reactor internals in nuclear power station%核电站金属堆内构件设备用12Cr2Mo1合金钢锻件落锤试验

    Institute of Scientific and Technical Information of China (English)

    韩建成; 王毅; 吴志军; 石长仁

    2011-01-01

    12Cr2Mo1合金钢锻件是高温气冷堆核电站示范工程金属堆内构件设备所用主体材质,为保证该材质在反应堆整个寿命期内的性能不发生失效,在制造时要求进行无延性转变温度RTNDT小于等于-25℃的落锤试验,而国内锻件生产企业以往从无此技术要求的铬钼合金钢锻件制造经验和实践活动,为实现在高温堆核岛主设备上的成功应用通过试验和科研攻关掌握了其核心制造技术.介绍了12Cr2Mo1材质锻件在制造过程中所出现的质量问题、原因分析、采取的改进措施及取得成效等,将对今后GW级高温堆核电设备的国产化用材提供借鉴.%12Cr2Mol alloy steel forging material is widely used for reactor internals in the demonstrate project of high temperature gas-cooled reactor nuclear power plant.In order to avoid the faiure of properties of the material during the service life cycle of the reactor,drop weight test under the condition of RTNDT≤-25℃ was required for the material manufacturing.Chinese manufacturers have less experience in the manufacturing before the project.To master the key manufacturing technique,experiments and research were carried on.The quality problems occurred in the manufacturing of the material,the cause analysis,improvement approach,and the achieved effects were introduced,which will provide a good experience for the domestic material application in 1000 MW high temperature gas-cooled reactor nuclear power plant.

  14. WATER BOILER REACTOR

    Science.gov (United States)

    King, L.D.P.

    1960-11-22

    As its name implies, this reactor utilizes an aqueous solution of a fissionable element salt, and is also conventional in that it contains a heat exchanger cooling coil immersed in the fuel. Its novelty lies in the utilization of a cylindrical reactor vessel to provide a critical region having a large and constant interface with a supernatant vapor region, and the use of a hollow sleeve coolant member suspended from the cover assembly in coaxial relation with the reactor vessel. Cool water is circulated inside this hollow coolant member, and a gap between its outer wall and the reactor vessel is used to carry off radiolytic gases for recombination in an external catalyst chamber. The central passage of the coolant member defines a reflux condenser passage into which the externally recombined gases are returned and condensed. The large and constant interface between fuel solution and vapor region prevents the formation of large bubbles and minimizes the amount of fuel salt carried off by water vapor, thus making possible higher flux densities, specific powers and power densities.

  15. MULTISTAGE FLUIDIZED BED REACTOR

    Science.gov (United States)

    Jonke, A.A.; Graae, J.E.A.; Levitz, N.M.

    1959-11-01

    A multistage fluidized bed reactor is described in which each of a number of stages is arranged with respect to an associated baffle so that a fluidizing gas flows upward and a granular solid downward through the stages and baffles, whereas the granular solid stopsflowing downward when the flow of fluidizing gas is shut off.

  16. Fusion reactor materials

    Energy Technology Data Exchange (ETDEWEB)

    none,

    1989-01-01

    This paper discuses the following topics on fusion reactor materials: irradiation, facilities, test matrices, and experimental methods; dosimetry, damage parameters, and activation calculations; materials engineering and design requirements; fundamental mechanical behavior; radiation effects; development of structural alloys; solid breeding materials; and ceramics.

  17. Integral Fast Reactor concept

    Energy Technology Data Exchange (ETDEWEB)

    Till, C.E.; Chang, Y.I.

    1986-01-01

    The Integral Fast Reactor (IFR) is an innovative LMR concept, being developed at Argonne National Laboratory, that fully exploits the inherent properties of liquid metal cooling and metallic fuel to achieve breakthroughs in economics and inherent safety. This paper describes key features and potential advantages of the IFR concept, technology development status, fuel cycle economics potential, and future development path.

  18. The First Reactor.

    Science.gov (United States)

    Department of Energy, Washington, DC.

    On December 2, 1942, in a racquet court underneath the West Stands of Stagg Field at the University of Chicago, a team of scientists led by Enrico Fermi created the first controlled, self-sustaining nuclear chain reaction. This updated and revised story of the first reactor (or "pile") is based on postwar interviews (as told to Corbin…

  19. Thermal Reactor Safety

    Energy Technology Data Exchange (ETDEWEB)

    1980-06-01

    Information is presented concerning fire risk and protection; transient thermal-hydraulic analysis and experiments; class 9 accidents and containment; diagnostics and in-service inspection; risk and cost comparison of alternative electric energy sources; fuel behavior and experiments on core cooling in LOCAs; reactor event reporting analysis; equipment qualification; post facts analysis of the TMI-2 accident; and computational methods.

  20. Chromatographic and Related Reactors.

    Science.gov (United States)

    1988-01-07

    special information about effects of surface heteroge- neity in the methanation reaction. Studies of an efficient multicolumn assembly for measuring...of organic basic catalysts such as pyridine and 4-methylpicoline. It was demonstrated that the chromatographic reactor gave special information about...Programmed Reaction to obtain special information about surface heterogeneity in the methanation reaction. Advantages of stopped flow over steady state

  1. New concepts for shaftless recycle reactors

    Energy Technology Data Exchange (ETDEWEB)

    Berty, J.M.; Berty, I.J.

    1987-01-01

    Berty Reaction Engineers, Ltd. (BREL) is developing two new laboratory recycle reactors, the ROTOBERTY and the TURBOBERTY. These new reactors are basically improved versions of the original Berty reactor. To understand why the reactors have the features that they do, it is first necessary to briefly review laboratory reactors in general and specifically the original Berty reactor.

  2. Study on secondary shutdown systems in Tehran research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jalali, H.R.; Fadaei, A.H., E-mail: Fadaei_amir@aut.ac.ir; Gharib, M.

    2015-09-15

    Highlights: • A study was undertaken to summarize the techniques for secondary shutdown systems (SSS). • Neutronic calculation performed for proposed systems as SSS. • Dumping the heavy water stored in the reflector vessel is capable to shut down reactor. • Neutronic and transient calculation was done for validating the selected SSS. • All calculation shown that this system has advantages in safety and neutron economy. - Abstract: One important safety aspect of any research reactor is the ability to shut down the reactor. Usually, research reactors, currently in operation, have a single shutdown system based on the simultaneous insertion of the all control rods into the reactor core through gravity. Nevertheless, the International Atomic Energy Agency currently recommends use of two shutdown systems which are fully independent from each other to guarantee secure shutdown when one of them fails. This work presents an investigative study into secondary shutdown systems, which will be an important safety component in the research reactor and will provide another alternative way to shut down the reactor emergently. As part of this project, a study was undertaken to summarize the techniques that are currently used at world-wide research reactors for recognizing available techniques to consider in research reactors. Removal of the reflector, removal of the fuels, change in critical shape of reactor core and insertion of neutron absorber between the core and reflector are selected as possible techniques in mentioned function. In the next step, a comparison is performed for these methods from neutronic aspects. Then, chosen method is studied from the transient behavior point of view. Tehran research reactor which is a 5 MW open-pool reactor selected as a case study and all calculations are carried out for it. It has 5 control rods which serve the purpose of both reactivity control and shutdown of reactor under abnormal condition. Results indicated that heavy

  3. Reactor Dosimetry State of the Art 2008

    Science.gov (United States)

    Voorbraak, Wim; Debarberis, Luigi; D'Hondt, Pierre; Wagemans, Jan

    2009-08-01

    Oral session 1: Retrospective dosimetry. Retrospective dosimetry of VVER 440 reactor pressure vessel at the 3rd unit of Dukovany NPP / M. Marek ... [et al.]. Retrospective dosimetry study at the RPV of NPP Greifswald unit 1 / J. Konheiser ... [et al.]. Test of prototype detector for retrospective neutron dosimetry of reactor internals and vessel / K. Hayashi ... [et al.]. Neutron doses to the concrete vessel and tendons of a magnox reactor using retrospective dosimetry / D. A. Allen ... [et al.]. A retrospective dosimetry feasibility study for Atucha I / J. Wagemans ... [et al.]. Retrospective reactor dosimetry with zirconium alloy samples in a PWR / L. R. Greenwood and J. P. Foster -- Oral session 2: Experimental techniques. Characterizing the Time-dependent components of reactor n/y environments / P. J. Griffin, S. M. Luker and A. J. Suo-Anttila. Measurements of the recoil-ion response of silicon carbide detectors to fast neutrons / F. H. Ruddy, J. G. Seidel and F. Franceschini. Measurement of the neutron spectrum of the HB-4 cold source at the high flux isotope reactor at Oak Ridge National Laboratory / J. L. Robertson and E. B. Iverson. Feasibility of cavity ring-down laser spectroscopy for dose rate monitoring on nuclear reactor / H. Tomita ... [et al.]. Measuring transistor damage factors in a non-stable defect environment / D. B. King ... [et al.]. Neutron-detection based monitoring of void effects in boiling water reactors / J. Loberg ... [et al.] -- Poster session 1: Power reactor surveillance, retrospective dosimetry, benchmarks and inter-comparisons, adjustment methods, experimental techniques, transport calculations. Improved diagnostics for analysis of a reactor pulse radiation environment / S. M. Luker ... [et al.]. Simulation of the response of silicon carbide fast neutron detectors / F. Franceschini, F. H. Ruddy and B. Petrović. NSV A-3: a computer code for least-squares adjustment of neutron spectra and measured dosimeter responses / J. G

  4. Challenges and Innovative Technologies On Fuel Handling Systems for Future Sodium-Cooled Fast Reactors

    OpenAIRE

    Chassignet, Mathieu; Dumas, Sebastien; Penigot, Christophe; Prele, Gerard; Capitaine, Alain; Rodriguez, Gilles; Sanseigne, Emmanuel; Beauchamp, Francois

    2011-01-01

    International audience; The reactor refuelling system provides the means of transporting, storing, and handling reactor core subassemblies. The system consists of the facilities and equipment needed to accomplish the scheduled refuelling operations. The choice of a FHS impacts directly on the general design of the reactor vessel (primary vessel, storage, and final cooling before going to reprocessing), its construction cost, and its availability factor. Fuel handling design must take into acc...

  5. Measuring techniques in gas–liquid and gas–liquid–solid reactors

    OpenAIRE

    Boyer, Cristophe; Billet, Anne-Marie; Wild, Gabriel

    2002-01-01

    International audience; This article offers an overview of the intrumentation techniques developped for multiphase flow analysis either in gas/liquid or in gas/liquid/solid reactors. To characterize properly such reactors, experimental data have to be acquired at different space scale or time frequency. The existing multiphase flow metering described give information concerning reactor hydrodynamics such as pressure, phases holdups, phases velocities, flow regime, size and shape of dispersed ...

  6. Design of a micro-channel reactor for decomposition of organic pollutants in waste water treatment

    OpenAIRE

    Charles, Guillaume; Corbel, Serge; Carré, Marie-Christiane; Roques-Carmes, Thibault; Zahraa, Orfan

    2009-01-01

    International audience; Photocatalytic micro-channel reactor was built by using stereolithography process. A reactor with a micro-channel as a support of TiO2 photocatalyst was designed in order to reduce dimensions while improving the efficiency. Photocatalytic activity of the micro-reactor at various flow rates was evaluated by the inlet and outlet concentrations of salicylic acid as a model of pollutant. Influence of the initial pollutant concentration, the irradiation intensity on the rat...

  7. Modeling Chemical Reactors I: Quiescent Reactors

    CERN Document Server

    Michoski, C E; Schmitz, P G

    2010-01-01

    We introduce a fully generalized quiescent chemical reactor system in arbitrary space $\\vdim =1,2$ or 3, with $n\\in\\mathbb{N}$ chemical constituents $\\alpha_{i}$, where the character of the numerical solution is strongly determined by the relative scaling between the local reactivity of species $\\alpha_{i}$ and the local functional diffusivity $\\mathscr{D}_{ij}(\\alpha)$ of the reaction mixture. We develop an operator time-splitting predictor multi-corrector RK--LDG scheme, and utilize $hp$-adaptivity relying only on the entropy $\\mathscr{S}_{\\mathfrak{R}}$ of the reactive system $\\mathfrak{R}$. This condition preserves these bounded nonlinear entropy functionals as a necessarily enforced stability condition on the coupled system. We apply this scheme to a number of application problems in chemical kinetics; including a difficult classical problem arising in nonequilibrium thermodynamics known as the Belousov-Zhabotinskii reaction where we utilize a concentration-dependent diffusivity tensor $\\mathscr{D}_{ij}(...

  8. Important problems of future thermonuclear reactors*

    Directory of Open Access Journals (Sweden)

    Sadowski Marek J.

    2015-06-01

    Full Text Available This paper concerns important and difficult problems connected with a design and construction of thermonuclear reactors, which have to use nuclear fusion reactions of heavy isotopes of hydrogen, i.e., deuterium (D and tritium (T. There are described conditions in which such reactions can occur, and different methods of a high-temperature plasma generation, i.e., high-current electrical discharges, intense microwave pulses, and injection of energetic neutral atoms (NBI. There are also presented experimental facilities which can contain hot plasma for an appropriate period, and particularly so-called tokamaks. The second part presents the technical problems which must be solved in order to build a thermonuclear reactor, that might be used for energetic purposes. There are considered problems connected with a choice of constructional materials for a vacuum chamber, its internal parts, external windings generating a magnetic field, and necessary shields. The next part considers the handling of radioactive tritium; the using of alpha particles (4He for additional heating of plasma; recuperation of hydrogen isotopes absorbed in the tokamak internal parts, and a removal of a helium excess. There is presented a scheme of a future thermonuclear power plant and critical comments on a road map which should enable the construction of an industrial thermonuclear reactor (DEMO.

  9. Alternative approaches to fusion. [reactor design and reactor physics for Tokamak fusion reactors

    Science.gov (United States)

    Roth, R. J.

    1976-01-01

    The limitations of the Tokamak fusion reactor concept are discussed and various other fusion reactor concepts are considered that employ the containment of thermonuclear plasmas by magnetic fields (i.e., stellarators). Progress made in the containment of plasmas in toroidal devices is reported. Reactor design concepts are illustrated. The possibility of using fusion reactors as a power source in interplanetary space travel and electric power plants is briefly examined.

  10. International nuclear power status 2002; International kernekraftstatus 2002

    Energy Technology Data Exchange (ETDEWEB)

    Lauritzen, B.; Majborn, B.; Nonboel, E.; Oelgaard, P.L. (eds.)

    2003-03-01

    This report is the ninth in a series of annual reports on the international development of nuclear power with special emphasis on reactor safety. For 2002, the report contains: 1) General trends in the development of nuclear power; 2) Decommissioning of the nuclear facilities at Risoe National Laboratory: 3) Statistical information on nuclear power production (in 2001); 4) An overview of safety-relevant incidents in 2002; 5) The development in West Europe; 6) The development in East Europe; 7) The development in the rest of the world; 8) Development of reactor types; 9) The nuclear fuel cycle; 10) International nuclear organisations. (au)

  11. International nuclear power status 2001; International kernekraftstatus 2001

    Energy Technology Data Exchange (ETDEWEB)

    Lauritzen, B.; Majborn, B.; Nonboel, E.; Oelgaard, P.L. (eds.)

    2002-04-01

    This report is the eighth in a series of annual reports on the international development of nuclear power with special emphasis on reactor safety. For 2001, the report contains: 1) General trends in the development of nuclear power; 2) Nuclear terrorism; 3) Statistical information on nuclear power production (in 2000); 4) An overview of safety-relevant incidents in 2001; 5) The development in West Europe; 6) The development in East Europe; 7) The development in the rest of the world; 8) Development of reactor types; 9) The nuclear fuel cycle; 10) International nuclear organisations. (au)

  12. Reactor vessel support system. [LMFBR

    Science.gov (United States)

    Golden, M.P.; Holley, J.C.

    1980-05-09

    A reactor vessel support system includes a support ring at the reactor top supported through a box ring on a ledge of the reactor containment. The box ring includes an annular space in the center of its cross-section to reduce heat flow and is keyed to the support ledge to transmit seismic forces from the reactor vessel to the containment structure. A coolant channel is provided at the outside circumference of the support ring to supply coolant gas through the keyways to channels between the reactor vessel and support ledge into the containment space.

  13. Determinants of self-rated health in old age: A population-based, cross-sectional study using the International Classification of Functioning

    Directory of Open Access Journals (Sweden)

    Stenlund Hans

    2011-08-01

    Full Text Available Abstract Background Self-rated health (SRH is a widely used indicator of general health and multiple studies have supported the predictive validity of SRH in older populations concerning future health, functional decline, disability, and mortality. The aim of this study was to use the theoretical framework of the International Classification of Functioning, Disability and Health (ICF to create a better understanding of factors associated with SRH among community-dwelling older people in urban and rural areas. Methods The study design was population-based and cross-sectional. Participants were 185 Icelanders, randomly selected from a national registry, community-dwelling, 65-88 years old, 63% urban residents, and 52% men. Participants were asked: "In general, would you say your health is excellent, very good, good, fair, or poor?" Associations with SRH were analyzed with ordinal logistic regression. Explanatory variables represented aspects of body functions, activities, participation, environmental factors and personal factors components of the ICF. Results Univariate analysis revealed that SRH was significantly associated with all analyzed ICF components through 16 out of 18 explanatory variables. Multivariate analysis, however, demonstrated that SRH had an independent association with five variables representing ICF body functions, activities, and personal factors components: The likelihood of a better SRH increased with advanced lower extremity capacity (adjusted odds ratio [adjOR] = 1.05, p p = 0.040, household physical activity (adjOR = 1.01, p = 0.016, and older age (adjOR = 1.09, p = 0.006; but decreased with more depressive symptoms (adjOR = 0.79, p Conclusions The results highlight a collection of ICF body functions, activities and personal factors associated with higher SRH among community-dwelling older people. Some of these, such as physical capacity, depressive symptoms, and habitual physical activity are of particular interest due

  14. Lessons Learned From Developing Reactor Pressure Vessel Steel Embrittlement Database

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [ORNL

    2010-08-01

    Materials behaviors caused by neutron irradiation under fission and/or fusion environments can be little understood without practical examination. Easily accessible material information system with large material database using effective computers is necessary for design of nuclear materials and analyses or simulations of the phenomena. The developed Embrittlement Data Base (EDB) at ORNL is this comprehensive collection of data. EDB database contains power reactor pressure vessel surveillance data, the material test reactor data, foreign reactor data (through bilateral agreements authorized by NRC), and the fracture toughness data. The lessons learned from building EDB program and the associated database management activity regarding Material Database Design Methodology, Architecture and the Embedded QA Protocol are described in this report. The development of IAEA International Database on Reactor Pressure Vessel Materials (IDRPVM) and the comparison of EDB database and IAEA IDRPVM database are provided in the report. The recommended database QA protocol and database infrastructure are also stated in the report.

  15. Method of producing gaseous products using a downflow reactor

    Energy Technology Data Exchange (ETDEWEB)

    Cortright, Randy D; Rozmiarek, Robert T; Hornemann, Charles C

    2014-09-16

    Reactor systems and methods are provided for the catalytic conversion of liquid feedstocks to synthesis gases and other noncondensable gaseous products. The reactor systems include a heat exchange reactor configured to allow the liquid feedstock and gas product to flow concurrently in a downflow direction. The reactor systems and methods are particularly useful for producing hydrogen and light hydrocarbons from biomass-derived oxygenated hydrocarbons using aqueous phase reforming. The generated gases may find used as a fuel source for energy generation via PEM fuel cells, solid-oxide fuel cells, internal combustion engines, or gas turbine gensets, or used in other chemical processes to produce additional products. The gaseous products may also be collected for later use or distribution.

  16. Pebble Bed Reactor Dust Production Model

    Energy Technology Data Exchange (ETDEWEB)

    Abderrafi M. Ougouag; Joshua J. Cogliati

    2008-09-01

    The operation of pebble bed reactors, including fuel circulation, can generate graphite dust, which in turn could be a concern for internal components; and to the near field in the remote event of a break in the coolant circuits. The design of the reactor system must, therefore, take the dust into account and the operation must include contingencies for dust removal and for mitigation of potential releases. Such planning requires a proper assessment of the dust inventory. This paper presents a predictive model of dust generation in an operating pebble bed with recirculating fuel. In this preliminary work the production model is based on the use of the assumption of proportionality between the dust production and the normal force and distance traveled. The model developed in this work uses the slip distances and the inter-pebble forces computed by the authors’ PEBBLES. The code, based on the discrete element method, simulates the relevant static and kinetic friction interactions between the pebbles as well as the recirculation of the pebbles through the reactor vessel. The interaction between pebbles and walls of the reactor vat is treated using the same approach. The amount of dust produced is proportional to the wear coefficient for adhesive wear (taken from literature) and to the slip volume, the product of the contact area and the slip distance. The paper will compare the predicted volume with the measured production rates. The simulation tallies the dust production based on the location of creation. Two peak production zones from intra pebble forces are predicted within the bed. The first zone is located near the pebble inlet chute due to the speed of the dropping pebbles. The second peak zone occurs lower in the reactor with increased pebble contact force due to the weight of supported pebbles. This paper presents the first use of a Discrete Element Method simulation of pebble bed dust production.

  17. Methanogenesis in Thermophilic Biogas Reactors

    DEFF Research Database (Denmark)

    Ahring, Birgitte Kiær

    1995-01-01

    Methanogenesis in thermophilic biogas reactors fed with different wastes is examined. The specific methanogenic activity with acetate or hydrogen as substrate reflected the organic loading of the specific reactor examined. Increasing the loading of thermophilic reactors stabilized the process...... as indicated by a lower concentration of volatile fatty acids in the effluent from the reactors. The specific methanogenic activity in a thermophilic pilot-plant biogas reactor fed with a mixture of cow and pig manure reflected the stability of the reactor. The numbers of methanogens counted by the most...... against Methanothrix soehngenii or Methanothrix CALS-I in any of the thermophilic biogas reactors examined. Studies using 2-14C-labeled acetate showed that at high concentrations (more than approx. 1 mM) acetate was metabolized via the aceticlastic pathway, transforming the methyl-group of acetate...

  18. Uso de detectores de neutrinos para el monitoreo de reactores nucleares Uso de detectores de neutrinos para el monitoreo de reactores nucleares

    Directory of Open Access Journals (Sweden)

    Gerardo Moreno

    2012-02-01

    Full Text Available Se estudia la factibilidad del uso de los detectores de antineutrinos para el monitoreo de reactores nucleares. Usando un modelo sencillo de cascada de fisión a dos componentes, se ilustra la dependencia del número de antineutrinos detectados a una distancia L del reactor según la composición nuclear del combustible. Se explica el principio de detección de neutrinos de reactores en base al decaimiento beta inverso y se describe como los detectores de neutrinos pueden emplearse para el monitoreo de la producción de materiales fisibles en el reactor. Se comenta como generalizar este análisis al caso real de un reactor nuclear in situ y uno de los principales experimentos internacionales dedicados a este propósito. We study the feasibility to use antineutrinos detectors for monitoring of nuclear reactors. Using a simple model of fission shower with two components, we illustrate how the numbers of antineutrinos detected at a distance L from the reactor depend on the composition of the nuclear combustible. We explain the principles of reactor neutrino detection using inverse beta decays and we describe how neutrinos detectors can be used for monitoring the production of fissile materials within the reactors. We comment how to generalize this analysis to the realistic case of a nuclear reactor in situ and one of the main international experiments dedicated to study the use of neutrinos detectors as nuclear safeguards.

  19. BENCHMARK EVALUATION OF THE START-UP CORE REACTOR PHYSICS MEASUREMENTS OF THE HIGH TEMPERATURE ENGINEERING TEST REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    John Darrell Bess

    2010-05-01

    The benchmark evaluation of the start-up core reactor physics measurements performed with Japan’s High Temperature Engineering Test Reactor, in support of the Next Generation Nuclear Plant Project and Very High Temperature Reactor Program activities at the Idaho National Laboratory, has been completed. The evaluation was performed using MCNP5 with ENDF/B-VII.0 nuclear data libraries and according to guidelines provided for inclusion in the International Reactor Physics Experiment Evaluation Project Handbook. Results provided include updated evaluation of the initial six critical core configurations (five annular and one fully-loaded). The calculated keff eigenvalues agree within 1s of the benchmark values. Reactor physics measurements that were evaluated include reactivity effects measurements such as excess reactivity during the core loading process and shutdown margins for the fully-loaded core, four isothermal temperature reactivity coefficient measurements for the fully-loaded core, and axial reaction rate measurements in the instrumentation columns of three core configurations. The calculated values agree well with the benchmark experiment measurements. Fully subcritical and warm critical configurations of the fully-loaded core were also assessed. The calculated keff eigenvalues for these two configurations also agree within 1s of the benchmark values. The reactor physics measurement data can be used in the validation and design development of future High Temperature Gas-cooled Reactor systems.

  20. Closed Fuel Cycle and Minor Actinide Multirecycling in a Gas-Cooled Fast Reactor

    NARCIS (Netherlands)

    Van Rooijen, W.F.G.; Kloosterman, J.L.

    2009-01-01

    The Generation IV International Forum has identified the Gas-Cooled Fast Reactor (GCFR) as one of the reactor concepts for future deployment. The GCFR targets sustainability, which is achieved by the use of a closed nuclear fuel cycle where only fission products are discharged to a repository; all H