78 FR 7439 - Proposed Alterations to Pawtucket Dam on the Merrimack River in Lowell, MA; Public Meeting

Science.gov (United States)

2013-02-01

..., Lowell National Historical Park, Boott Mills, Second Floor, 115 John Street, Lowell, MA 01852. FOR... out, financially assist, or license on historic properties. When, as is the case here, an undertaking... the Lowell Historic Preservation District (LHPD) and the National Park Service's List of Classified...

Technical description of other types of reactors

International Nuclear Information System (INIS)

Vollmer, H.

1977-01-01

The paper reviews the development of reactor systems other than LWR, i. e. gas cooled reactors, heavy water reactors and fast breeders. The specific features of these reactors are discussed. Technical details on plant design of the various systems will be given as well as the present status-of-the-art. (orig.) [de

As Big and As Good As It Gets: The Large Monolithic Imager for Lowell Observatory's 4.3-m Discovery Channel Telescope

Science.gov (United States)

Massey, Philip; Dunham, E. W.; Bida, T. A.; Collins, P.; Hall, J. C.; Hunter, D. A.; Lauman, S.; Levine, S.; Neugent, K.; Nye, R.; Oliver, R.; Schleicher, D.; Zoonematkermani, S.

2013-01-01

The Large Monolithic Imager (LMI), a camera built at Lowell Observatory, is currently undergoing commissioning on Lowell's new 4.3-m Discovery Channel Telescope (DCT). At the heart of the LMI is the largest charge-coupled device (CCD) that can be built using current fabrication techniques, and the first of its kind to be made by e2v. The active area of the chip is 92.2mmx92.4mm, and has 6144 by 6160 15-micron pixels. Our choice of a single chip over a mosaic of smaller ones was inspired by the success of USNO in deploying a similarly ginormous device made by Semiconductor Technology Associates, Inc. There are some significant advantages that a (very!) large single CCD has over a mosaic of smaller ones. With a mosaic, one has to dither to fill in the gaps between the chips for complete areal coverage. This is not only costly in overhead, but it also poses a limitation in faint surface brightness studies, as the sky brightness is constantly changing during the dithering process. In addition, differences in the wavelength dependence of the DQE can lead to differences in the color terms from chip to chip in mosaics, requiring one to deal with each chip as a separate instrument (see the Local Group Galaxy photometry of Massey et al. 2006, AJ, 131, 2478). The LMI avoids these problems. The Discovery Channel Telescope is being built by Lowell Observatory in partnership with Discovery Communications. First light took place in May 2012. Institutional DCT partners include Boston University (in perpetuity), the University of Maryland, and the University of Toledo. More about the DCT can be found in the adjacent poster by Hall et al. The LMI has been made possible thanks to a National Science Foundation grant (AST-1005313). We are currently doing on-sky evaluation of the camera, as commissioning of the DCT progresses, determining color terms, photometric zero-points, astrometric characteristics, etc. We will present these results, along with technical details and many pretty

Research reactors spent fuel management in the Nuclear Research Institute Rez

International Nuclear Information System (INIS)

Rychecky, J.

2001-01-01

In Czech Republic 3 research and testing nuclear reactors are operated at present time, with the biggest one being the Nuclear Research Institute (NRI) reactor LVR-15, operated with maximum power 10 MW. This reactor serves as a radiation source for material testing, producing of ionizing radiation sources, theoretical studies, and, most recently, for boron neutron capture therapy. Another NRI reactor LR-0 is a reactor of zero power used mainly for the studies of WWER 1000 spent fuel criticality. For training of students the reactor called VRABEC (VR-1), operated also with very low power, serves since 1990 at the Faculty of Nuclear Engineering, of Czech Technical University. The similar testing type reactor (SR-0), already decommissioned, was also used since 1974 to 1989 in Skoda, Nuclear Machinery, Plzen. This contribution summarizes the present state of the spent fuel (SF) management of these nuclear reactors. As the SF management is different for very low or zero power reactors and power reactors, the first type will be only briefly discussed, and then the main attention will be devoted to SF management of the NRI experimental reactor LVR-15

Research reactors spent fuel management in the Nuclear Research Institute Rez

Energy Technology Data Exchange (ETDEWEB)

Rychecky, J. [Nuclear Research Institute, 25068 Rez (Czech Republic)

2001-07-01

In Czech Republic 3 research and testing nuclear reactors are operated at present time, with the biggest one being the Nuclear Research Institute (NRI) reactor LVR-15, operated with maximum power 10 MW. This reactor serves as a radiation source for material testing, producing of ionizing radiation sources, theoretical studies, and, most recently, for boron neutron capture therapy. Another NRI reactor LR-0 is a reactor of zero power used mainly for the studies of WWER 1000 spent fuel criticality. For training of students the reactor called VRABEC (VR-1), operated also with very low power, serves since 1990 at the Faculty of Nuclear Engineering, of Czech Technical University. The similar testing type reactor (SR-0), already decommissioned, was also used since 1974 to 1989 in Skoda, Nuclear Machinery, Plzen. This contribution summarizes the present state of the spent fuel (SF) management of these nuclear reactors. As the SF management is different for very low or zero power reactors and power reactors, the first type will be only briefly discussed, and then the main attention will be devoted to SF management of the NRI experimental reactor LVR-15.

Netherlands Interuniversity Reactor Institut

International Nuclear Information System (INIS)

1978-01-01

This is the annual report of the Interuniversity Reactor Institute in the Netherlands for the Academic Year 1977-78. Activities of the general committee, the daily committee and the scientific advice board are presented. Detailed reports of the scientific studies performed are given under five subjects - radiation physics, reactor physics, radiation chemistry, radiochemistry and radiation hygiene and dosimetry. Summarised reports of the various industrial groups are also presented. Training and education, publications and reports, courses, visits and cooperation with other institutes in the area of scientific research are mentioned. (C.F.)

Design, Manufacturing and Irradiation Behaviour of Fast Reactor Fuel. Proceedings of a Technical Meeting

International Nuclear Information System (INIS)

2013-04-01

Fast reactors are vital for ensuring the sustainability of nuclear energy in the long term. They offer vastly more efficient use of uranium resources and the ability to burn actinides, which are otherwise the long-lived component of high level nuclear waste. These reactors require development, qualification, testing and deployment of improved and innovative nuclear fuel and structural materials having very high radiation resistance, corrosion/erosion and other key operational properties. Several IAEA Member States have made efforts to advance the design and manufacture of technologies of fast reactor fuels, as well as to investigate their irradiation behaviour. Due to the acute shortage of fast neutron testing and post-irradiation examination facilities and the insufficient understanding of high dose radiation effects, there is a need for international exchange of knowledge and experience, generation of currently missing basic data, identification of relevant mechanisms of materials degradation and development of appropriate models. Considering the important role of nuclear fuels in fast reactor operation, the IAEA Technical Working Group on Fuel Performance and Technology (TWGFPT) proposed a Technical Meeting (TM) on 'Design, Manufacturing and Irradiation Behaviour of Fast Reactors Fuels', which was hosted by the Institute of Physics and Power Engineering (IPPE) in Obninsk, Russian Federation, from 30 May to 3 June 2011. The TM included a technical visit to the fuel production plant MSZ in Elektrostal. The purpose of the meeting was to provide a forum to share knowledge, practical experience and information on the improvement and innovation of fuels for fast reactors through scientific presentations and brainstorming discussions. The meeting brought together 34 specialists from national nuclear agencies, R and D and design institutes, fuel vendors and utilities from 10 countries. The presentations were structured into four sections: R and D Programmes on FR Fuel

Technical modifications and management innovations in exporting nuclear reactor projects

International Nuclear Information System (INIS)

Mao Xiaoming; Qin Xijiu; Ding Hu; Xue Zhaoqun; Wen Shengjun

2009-01-01

As a main channel for the foreign economic cooperation of China nuclear industry, China Zhongyuan Engineering Corporation (CZEC) has been constantly engaged in technical modifications and management innovations in its exporting nuclear reactor projects. In the implementation of heavy water research reactor contract in Algeria, CZEC had established a complete and adequate design standards system in compliance with the international standards, and made significant modifications to the reference reactor in the aspects of reactor power and reactor safety, solved quite some technical issues which-affected the reactor technical performance. The modifications and improvements enabled the technical parameters, safety features, reactor multipurpose application to attain to the advanced level in the world. In the 300 MWe PWR NPPs in Pakistan, safety features had been updated in line with upgrading regulatory requisites. The design philosophy and technology application demonstrated CZEC' s creation and innovation on basis of constant safety enhancement of nuclear power projects. Efforts had also been made by CZEC' s creation and innovation on basis of constant safety enhancement of nuclear power projects. Efforts had also been made by CZEC in promoting China made equipment items and components exportation. (authors)

Results of the scientific and technical activities of the Nuclear Reactors and Thermal Physics Institute for 2014. Scientific and technical collection

International Nuclear Information System (INIS)

Trufanov, A.A.; Sorokin, A.P.; Vereshchagina, T.N.

2015-01-01

In the collection there are the main results of research and development obtained by the researchers of the Nuclear Reactors and Thermal Physics Institute FSUE SSC RF - IPPE in 2014, the problems and questions of further investigations are formulated and discussed. Considerable body of data on neutronic, thermohydraulic and technological studies carried out in the frameworks of Proryv project are presented, calculational and experimental justification of design choices and safety of projects on RU BN-1200, multipurpose research reactor MBIR with sodium coolant, RU BREST-OD-300 with lead coolant are among them. The results of experimental and calculational thermophysical investigations in justification of operation conditions and safety of nuclear power plants with water-cooled reactors (WWER-1000, WWER-TOI), pilot studies on innovation project WWER-SKD with supercritical water, in justification of thermonuclear reactor blanket are given [ru

Assessment of the technical viability of reactor options for plutonium disposition

International Nuclear Information System (INIS)

Primm, R.T. III.

1996-01-01

Various reactor concepts for the disposition of surplus Pu have been proposed by reactor vendors; not all have attained the same level of technical viability. Studies were performed to differentiate between reactor concepts by devising a quantitative index for technical viability. For a quantitative assessment, three issues required resolution: the definition of a technical maturity scale, the treatment of ''subjective'' factors which cannot be easily represented in a quantitative format, and the protocol for producing a single technical viability figure-of-merit for each alternative. Alternatives involving the use of foreign facilities were found to be the most technically viable

10 years Institute for Reactor Development

International Nuclear Information System (INIS)

1975-05-01

Ten years ago the Institute of Reactor Development was founded. This report contains a review about the research work of the institute in these past ten years. The work was mainly performed within the framework of the Fast Breeder Project, the Nuclear Safety Project and Computer Aided Design. Especially the following topics are discussed: design studies for different fast breeder reactors, development works for fast breeders, investigations of central safety problems of sodium cooled breeder reactors (such as local and integral coolant disturbances and hypothetical accident analysis), special questions of light water reactor safety (such as dynamic stresses in pressure suppression systems and fuel rod behaviour under loss of coolant conditions), and finally computer application in various engineering fields. (orig.) [de

Savannah River Site production reactor technical specifications. K Production Reactor

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-02-01

These technical specifications are explicit restrictions on the operation of the Savannah River Site K Production Reactor. They are designed to preserve the validity of the plant safety analysis by ensuring that the plant is operated within the required conditions bounded by the analysis, and with the operable equipment that is assumed to mitigate the consequences of an accident. Technical specifications preserve the primary success path relied upon to detect and respond to accidents. This report describes requirements on thermal-hydraulic limits; limiting conditions for operation and surveillance for the reactor, power distribution control, instrumentation, process water system, emergency cooling and emergency shutdown systems, confinement systems, plant systems, electrical systems, components handling, and special test exceptions; design features; and administrative controls.

Technical report: technical development on the silicide plate-type fuel experiment at nuclear safety research reactor

International Nuclear Information System (INIS)

Yanagisawa, Kazuaki; Soyama, Kazuhiko; Ichikawa, Hiroki

1991-08-01

According to a reduction of fuel enrichment from 45 w/o 235 U to 20 w/o, an aluminide plate-type fuel used currently in the domestic research and material testing reactors will be replaced by a silicide plate-type one. One of the major concern arisen from this alternation is to understand the fuel behavior under simulated reactivity initiated accident (RIA) conditions, this is strongly necessary from the safety and licensing point of view. The in-core RIA experiments are, therefore, carried out at Nuclear Safety Research Reactor (NSRR) in Japan Atomic Energy Research Institute (JAERI). The silicide plate-type fuel consisted of the ternary alloy of U-Al-Si as a meat with uranium density up to 4.8 g/cm 3 having thickness by 0.51 mm and the binary alloy of Al-3%Mg as a cladding by thickness of 0.38 mm. Comparison of the physical properties of this metallic plate fuel with the UO 2 -zircaloy fuel rod used conventionally in commercial light water reactors shows that the heat conductivity of the former is of the order of about 13 times greater than the latter, however the melting temperature is only one-half (1570degC). Prior to in-core RIA experiments, there were some difficulties lay in our technical path. This report summarized the technical achievements obtained through our four years work. (J.P.N.)

Report of results on the research and development work 1979 of the Institute for Technical Chemistry

International Nuclear Information System (INIS)

1980-02-01

The results report on the research and development work in 1979 carried out at the Institute for Technical Physics of the Nuclear Research Centre, Karlsruhe is concerned here. The main field of this development work is the research into superconducting magnets for fusion reactors. Studies are published on the material, processing and shape of these magnets. Furthermore, a report is given on fusion magnet technology, superconducting fundamentals and technical superconductors, as well as an cryonergy technique and cryotechnique. (KBI) [de

High Flux Isotope Reactor technical specifications

International Nuclear Information System (INIS)

1985-11-01

This report gives technical specifications for the High Flux Isotope Reactor (HFIR) on the following: safety limits and limiting safety system settings; limiting conditions for operation; surveillance requirements; design features; and administrative controls

Annual report 1986 Interfacultary Reactor Institute

International Nuclear Information System (INIS)

1987-01-01

In this annual report of the Dutch Interfacultary Reactor Institute, summary- and detailed reports are presented of Current research during 198? of the departments radiochemistry, radiation chemistry, radiation physics, and reactor physics. (H.W.). refs.; figs.; tabs

NRC review of Electric Power Research Institute's Advanced Light Reactor Utility Requirements Document - Program summary, Project No. 669

International Nuclear Information System (INIS)

1992-08-01

The staff of the US Nuclear Regulatory Commission has prepared Volume 1 of a safety evaluation report (SER), ''NRC Review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document -- Program Summary,'' to document the results of its review of the Electric Power Research Institute's ''Advanced Light Water Reactor Utility Requirements Document.'' This SER provides a discussion of the overall purpose and scope of the Requirements Document, the background of the staff's review, the review approach used by the staff, and a summary of the policy and technical issues raised by the staff during its review

Annual report 1984 Interfacultary Reactor Institute

International Nuclear Information System (INIS)

1985-01-01

In this annual report of the Dutch Interfacultary Reactor Institute, summary- and detailed reports are presented of current research during 1984 of the departments radiochemistry, radiation chemistry, radiation physics, reactor physics, and radiation hygiene and dosimetry. (H.W.). refs.; figs.; tabs

Technical specifications: Health Physics Research Reactor

International Nuclear Information System (INIS)

1986-03-01

These technical specifications define the key limitations that must be observed for safe operation of the Health Physics Research Reactor (HPRR) and an envelope of operation within which there is assurance that these limits will not be exceeded

Report on generation IV technical working group 3 : liquid metal reactors

International Nuclear Information System (INIS)

Lineberry, M. J.; Rosen, S. L.; Sagayama, Y.

2002-01-01

This paper reports on the first round of R and D roadmap activities of the Generation IV (Gen IV) Technical Working Group (TWG) 3, on liquid metal-cooled reactors. Liquid metal coolants give rise to fast spectrum systems, and thus the reactor systems considered in this TWG are all fast reactors. Gas-cooled fast reactors are considered in the context of TWG 2. As is noted in other Gen IV papers, this first round activity is termed ''screening for potential'', and includes collecting the most complete set of liquid metal reactor/fuel cycle system concepts possible and evaluating the concepts against the Gen IV principles and goals. Those concepts or concept groups that meet the Gen IV principles and which are deemed to have reasonable potential to meet the Gen IV goals will pass to the next round of evaluation. Although we sometimes use the terms ''reactor'' or ''reactor system'' by themselves, the scope of the investigation by TWG 3 includes not only the reactor systems, but very importantly the closed fuel recycle system inevitably required by fast reactors. The response to the DOE Request for Information (RFI) on liquid metal reactor/fuel cycle systems from principal investigators, laboratories, corporations, and other institutions, was robust and gratifying. Thirty three liquid metal concept descriptions, from eight different countries, were ultimately received. The variation in the scope, depth, and completeness of the responses created a significant challenge for the group, but the TWG made a very significant effort not to screen out concepts early in the process

A String-Inspired Model for the Low-$\\ell$ CMB

CERN Document Server

Kitazawa, N.

2015-07-09

We present a semi--analytic exploration of some low--$\\ell$ angular power spectra inspired by "Brane Supersymmetry Breaking". This mechanism splits Bose and Fermi excitations in String Theory, leaving behind an exponential potential that is just too steep for the inflaton to emerge from the initial singularity while descending it. As a result, the scalar generically bounces against the exponential wall, which typically introduces an infrared depression and a pre--inflationary peak in the power spectrum of scalar perturbations. We elaborate on a possible link between this phenomenon and the low--$\\ell$ CMB. For the first 32 multipoles, combining the hard exponential with a milder one leading to $n_s\\simeq 0.96$ and with a small gaussian bump we have attained a reduction of $\\chi^{\\,2}$ to about 46% of the standard $\\Lambda$CDM setting, with both WMAP9 and PLANCK 2013 data. This result corresponds to a $\\chi^{\\,2}/DOF$ of about 0.45, to be compared with a $\\Lambda$CDM value of about 0.85. The preferred choices ...

Technical and management challenges associated with structural materials degradation in nuclear reactors in the future

International Nuclear Information System (INIS)

Ford, F.P.

2007-01-01

issue is compounded by the fact that some of the future GEN IV reactor designs involve fast neutron spectra, and all involve increases in temperature to the range 500 o C - 1250 o C. Comparatively little is known of the effect of, for instance, creep-fatigue interactions in high irradiation fluxes on the structural integrity of the potential materials of construction. In spite of these technical concerns there is the business management expectation that all of these reactors will experience very few materials degradation problems that might affect the economics of operation. The paper starts with a review of our present capability to predict the materials degradation modes encountered in the current BWR and PWR reactor designs. This capability is the basis for any analysis of the future degradation problems (and their mitigation) in the current reactors and in the evolutionary water-cooled reactor designs. This section concludes with an overview of assessments of future materials degradation issues that might be expected in these water-cooled reactors. These preliminary discussions are then broadened to cover some of the more obvious technical problems likely to be encountered with the more advanced GEN IV designs, such as the Very High Temperature Reactor (VHTR) and the Super Critical Water Cooled Reactor (SCWR). The article concludes with a brief discussion of some of the challenges facing the technical management/leadership, with some suggestions on how to overcome them. These challenges may become especially severe given the fact that the technical problems must be overcome in a time frame that is short compared with that taken to resolve the issues that have faced us over the last 30 years. Some specific management challenges include: The decrease in the number of experienced experimentalists and analysts over the last 10 years; The decrease in 'institutional' memory as it relates to the operation of the current reactors, and the design and construction of

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

International Nuclear Information System (INIS)

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

2004-01-01

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

Technical evaluation of corium cooling at the reactor cavity

International Nuclear Information System (INIS)

Yang, Soo Hyung; Chan, Eun Sun; Lee, Jae Hun; Lee, Jong In

1998-01-01

To terminate the progression of the severe accident and mitigate the accident consequences, corium cooling has been suggested as one of most important design features considered in the severe accident mitigation. Till now, some kinds of cooling methodologies have been identified and, specially, the corium cooling at the reactor cavity has been considered as one of the most promising cooling methodologies. Moreover, several design requirements related to the corium cooling at the reactor cavity have been also suggested and applied to the design of the next generation reactor. In this study, technical descriptions are briefly described for the important issues related to the corium cooling at the reactor cavity, i.e. cavity area, cavity flooding system, etc., and simple evaluations for those items have been performed considering present technical levels including the experiment and analytical works

Scope of activities and organization of an interuniversity reactor institute

International Nuclear Information System (INIS)

de Bruin, M.

1990-01-01

The Reactor Instituut Delft was founded in 1958 and was at that time part of the Delft University of Technology. In 1969, the institute was converted into an interuniversity institute, owned and directed by the combined Dutch universities. Since 1987, the institute has again constituted part of the Delft University of Technology, still continuing its role as an interuniversity institute and with provisions in the organizational structure to secure this role. The major facility of the institute is the Hoger Onderwijs reactor, a 2-MW swimming pool reactor operated 24 h/day, 5 days/week. The reactor is used in neutron beam studies, reactor physics research, and for isotope production, neutron activation analysis (NAA), and commercial irradiation. The institute's 3-MeV Van de Graaff electron accelerator is mainly used for radiation chemistry. It can deliver subnanosecond high-current pulses and is provided with fast optical and conductivity measuring equipment. The variable energy positron source is being used for the study of defects at metal surfaces and interfaces. The experience obtained with this source is used in the development of a much stronger source as the basis of a positron microbeam in one of the reactor beam tubes

Factors Affecting Women Enrolment In Technical Institutions In Tanzania A Case Study Of Arusha Technical College

OpenAIRE

Glory B. Kaaya; Dr. Esther Waiganjo

2015-01-01

Abstract The purpose of this study was to investigate the factors affecting women enrolment in Technical Institutions in Tanzania by focusing on the women. The study examined in detail the factors affecting women enrolment in Technical institutions and Arusha Technical College in particular as well as exploring sexual dynamics within Technical Institutions. Moreover a number of theoretical frameworks were concerned basing on the objectives a case study design was employed involving both qua...

Conversion of research and test reactors to low enriched uranium fuel: technical overview and program status

International Nuclear Information System (INIS)

Roglans-Ribas, J.

2008-01-01

Many of the nuclear research and test reactors worldwide operate with high enriched uranium fuel. In response to worries over the potential use of HEU from research reactors in nuclear weapons, the U.S Department of Energy (DOE) initiated a program - the Reduced Enrichment for Research and Test Reactors (RERTR) - in 1978 to develop the technology necessary to reduce the use of HEU fuel by converting research reactors to low enriched uranium (LEU) fuel. The Reactor Conversion program is currently under the DOE's National Nuclear Security Administration's Global Threat Reduction Initiative (GTRI). 55 of the 129 reactors included in the scope have been already converted to LEU fuel or have shutdown prior to conversion. The major technical activities of the Conversion Program include: (1) the development of advanced LEU fuels; (2) conversion analysis and conversion support; and (3) technology development for the production of Molybdenum-99 (Mo 99 ) with LEU targets. The paper provides an overview of the status of the program, the technical challenges and accomplishments, and the role of international collaborations in the accomplishment of the Conversion Program objectives. Nuclear research and test reactors worldwide have been in operation for over 60 years. Many of these facilities operate with high enriched uranium fuel. In response to increased worries over the potential use of HEU from research reactors in the manufacturing of nuclear weapons, the U.S Department of Energy (DOE) initiated a program - the Reduced Enrichment for Research and Test Reactors (RERTR) - in 1978 to develop the technology necessary to reduce the use of HEU fuel in research reactors by converting them to low enriched uranium (LEU) fuel. The reactor conversion program was initially focused on U.S.-supplied reactors, but in the early 1990s it expanded and began to collaborate with Russian institutes with the objective of converting Russian supplied reactors to the use of LEU fuel.

Standard technical specifications for Westinghouse pressurized water reactors (revision issued Fall 1981). Technical report

International Nuclear Information System (INIS)

Virgilio, M.J.

1981-11-01

The Standard Technical Specifications for Westinghouse Pressurized Water Reactors (W-STS) is a generic document prepared by the U.S. NRC for use in the licensing process of current Westinghouse Pressurized Water Reactors. The W-STS sets forth the Limits, Operating Conditions and other requirements applicable to nuclear reactor facility operation as set forth in Section 50.36 of 10 CFR Part 50 for the protection of the health and safety of the public

Technical and institutional preparedness for introduction of evolutionary water cooled reactors

International Nuclear Information System (INIS)

Jones, R.L.; Machiels, A.J.; Vine, G.V.

1999-01-01

Since 1982, US utilities have been leading an industry-wide effort to establish a technical foundation for the design of the next generation of light water reactors in the United States: the Advanced Light Water Reactor (ALWR) Program. This program provided a foundation for a comprehensive initiative for revitalizing Nuclear Power in the US, as set forth in the Nuclear Energy Industry's 'Strategic Plan for Building New Nuclear Power Plants'. The Strategic Plan contains fourteen building blocks, each of which is considered essential to building new nuclear plants. At its inception, the ALWR Program envisioned new plant orders in the US shortly after the turn of the century, and geared its milestones and deliverables to enabling ALWRs as an option for utilities by about 2000. However, in the US, new orders for nuclear plants are not imminent. There are three primary reasons for this - the lack of demand today for major new construction of baseload capacity, the economic and structural uncertainty associated with deregulation, and the lack of an assured resolution to public concerns over long term management of nuclear waste. Deregulation will likely drive further consolidation of the electricity business, as evidenced in recent nuclear utility mergers, acquisitions, and plant purchases by the larger utilities intent on remaining in the generation business. Deregulation has focused attention on some of the inefficiencies in the current regulatory regime for nuclear energy, and is likely to drive the US government to find more efficient and less expensive ways of providing adequate protection of public health and safety. Deregulation has also focused the industry on the significant variations in production costs among plants, fueling the belief that the industry as a whole can make further improvements in this area to match the stable, low cost performance of the top ten plants. Finally, deregulation has focused the nuclear industry on the imperative for ensuring that

Standard Technical Specifications for Westinghouse pressurized water reactors

International Nuclear Information System (INIS)

Virgilio, M.J.

1980-09-01

The Standard Technical Specifications for Westinghouse Pressurized Water Reactors (W-STS) is a generic document prepared by the U.S. NRC for use in the licensing process of current Westinghouse Pressurized Water Reactors. The W-STS sets forth the Limits, Operating Conditions and other requirements applicable to nuclear reactor facility operation as set forth in by Section 50.36 of 10 CFR Part 50 for the protection of the health and safety of the public. This document is revised periodically to reflect current licensing requirements

Technical specifications for the bulk shielding reactor

International Nuclear Information System (INIS)

1986-05-01

This report provides information concerning the technical specifications for the Bulk Shielding Reactor. Areas covered include: safety limits and limiting safety settings; limiting conditions for operation; surveillance requirements; design features; administrative controls; and monitoring of airborne effluents. 10 refs

Technical and safe development features of modern research reactor

International Nuclear Information System (INIS)

Wang Jiaying; Dong Duo

1998-01-01

The development trend of research reactor in the world, and development situation in China are introduced. Up to now, some research reactors have serviced for long time and equipment have aged, not to be satisfied for requirement of science and technology development. New research reactors must been developed. The technical features and safe features of new type research reactor in China, for example: multi-pile utilization, compact core of high flux, high automation level of control, reactor two independent shutdown systems, great coefficient of negative temperature, passive safety systems, reliable residual heat removal system are studied

Technical meeting on decommissioning of fast reactors after sodium draining. Working material

International Nuclear Information System (INIS)

2005-01-01

The objective of the technical meeting was to provide a forum for in-depth scientific and technical exchange on topics related to the decommissioning experience with fast reactors, in particular with regard to the decommissioning of components after sodium draining. Accordingly, the scope of the meeting covers the review and analyses of the experience gained from the decommissioning of both active sodium loops and sodium cooled fast reactors (e.g., KNK II, Superphenix, RAPSODIE, EBR-II, FERMI, BN-350, BR-10). It is expected that the outcome of the meeting will contribute to the Agency initiative to preserve fast reactor data and knowledge. The main focus of the technical meeting was given on the decommissioning of both active loop and reactor components (e.g., the primary vessel of a sodium-cooled reactor) that have been drained of sodium, but that still conserve some residual amounts of sodium (e.g., films covering the entire surface of the component, or particular sodium heels that cannot be drained)

Technical and institutional safety features of nuclear power plants in Brazil. Aspectos tecnicos e institucionais da seguranca dos reatores nucleares no Brasil

Energy Technology Data Exchange (ETDEWEB)

Rosa, L P [Sociedade Brasileira de Fisica, Rio de Janeiro, RJ (Brazil)

1986-01-01

This work reports technical, political and institutional safety features of nuclear power plants in Brazil. It is mainly concerned with reactor accidents and personnel safety. The three mile Island and Chernobyl accidents are also discussed and taken as examples. (A.C.A.S.).

Technical note: Development of a Linear Flow Channel Reactor for ...

African Journals Online (AJOL)

Technical note: Development of a Linear Flow Channel Reactor for sulphur removal ... AFRICAN JOURNALS ONLINE (AJOL) · Journals · Advanced Search ... 000 mg∙ℓ-1 Na2SO4 solution) and the Liner Flow Channel Reactors (surface area ...

Analysis of Technical Feasibility of Traveling Wave Reactor

International Nuclear Information System (INIS)

Kim, Sang Ji; Yoo, Jae Woon; Bae, In Ho

2011-01-01

The status and trend of TWR, patent status and its major technical characteristics were examined in this study. Main technical features of traveling wave reactor can be characterized as a reactor operation without refueling up to the reactor life more than 60 years and TWR utilizes depleted uranium which would be produced from the enrichment process as a byproduct. Enriched fuel is only loaded to an igniter which is required for initiation of burning wave. In this study, quantitative analysis of TWR arising from the technical features was carried out in terms of resource utilization, safety and integrity, and proliferation resistance. In parallel with the concept review of TerraPower SWR design concepts, independent analysis of SWR design by altering a design specification and operation strategy was done in this study. The fuel rod design of SWR was also investigated based on the current database of fuel irradiation and performance. The technical issues of TWR or SWR which should be prior to detailed research and development can be summarized as follows: ·Strong physical protection is required during the shuffling or in-service inspection period to improve the proliferation resistance. ·New flow control logic or device is required for distributing the assembly-wise flow to be corresponded with power swing of fuel assembly. ·High integrity cladding material need to be developed for covering the high fast neutron fluence more than three times of current limit which result from the high burnup and long fuel cycle. The metal fuel under the high burnup condition should be validated through the irradiation test

Technical specifications: Tower Shielding Reactor II

International Nuclear Information System (INIS)

1979-02-01

The technical specifications define the key limitations that must be observed for safe operation of the Tower Shielding Reactor II (TSR-II) and an envelope of operation within which there is reasonable assurance that these limits cannot be exceeded. The specifications were written to satisfy the requirements of the Department of Energy (DOE) Manual Chapter 0540, September 1, 1972

Technical specifications: Health Physics Research Reactor

International Nuclear Information System (INIS)

1979-02-01

The technical specifications define the key limitations that must be observed for safe operation of the Health Physics Research Reactor (HPRR) and an envelope of operation within which there is assurance that these limits will not be exceeded. The specifications were written to satisfy the requirements of the Department of Energy (DOE) Manual Chapter 0540, September 1, 1972

Workshop on nuclear technology: A joint effort between ANS and the University of Massachusetts-Lowell

International Nuclear Information System (INIS)

Brown, G.J.; McDevitt, M.A.; Schmidt, D.

1992-01-01

The University of Massachusetts Lowell (UML) (formerly University of Lowell) sponsored, along with the American Nuclear Society (ANS), a 5-day workshop entitled 'Understanding and Teaching about Nuclear Technology and Its Place in Our Society.' More than 30 middle and high school teachers from the New England area (Connecticut, New Hampshire, and Massachusetts) attended the workshop, which was held June 24 through 28, 1991. Based on this experience, and with the expectation of replicating if not improving upon initial success, plans are now under way to offer a similar workshop at UML from June 29 through July 3, 1992

Safety of RBMK reactors: Setting the technical framework

International Nuclear Information System (INIS)

Lederman, L.

1996-01-01

This article reviews major efforts for improving the safety of RBMK reactors through a co-operative IAEA programme initiated in 1992. Specifically covered are technical findings of safety reviews related to the design and operation of the plants, and the documentation of findings through an Agency database intended to facilitate the technical co-ordination of ongoing national and international efforts for improving RBMK safety

Small reactors with simplified design. Proceedings of a technical committee meeting

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-11-01

There is a potential future need for small reactors for applications such as district heating, electricity production at remote locations and desalination. Nuclear energy can provide an environmentally benign alternative to meet these needs. For successful deployment, small reactors must satisfy the requirements of users, regulators and the general public. The IAEA has been following the developments in the field of small reactors as a part of the sub-programme on advanced reactor technology. In accordance with the interests of Member States, a Technical Committee meeting (TCM) was organized in Mississauga, Ontario, Canada, 15-19 May 1995 to discuss the status of designs and design requirements related to small reactors for diverse applications. The papers presented at the TCM and a summary of the discussions are contained in this TECDOC which, it is hoped, will serve the Member States as a useful source of technical information on the development of small reactors with simplified design. Refs, figs, tabs.

Small reactors with simplified design. Proceedings of a technical committee meeting

International Nuclear Information System (INIS)

1997-11-01

There is a potential future need for small reactors for applications such as district heating, electricity production at remote locations and desalination. Nuclear energy can provide an environmentally benign alternative to meet these needs. For successful deployment, small reactors must satisfy the requirements of users, regulators and the general public. The IAEA has been following the developments in the field of small reactors as a part of the sub-programme on advanced reactor technology. In accordance with the interests of Member States, a Technical Committee meeting (TCM) was organized in Mississauga, Ontario, Canada, 15-19 May 1995 to discuss the status of designs and design requirements related to small reactors for diverse applications. The papers presented at the TCM and a summary of the discussions are contained in this TECDOC which, it is hoped, will serve the Member States as a useful source of technical information on the development of small reactors with simplified design

Technical aspects of high converter reactors

International Nuclear Information System (INIS)

1992-02-01

The meeting provided an opportunity to review and discuss national R and D programs, various technical aspects of and worldwide progress in the implementation of high conversion reactors. The meeting was attended by 66 participants from 18 countries and 2 international organizations. 33 papers were presented. A separate abstract was prepared for each of these papers. Refs, figs, tabs, slides and diagram

Development of technical specifications for research reactors

International Nuclear Information System (INIS)

Anon.

1983-01-01

This standard identifies and establishes the content of technical specifications for research reactors. Areas addressed are: definitions, safety limits, limiting safety system settings, limiting conditions for operation, surveillance requirements, design features and administrative controls. Sufficient detail is incorporated so that applicable specifications can be derived or extracted

Students from UMass/Lowell Win $15,000 EPA Grant for Innovative Technology Project

Science.gov (United States)

A student research team from the University of Massachusetts in Lowell has been awarded $15,000 from the US Environmental Protection Agency to research a technology that would turn seafood shells and waste into fertilizer.

High temperature gas cooled reactor technology development. Proceedings of a technical committee meeting

International Nuclear Information System (INIS)

1997-12-01

The successful introduction of an advanced nuclear power plant programme depends on many key elements. It must be economically competitive with alternative sources of energy, its technical development must assure operational dependability, the support of society requires that it be safe and environmentally acceptable, and it must meet the regulatory standards developed for its use and application. These factors interrelate with each other, and the ability to satisfy the established goals and criteria of all of these requirements is mandatory if a country or a specific industry is to proceed with a new, advanced nuclear power system. It was with the focus on commercializing the high temperature gas cooled reactor (HTGR) that the IAEA's International Working Group on Gas Cooled Reactors recommended this Technical Committee Meeting (TCM) on HTGR Technology Development. Over the past few years, many Member States have instituted a re-examination of their nuclear power policies and programmes. It has become evident that the only realistic way to introduce an advanced nuclear power programme in today's world is through international co-operation between countries. The sharing of expertise and technical facilities for the common development of the HTGR is the goal of the Member States comprising the IAEA's International Working Group on Gas Cooled Reactors. This meeting brought together key representatives and experts on the HTGR from the national organizations and industries of ten countries and the European Commission. The state electric utility of South Africa, Eskom, hosted this TCM in Johannesburg, from 13 to 15 November 1996. This TCM provided the opportunity to review the status of HTGR design and development activities, and especially to identify international co-operation which could be utilized to bring about the commercialization of the HTGR

New training reactor at Dresden Technical University

International Nuclear Information System (INIS)

Hansen, W.; Knorr, J.; Wolf, T.

2006-01-01

A total of 14 low-power (up to 10 W) training reactors have been operated at German universities, 9 of them officially classified as being operational in 2004, though for very different uses. This number is expected to drop sharply. The only comprehensive upgrading of a training reactor took place at Dresden Technical University: AKR-2, the most modern facility in Germany, started routine operation in April 2005, under a newly granted license pursuant to Sec. 7, Subsec. 1 of the German Atomic Energy Act, for training students in nuclear technology, for suitable research projects, and a a center of information about reactor technology and nuclear technology for the interested public. One special aspect of this refurbishment was the installation of digital safety I and C systems of the TELEPERM XS line, which are used also in other modern plants. This fact, plus the easy possibility to use the plant for many basic experiments in reactor physics and radiation protection, make the AKR-2 attractive also to other users (e.g. for training reactor personnel or other persons working in nuclear technology). (orig.)

Standard Technical Specifications for Combustion Engineering Pressurized Water Reactors

International Nuclear Information System (INIS)

Vito, D.J.

1980-12-01

The Standard Technical Specifications for Combustion Engineering Pressurized Water Reactors (CE-STS) is a generic document prepared by the US NRC for use in the licensing process of current Combustion Engineering Pressurized Water Reactors. The CE-STS sets forth the limits, operating conditions, and other requirements applicable to nuclear reactor facility operation as set forth by Section 50.36 of 10 CFR 50 for the protection of the health and safety of the public. The document is revised periodically to reflect current licensing requirements

Zero energy reactor RB technical characteristics and experimental possibilities

Energy Technology Data Exchange (ETDEWEB)

Jovanovic, S; Takac, S; Raisic, N; Lolic, B; Markovic, H [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)

1963-04-15

The zero energy reactor RB was constructed in 1958 in accordance with the nuclear reactor development programme of the Boris Kidric Institute of Nuclear Sciences. The reactor was in operation until the middle of 1959 when the heavy water, serving as the moderator, was transported to the high flux reactor RA, built at the same time at the Boris Kidric Institute. Owing to the fact that the purchase of new quantities of heavy water was planned for 1961 it was decided to reconstruct the RB reactor in order to improve the safety of the system and to obtain better flexibility in performing the experiments. New control, safety and radiation monitoring systems were constructed. Some changes were also made on the reactor tank, water circulation system and the water level monitoring equipment. The reconstruction was completed in 1961. and the heavy water was delivered early in 1962. The reconstructed reactor was critical for the first time in summer 1962, and from that time was in continuous operation. This report presents an outline of the design and construction characteristics of the reactor. The main intention is to inform potential users of the reactor about experimental possibilities, advantages and disadvantages of such a critical facility.

Zero energy reactor RB technical characteristics and experimental possibilities

International Nuclear Information System (INIS)

Jovanovic, S.; Takac, S.; Raisic, N.; Lolic, B.; Markovic, H.

1963-04-01

The zero energy reactor RB was constructed in 1958 in accordance with the nuclear reactor development programme of the Boris Kidric Institute of Nuclear Sciences. The reactor was in operation until the middle of 1959 when the heavy water, serving as the moderator, was transported to the high flux reactor RA, built at the same time at the Boris Kidric Institute. Owing to the fact that the purchase of new quantities of heavy water was planned for 1961 it was decided to reconstruct the RB reactor in order to improve the safety of the system and to obtain better flexibility in performing the experiments. New control, safety and radiation monitoring systems were constructed. Some changes were also made on the reactor tank, water circulation system and the water level monitoring equipment. The reconstruction was completed in 1961. and the heavy water was delivered early in 1962. The reconstructed reactor was critical for the first time in summer 1962, and from that time was in continuous operation. This report presents an outline of the design and construction characteristics of the reactor. The main intention is to inform potential users of the reactor about experimental possibilities, advantages and disadvantages of such a critical facility

Technical meeting to 'Review of national programmes on fast reactors and accelerator driven systems (ADS)'. Working material

International Nuclear Information System (INIS)

2003-01-01

36th Annual Meeting of the Technical Working Group on Fast Reactors, the IAEA Technical Meeting (TM) on 'Review of National Programmes on Fast Reactors and Accelerator Driven Systems (ADS)', hosted by the Korean Atomic Energy Research Institute (KAERI) was attended by TWG-FR Members and Advisers from the following Member States (MS) and International Organizations: Brazil, France, Germany, India, Japan, the Republic of Kazakhstan, the Republic of Korea, the Russian Federation, the United Kingdom, the United States of America, and the OECD/NEA. The objectives of the meeting were to: 1) exchange information on the national programmes on Fast Reactors (FR) and Accelerator Driven Systems (ADS); 2) review the progress since the 35th TWG-FR Annual Meeting, including the status of the actions; 3) consider meeting arrangements for 2003 and 2004; 4) review the Agency's co-ordinated research activities in the field of FRs and ADS, as well as co-ordination of the TWG-FR's activities with other organizations. The participants made presentations on the status of the respective national programmes on FR and ADS development. A summary of the highlights for the period since the 35th TWG-FR Annual Meeting

Annual report 1983 Interfacultary Reactor Institute

International Nuclear Information System (INIS)

1984-01-01

In this annual report of the Dutch Interfacultary Reactor Institute, summary- and detailed reports are presented of current research during 1983 of the departments radiochemistry, radiation chemistry, radiation physics, reacotor physics, and radiation hygiene and dosimetry. (H.W.). refs.; figs.; tabs

Disposal of mixed waste: Technical, institutional, and policy factors

International Nuclear Information System (INIS)

Waters, R.D.; Gruebel, M.M.; Letourneau, M.J.; Case, J.T.

1996-01-01

In conjunction with the affected States as part of their interactions required by the Federal Facilities Compliance Act, the Department of Energy has been developing a process for a disposal configuration for its mixed low-level waste (MLLW). This effort, spanning more than two years, has reduced the potential disposal sites from 49 to 15. The remaining 15 sites have been subjected to a performance evaluation to determine their strengths and weaknesses for disposal of MLLW. The process has included institutional and policy factors as well as strictly technical analyses, and technical analyses must be supported by technical analyses, and technical analyses must be performed within a framework which includes some institutional considerations, with the institutional considerations selected for inclusion largely a matter of policy. While the disposal configuration process is yet to be completed, the experience to date offers a viable approach for solving some of these issues. Additionally, several factors remain to be addressed before an MLLW disposal configuration can be developed

Bedrock geologic map of parts of the Eden, Albany, Lowell, and Irasburg quadrangles, Vermont

Data.gov (United States)

Vermont Center for Geographic Information — Digital Data from VG09-4 (Digitized draft of VG97-5): Kim, J., 2009, Bedrock geologic map of parts of the Eden, Albany, Lowell, and Irasburg quadrangles, VGS...

Technical Requirements For Reactors To Be Deployed Internationally For the Global Nuclear Energy Partnership

International Nuclear Information System (INIS)

Ingersoll, Daniel T.

2007-01-01

The Global Nuclear Energy Partnership (GNEP) seeks to create an international regime to support large-scale growth in the worldwide use of nuclear energy. Fully meeting the GNEP vision may require the deployment of thousands of reactors in scores of countries, many of which do not use nuclear energy currently. Some of these needs will be met by large-scale Generation III and III+ reactors (>1000 MWe) and Generation IV reactors when they are available. However, because many developing countries have small and immature electricity grids, the currently available Generation III(+) reactors may be unsuitable since they are too large, too expensive, and too complex. Therefore, GNEP envisions new types of reactors that must be developed for international deployment that are 'right sized' for the developing countries and that are based on technologies, designs, and policies focused on reducing proliferation risk. The first step in developing such systems is the generation of technical requirements that will ensure that the systems meet both the GNEP policy goals and the power needs of the recipient countries. Reactor systems deployed internationally within the GNEP context must meet a number of requirements similar to the safety, reliability, economics, and proliferation goals established for the DOE Generation IV program. Because of the emphasis on deployment to nonnuclear developing countries, the requirements will be weighted differently than with Generation IV, especially regarding safety and non-proliferation goals. Also, the reactors should be sized for market conditions in developing countries where energy demand per capita, institutional maturity and industrial infrastructure vary considerably, and must utilize fuel that is compatible with the fuel recycle technologies being developed by GNEP. Arrangements are already underway to establish Working Groups jointly with Japan and Russia to develop requirements for reactor systems. Additional bilateral and multilateral

Overview of technical Issues Associated with the Long Term Storage of Light Water Reactor used Nuclear Fuel

International Nuclear Information System (INIS)

Sorenson, Ken B.

2014-01-01

The nuclear power technical community is developing the technical basis for demonstrating the safety of storing used nuclear fuel for extended periods of time. The combination of reactor operations that off-load spent fuel to interim storage, coupled with delays in repository construction, has resulted in the expectation that storage periods may be for longer periods of time than originally intended. As more fuel continues to be off-loaded from operating reactors, the need for expanded interim storage also increases. As repository programs are delayed, interim storage requirements will likely exceed licensing term limits. To address these operational realities, there has been a concerted international effort to identify and prioritize the technical issues that need to be addressed in order to demonstrate the safety of storing used nuclear fuel for extended periods of time. Since this is an international effort, different storage systems, regulations, and policies need to be considered. This results in differences in technical issues, as well as differences in priorities. However, this effort also identifies important commonalities in some technical areas that need to be addressed. A broad-based international evaluation of these technical issues provides a better understanding of technical concerns as they relate to individual storage systems and specific national regulatory frameworks. While there are several international activities underway that are focused on long term storage, this paper will discuss the activities of the Electric Power Research Institute (EPRI)/Extended Storage Collaboration Program (ESCP) International Subcommittee. A status report detailing the identification and prioritization of the technical issues was presented at the PSAM11 Conference in June 2012 (1). Since that conference, a final report has been completed by the EPRI/ESCP International Subcommittee (2). This paper will provide important results of the final report as well as

Digital data for the Hazens Notch and a portion of the Lowell quadrangles, Vermont

Data.gov (United States)

Vermont Center for Geographic Information — Digital Data from VG03-3B Digital data for the Hazens Notch and a portion of the Lowell quadrangles, Vermont: Vermont Geological Survey Open File Report VG03-3B, The...

Main research results in the field of nuclear power engineering of the Nuclear Reactors and Thermal Physics Institute in 2014

International Nuclear Information System (INIS)

Trufanov, A.A.; Orlov, Yu.I.; Sorokin, A.P.; Chernonog, V.L.

2015-01-01

The main results of scientific and technological activities for last years of the Nuclear Reactors and Thermal Physics Institute FSUE SSC RF - IPPE in solving problems of nuclear power engineering are presented. The work have been carried out on the following problems: justification of research and development solutions and safety of NPPs with fast reactors of new generation with sodium (BN-1200, MBIR) and lead (BREST-OD-300) coolants, justification of safety of operating and advanced NPPs with WWER reactor facilities (WWER-1000, AEhS 2006, WWER-TOI), development and benchmarking of computational codes, research and development support of Beloyarsk-3 (BN-600) and Bilibino (BN-800) NPPs operation, decommissioning of AM and BR-10 research reactors, pilot scientific studies (WWER-SKD, ITER), international scientific and technical cooperation. Problems for further investigations are charted [ru

Technical Meeting on Passive Shutdown Systems for Liquid Metal-Cooled Fast Reactors. Working Material

International Nuclear Information System (INIS)

2015-01-01

A major focus of the design of modern fast reactor systems is on inherent and passive safety. Specific systems to improve reactor safety performance during accidental transients have been developed in nearly all fast reactor programs, and a large number of proposed systems have reached various stages of maturity. This Technical Meeting on Passive Shutdown Systems for Fast Reactors, which was recommended by the Technical Working Group on Fast Reactors (TWG-FR), addressed Member States’ expressed need for information exchange on projects and programs in the field, as well as for the identification of priorities based on the analysis of technology gaps to be covered through R&D activities. This meeting was limited to shutdown systems only, and did not include other passive features such as natural circulation decay heat removal systems etc.; however the meeting catered to passive shutdown safety devices applicable to all types of fast neutron systems. It was agreed to initiate a new study and produce a Nuclear Energy Series (NES) Technical Report to collect information about the existing operational systems as well as innovative concepts under development. This will be a useful source for member states interested in gaining technical expertise to develop passive shutdown systems as well as to highlight the importance and development in this area

10 CFR 50.36a - Technical specifications on effluents from nuclear power reactors.

Science.gov (United States)

2010-01-01

...; Ineligibility of Certain Applicants § 50.36a Technical specifications on effluents from nuclear power reactors..., including expected occurrences, as low as is reasonably achievable, each licensee of a nuclear power reactor... the design, construction, and operation of nuclear power reactors indicates that compliance with the...

GENDER PARTICIPATION IN TECHNICAL TRAINING INSTITUTIONS

African Journals Online (AJOL)

The paper focuses on skill training opportunities for females in Technical Education Programmes (TEP) in Kenya. In Africa, labour markets have become so competitive that females need to be assisted to enter such markets. Expanding skill-training opportunities for females in training institutions could meet this demand.

Summary report of the final technical meeting on 'International Reactor Dosimetry File: IRDF-2002'

International Nuclear Information System (INIS)

Griffin, Patrick J.; Paviotti-Corcuera, R.

2003-10-01

Presentations, recommendations and conclusions of the Final Technical Meeting on 'International Reactor Dosimetry File: IRDF-2002' are summarized in this report. The main aims of this meeting were to discuss scientific and technical matters related to reactor dosimetry and to assign responsibilities for the preparation of the final version of the IRDF- 2002 library and the associated TECDOC. Tasks were assigned and deadlines were agreed. Participants emphasized that accurate and complete nuclear data for reactor dosimetry are essential to improve the assessment accuracies for reactor pressure vessel service lifetimes in nuclear power plants, as well as for other neutron metrology applications such as boron neutron capture therapy, therapeutic use of medical isotopes, nuclear physics measurements, and reactor safety applications. (author)

Technical/institutional prerequisite for nuclear forensics response framework

International Nuclear Information System (INIS)

Tamai, Hiroshi; Okubo, Ayako; Kimura, Yoshiki; Kokaji, Lisa; Shinohara, Nobuo; Tomikawa, Hirofumi

2016-01-01

Nuclear Forensics capability has been developed under the international collaborations. For its effective function, technical development in analysis of seized nuclear materials as well as the institutional development in comprehensive response framework are required under individual national responsibility. In order to keep the “chain of custody” in the proper operation of sample collection at the event scene, radiological analysis at the laboratory, storage of the samples, and further inspection and trial, close cooperation and information sharing between relevant organisations are essential. IAEA issues the Implementing Guide to provide the model action plan and assists individual national development. Some countries at the advancing stage of national response framework, promote the international cooperation for the technical improvement and awareness cultivation. Examples in such national developments will be introduced and prospective technical/institutional prerequisite for nuclear forensics response framework will be studied. (author)

Pareto analysis of critical factors affecting technical institution evaluation

Directory of Open Access Journals (Sweden)

Victor Gambhir

2012-08-01

Full Text Available With the change of education policy in 1991, more and more technical institutions are being set up in India. Some of these institutions provide quality education, but others are merely concentrating on quantity. These stakeholders are in a state of confusion about decision to select the best institute for their higher educational studies. Although various agencies including print media provide ranking of these institutions every year, but their results are controversial and biased. In this paper, the authors have made an endeavor to find the critical factors for technical institution evaluation from literature survey. A Pareto analysis has also been performed to find the intensity of these critical factors in evaluation. This will not only help the stake holders in taking right decisions but will also help the management of institutions in benchmarking for identifying the most important critical areas to improve the existing system. This will in turn help Indian economy.

Technical findings related to Generic Issue 23: Reactor coolant pump seal failure

International Nuclear Information System (INIS)

Ruger, C.J.; Luckas, W.J. Jr.

1989-03-01

Reactor coolant pumps contain mechanical seals to limit the leakage of pressurized coolant from the reactor coolant system to the containment. These seals have the potential to leak, and a few have degraded and even failed resulting in a small break loss of coolant accident (LOCA). As a result, ''Reactor Coolant Pump Seal Failure,'' Generic Issue 23 was established. This report summarizes the findings of a technical investigation generated as part of the program to resolve this issue. These technical findings address the various fact-finding issue tasks developed for the action plan associated with the generic issue, namely background information on seal failure, evaluation of seal cooling, and mechanical- and maintenance-induced failure mechanisms. 46 refs., 15 figs., 14 tabs

Sharing of Rensselaer Polytechnic Institute Reactor Critical Facility (RCF)

International Nuclear Information System (INIS)

1995-01-01

The RPI Reactor Critical Facility (RCF) operated successfully over the period fall 1994 - fall 1995. During this period, the RCF was used for Critical Reactor Laboratory spring 1995 (12 students); Reactor Operations Training fall 1994 (3 students); Reactor Operations Training spring 1995 (3 students); and Reactor Operations Training fall 1995 (3 students). Thirty-two Instrumentation and Measurement students used the RCF for one class for hands-on experiments with nuclear instruments. In addition, a total of nine credits of PhD thesis work were carried out at the RCF. This document constitutes the 1995 Report of the Rensselaer Polytechnic Institute's Reactor Critical Facility (RCF) to the USNRC, to the USDOE, and to RPI management

Relevance of Technical Training Institutions Taught Skills to Current ...

African Journals Online (AJOL)

Technical education in Kenya is a necessary condition that enables members of society to productively function in technologically rapidly changing society. Technical training institutions have the responsibility to develop skilled Artisans, Craftsmen and Technician for employment in business organizations and industries.

Technical safety appraisal of the N-Reactor

International Nuclear Information System (INIS)

1986-07-01

This report presents the results of a Technical Safety Appraisal conducted at the Hanford N-Reactor. A team of specialists gathered information for about three weeks on all areas related to safety at the plant. Operational practices, maintenance practices, training drills, and hardware condition were observed. Several recommendations are made in order to correct incomplete rule implementation, to correct hazardous practices, and to promote improvement in satisfactory areas

Technical Meeting on Liquid Metal Reactor Concepts: Core Design and Structural Materials. Presentations

International Nuclear Information System (INIS)

2013-01-01

The objective of the Technical Meeting is to present and discuss innovative liquid metal fast reactor (LMFR) core designs with special focus on the choice, development, testing and qualification of advanced reactor core structural materials

The Importance of Institutional Culture at a Technical College

Science.gov (United States)

Adams, Sarah Y.

2014-01-01

This is a qualitative study about the importance of institutional culture at a particular technical college in Wisconsin. It examines the administrators' understanding of their institution's culture, the importance they attach to the culture, and how they perceive they both manage and influence their institution's culture. Bergquist and Pawlak's…

Russian RBMK reactor design information

International Nuclear Information System (INIS)

1993-11-01

This document concerns the systems, design, and operations of the graphite-moderated, boiling, water-cooled, channel-type (RBMK) reactors located in the former Soviet Union (FSU). The Russian Academy of Sciences Nuclear Safety Institute (NSI) in Moscow, Russia, researched specific technical questions that were formulated by the Pacific Northwest Laboratory (PNL) and provided detailed technical answers to those questions. The Russian response was prepared in English by NSI in a question-and-answer format. This report presents the results of that technical exchange in the context they were received from the NSI organization. Pacific Northwest Laboratory is generating this document to support the US Department of Energy (DOE) community in responding to requests from FSU states, which are seeking Western technological and financial assistance to improve the safety systems of the Russian-designed reactors. This report expands upon information that was previously available to the United States through bilateral information exchanges, international nuclear society meetings, International Atomic Energy Agency (IAEA) reactor safety programs, and Research and Development Institute of Power Engineering (RDIPE) reports. The response to the PNL questions have not been edited or reviewed for technical consistency or accuracy by PNL staff or other US organizations, but are provided for use by the DOE community in the form they were received

Standard Technical Specifications for General Electric Boiling Water Reactors (BWR/5)

International Nuclear Information System (INIS)

Bottimore, R.R.

1980-12-01

The Standard Technical Specifications for General Electric Boiling Water Reactors (GE-STS) is a generic document prepared by the US NRC for use in the licensing process of current General Electric Boiling Water Reactors. The GE-STS sets forth the limits, operating conditions, and other requirements applicable to nuclear reactor facility operation as set forth by Section 50.36 of 10 CFR Part 50 for the protection of the health and safety of the public. The document is revised periodically to reflect current licensing requirements

Standard technical specifications for Babcock and Wilcox pressurized water reactors. Revision 4. Technical report

International Nuclear Information System (INIS)

Virgilio, M.J.

1980-10-01

The Standard Technical Specifications for Babcock and Wilcox Pressurized Water Reactors (BandW-STS) is a generic document prepared by the U.S. NRC for use in the licensing process. The BandW-STS provide applicants with model specifications to be used in formulation plant-specific technical specifications required by 10 CFR Part 50, Section 50.36, which set forth the specific characteristics of the facility and the conditions for its operation that are required to provide adequate protection to the health and safety of the public. This document is revised periodically to reflect current licensing requirements

Advanced Fuel Pellet Materials and Fuel Rod Design for Water Cooled Reactors. Proceedings of a Technical Committee Meeting

International Nuclear Information System (INIS)

2010-10-01

The economics of current nuclear power plants have improved through increased fuel burnup and longer fuel cycles, i.e. increasing the effective time that fuel remains in the reactor core and the amount of energy it generates. Efficient consumption of fissile material in the fuel element before it is discharged from the reactor means that less fuel is required over the reactor's life cycle, which results in lower amounts of fresh fuel, lower spent fuel storage costs, and less waste for ultimate disposal. Better utilization of fissile nuclear materials, as well as more flexible power manoeuvring, place challenging operational demands on materials used in reactor components, and first of all, on fuel and cladding materials. It entails increased attention to measures ensuring desired in-pile fuel performance parameters that require adequate improvements in fuel material properties and fuel rod designs. These are the main reasons that motivated the IAEA Technical Working Group on Fuel Performance and Technology (TWG-FPT) to recommend the organization of a Technical Committee Meeting on Advanced Fuel Pellet Materials and Fuel Rod Designs for Power Reactors. The proposal was supported by the IAEA TWGs on Advanced Technologies for Light and Heavy Water-Cooled Reactors (TWG-LWR and TWG-HWR), and the meeting was held at the invitation of the Government of Switzerland at the Paul Scherrer Institute in Villigen, from 23 to 26 November 2009. This was the third IAEA meeting on these subjects (the first was held in 1996 in Tokyo, Japan, and the second in 2003 in Brussels, Belgium), which reflects the continuous interest in the above issues among Member States. The purpose of the meeting was to review the current status in the development of fuel pellet materials and to explore recent improvements in fuel rod designs for light and heavy water cooled power reactors. The meeting was attended by 45 specialists representing fuel vendors, nuclear utilities, research and development

Technical-scientific production of IPEN - 1992

International Nuclear Information System (INIS)

1993-01-01

The technical-scientific production from Nuclear and Energetic Research Institute (IPEN), Brazil, is presented, including 471 references, where 421 papers are from the 1992 year and 50 from the earlier years. The main areas that the IPEN acts are: technical-scientific development of materials; nuclear processes; research reactors; nuclear utilization; nuclear safety and human resources formation. (C.G.C.)

Annual report 90. Institute for advanced materials

International Nuclear Information System (INIS)

1991-01-01

The Annual Report 1990 of the Institute for Advanced Materials of the JRC highlights the Scientific Technical Achievements and presents in the Annex the Institute's Competence and Facilities available to industry for services and research under contract. The Institute executed in 1990 the R and D programme on advanced materials of the JRC and contributed to the programmes: reactor safety, radio-active waste management, fusion technology and safety, nuclear fuel and actinide research. The supplementary programme: Operation of the High Flux Reactor is presented in condensed form. A full report is published separately

Summary report of the technical meeting on 'International Reactor Dosimetry File: IRDF-2002'

International Nuclear Information System (INIS)

Greenwood, L.R.; Paviotti-Corcuera, R.

2002-09-01

This report summarizes the presentations, recommendations and conclusions of the Technical Meeting on 'International Reactor Dosimetry File: IRDF-2002.' The purpose of this meeting was to discuss scientific and technical matters related to the subject and coordinate related tasks. Discussions were held and recommendations were given for the preparation of the files on topics related to: reactions to be included, need for new evaluations or revisions, decay data, radiation damage data, integral testing in benchmark fields, and computer codes to be included. Tasks were assigned and deadlines were set. The participants emphasized that accurate and complete knowledge of nuclear data for reactor dosimetry are essential for improving the accuracy of the reactor pressure vessel service life assessment of nuclear power plants as well as in other neutron metrology applications such as boron neutron capture therapy, therapeutic use of medical isotopes, nuclear physics measurements, and reactor safety applications. (author)

AKR-1 nuclear training reactor of Dresden Technical University turns twenty-five

International Nuclear Information System (INIS)

Hansen, W.

2003-01-01

Twenty-five years ago, in the night of July 27 to 28, 1978, the AKR-1 nuclear training reactor of the Dresden Technical University went critical for the first time and was commissioned. On the occasion of this anniversary, a colloquy was arranged with representatives from science, politics and industry, at which the reactor's history, the excellent achievements in research and training with the reactor, and the status and perspectives of this research facility were described. The AKR-1 had been built within the framework of the Nuclear Development Program of the then German Democratic Republic (GDR). The Nuclear Power Scientific Division of the Dresden Technical University had been entrusted with the responsibility, among other things, to train university personnel for the GDR Nuclear Power Program. The review by an expert group in 1996 of this plant had resulted in a recommendation in favor of long-term plant operation. A nuclear licensing procedure to this effect was initiated, and the necessary technical backfitting measures were implemented. The AKR-1 plant now equally serves for the specialized training of students and for research. (orig.) [de

Development of Regulatory Technical Requirements for the Advanced Integral Type Research Reactor

International Nuclear Information System (INIS)

Jo, Jong Chull; Yune, Young Gill; Kim, Woong Sik; Kim, Hho Jung

2004-01-01

This paper presents the current status of the study on the development of regulatory technical requirements for the licensing review of an advanced integral type research reactor of which the license application is expected in a few years. According to the Atomic Energy Act of Korea, both research and education reactors are subject to the technical requirements for power reactors in the licensing review. But, some of the requirements may not be applicable or insufficient for the licensing reviews of reactors with unique design features. Thus it is necessary to identify which review topics or areas can not be addressed by the existing requirements and to develop the required ones newly or supplement appropriately. Through the study performed so far, it has been identified that the following requirements need to be developed newly for the licensing review of SMART-P: the use of proven technology, the interfacial facility, the non-safety systems, and the metallic fuels. The approach and basis for the development of each of the requirements are discussed. (authors)

Fuel cycle options for light water reactors and heavy water reactors. Proceedings of a technical committee meeting

International Nuclear Information System (INIS)

1999-11-01

In the second half of the 20th century nuclear power has evolved from the research and development environment to an industry that supplies 16% of the world's electricity. By the end of 1997, over 8500 reactor-years of operating experience had been accumulated. Global environmental change, and the continuing increase in global energy supply required to provide increasing populations with an improving standard of living, make the contribution from nuclear energy even more important for the next century. For nuclear power to achieve its full potential and make its needed contribution, it must be safe, economical, reliable and sustainable. All of these factors can be enhanced by judicious choice and development of advanced fuel cycle options. The Technical Committee Meeting (TCM) on Fuel Cycle Options for Light Water Reactors and Heavy Water Reactors was hosted by Atomic Energy of Canada Limited (AECL) on behalf of the Canadian Government and was jointly conducted within the frame of activities of the IAEA International Working Group on Advanced Technologies for Light Water Reactors (IWG-LWR) and the IAEA International Working Group on Advanced Technologies for Heavy Water Reactors (IWG-HWR). The TCM provided the opportunity to have in-depth discussions on important technical topics which were highlighted in the International Symposium on Nuclear Fuel Cycle and Reactor Strategies: Adjusting to New Realities, held in Vienna, 3-6 June 1997. The main results and conclusions of the TCM were presented as input for discussion at the first meeting of the IAEA newly formed International Working Group on Fuel Cycle Options

Technical Meeting on Fast Reactors and Related Fuel Cycle Facilities with Improved Economic Characteristics. Presentations

International Nuclear Information System (INIS)

2013-01-01

The objectives of the meeting were: • To identify the main issues and technical features that affect capital and energy production costs of fast reactors and related fuel cycle facilities; • To present fast reactor concepts and designs with enhanced economic characteristics, as well as innovative technical solutions (components, subsystems, etc.) that have the potential to reduce the capital costs of fast reactors and related fuel cycle facilities; • To present energy models and advanced tools for the cost assessment of innovative fast reactors and associated nuclear fuel cycles; • To discuss the results of studies and ongoing R&D activities that address cost reduction and the future economic competitiveness of fast reactors; • To identify research and technology development needs in the field, also in view of new IAEA initiatives to help and support Member States in improving the economic competitiveness of fast reactors and associated nuclear fuel cycles

Practice in development and utilization of program-technical complex (PTK) in in-reactor control systems

International Nuclear Information System (INIS)

Gribov, A.A.; Kuzil, A.C.; Padun, S.P.; Surnachov, S.I.; Jakovlev, G.V.

2001-01-01

Experience with the development and utilization of the program-technical complex PTK 'KRUIZ' is analyzed in the paper. A peculiarity of PTK is the orientation on acquisition, processing and diagnostics of signals from in-reactor sensors (thermocouples and SPD). The PTK 'KRUIZ' represents a new generation of tools open for further development, oriented specifically on the use in in-reactor control systems in modernized and built power units of the WWER type. In the PTK 'KRUIZ', methods, models and algorithms proved in nuclear power plants are used accounting for the utilization of up to date technical tools and systematic technical solutions. Experience with the use of basic elements of the PTK 'KRUIZ' at existing WWER reactors including peculiarities of temperature control in nuclear power plants are also dealt within the paper. (Authors)

Decommissioning and decontrolling the R1-reactor

International Nuclear Information System (INIS)

Bergman, C.; Holmberg, B.T.

1985-01-01

Sweden's first nuclear reactor - the research reactor R1 - situated in bedrock under the Royal Technical Institute of Stockholm, has in the period 1981-1983 been subject to a complete decommissioning. The National Institute for Radiation Protection has followed the work in detail, and has after the completion of the decommissioning performed measurements of radioactivity on site. The report gives an account of the work the Institute has done in preparation for- and during decommissioning and specifically report on the measurements for classification of the local as free for non-nuclear use. (aa)

Science Hall of Atomic Energy in Research Reactor Institute, Kyoto University

International Nuclear Information System (INIS)

Hayashi, Takeo

1979-01-01

The Science Hall of Atomic Energy was built as a subsidiary facility of the Research Reactor Institute, Kyoto University. The purpose of this facility is to accept outside demands concerning the application of the research reactor. The building is a two story building, and has the floor area of 901.47 m 2 . There are an exhibition room, a library, and a big lecture room. In the exhibition room, models of the Kyoto University Research Reactor and the Kyoto University Critical Assembly are placed. Various pictures concerning the application of the reactor are on the wall. In the library, people from outside of the Institute can use various books on science. Books for boys and girls are also stocked and used for public use. At the lecture room, various kinds of meeting can be held. (Kato, T.)

Summary report of activities under visiting research program in Research Reactor Institute, Kyoto University, second half of 1989

International Nuclear Information System (INIS)

1990-07-01

The Technical Report is published on occasion by summarizing in the form of prompt report the data required at the time of research and experiment, such as the results of the functional test on various experimental facilities, the test results for the articles made for trial, the state of radiation control and waste treatment, the reports of study meetings and so on, or the remarkable results and new methods obtained in research, the discussion on other papers and reports and others in the Research Reactor Institute, Kyoto University. In this report, the gists of 69 studies carried out by using the Research Reactor and 15 studies by using the Kyoto University Critical Assembly are collected. Adoption number, classification, title, the names of reporters and gist are given for each report. (K.I.)

DOE University Reactor Sharing Program. Final technical report for 1996--1997

International Nuclear Information System (INIS)

Chappas, W.J.; Adams, V.G.

1998-01-01

The Department of Energy University Reactor Sharing Program at University of Maryland, College Park (UMCP) has, once again, stimulated a broad use of the reactor and radiation facilities by undergraduate and graduate students, visitors, and professionals. Participants are exposed to topics such as nuclear engineering, radiation safety, and nuclear reactor operations. This information is presented through various means including tours, slide presentations, experiments, and discussions. Student research using the MUTR is also encouraged. In addition, the Reactor Sharing Program here at the University of Maryland does not limit itself to the confines of the TRIGA reactor facility. Incorporated in the program are the Maryland University Radiation Effects Laboratory, and the UMCP 2 x 4 Thermal Hydraulic Loop. These facilities enhance and give an added dimension to the tours and experiments. The Maryland University Training Reactor (MUTR) and the associated laboratories are made available to any interested institution six days a week on a scheduled basis. Most institutions are scheduled at the time of their first request--a reflection of their commitment to the Reactor Sharing Program. The success of the past years by no means guarantees future success. Therefore, the reactor staff is more aggressively pursuing its outreach program, especially with junior colleges and universities without reactor or radiation facilities; more aggressively developing demonstration and training programs for students interested in careers in nuclear power and radiation technology; and more aggressively up-grading the reactor facilities--not only to provide a better training facility but to prepare for relicensing in the year 2000

Report of research by common utilization in Research Reactor Institute, Kyoto University, in latter half of fiscal 1980

International Nuclear Information System (INIS)

1982-01-01

In the technical report, the data required for research and experiment, such as the result of functional test of various experimental facilities, the test results of the products manufactured for trial, the state of radiation control and waste treatment, and the reports of study meetings, or the remarkable results and new methods obtained in research and the discussion on other papers and reports in the Research Reactor Institute, Kyoto University, are summarized as prompt report. The subject, reporters and synopsis of 69 papers are reported in this publication. (Kako, I.)

Report of research by common utilization in Research Reactor Institute, Kyoto University, in first half of fiscal 1980

International Nuclear Information System (INIS)

1982-01-01

In the technical report, the data required for research and experiment, such as the result of functional test of various experimental facilities, the test results of the products manufactured for trial, the state of radiation control and waste treatment, and the reports of study meetings, or the remarkable results and new methods obtained in research, and the discussion on other papers and reports in the Research Reactor Institute, Kyoto University, are summarized as prompt report. The subject, reporters and synopsis of 54 papers are reported in this publication. (Kako, I.)

Technical evolution and operation of French CO2 cooled reactors (UNGG)

International Nuclear Information System (INIS)

Berthion, Y.

1986-10-01

The technical evolution of the five French CO 2 cooled reactors (UNGG) from 1981 to 1986 needs to be outlined. These technical evolutions concerned the fuel element of Bugey 1 which is now slightly enriched, as well as the load reduction operation required by the grid. In addition work in underway to increase the safety at the two St Laurent units, or to repair the hot steel upper-structures of Chinon-3 unit

Extension of the technical scope of the Paris and Vienna Conventions: fusion reactors and reactors in means of transport

International Nuclear Information System (INIS)

Reye, S.

1993-01-01

This paper examines the possibility of extending the technical scope of the Vienna and Paris Conventions to two types of nuclear installation presently excluded. Industrial use of fusion reactors is not expected for several decades, but the present revision of the liability regime provides a useful opportunity to ensure in advance that future industrial reactors will be covered, as well as covering risks arising from existing research reactors. Inclusion of nuclear reactors comprised in means of transport (in practice, in ships) in the liability regime would have certain advantages, but given their almost exclusively military use, such a proposal would be politically controversial. 18 refs

Results of research and development activities in 1989 of the Institute for Neutron Physics and Reactor Technology

International Nuclear Information System (INIS)

1990-03-01

The Institute for Neutron Physics and Reactor Technology treats research problems of nuclear engineering, mainly those that are related to the development of sodium-cooled fast breeder reactors and fusion reactor technology. The activities are in approximately equal parts of an experimental and theoretical nature. A great part of the research activities is performed in co-operation with other institutes and industrial groups in the framework of projects. For the Fast Breeder Reactor Project the Institute works on reactor physical design and safety problems by the core of large-scale fast breeder reactors. Questions concerning the consequences of accidents in light water reactors upon the environment and the population are treated as part of the Nuclear Safety Project. The Institute contributes to the Reprocessing Project with theoretical investigations on the physics of the fuel cycle and by developing control devices for a reprocessing plant. In the framework of the Fusion Project the Institute is concerned with neutron physical and technological questions of the breeder blanket. (orig.) [de

The Environmental Trilogy project: Balancing technical, institutional, and cultural perspectives to environmental management

International Nuclear Information System (INIS)

Kurstedt, Pamela S.; Jim, Russell; Wadsworth, Bonnie C.W.; Burke, William H.; Kurstedt, Harold A. Jr.

1992-01-01

'The significant problems we face cannot be solved at the same level of thinking we were at when we created them.' Albert Einstein. I've identified an initial set of three perspectives important to building an integrated, comprehensive approach to managing the environment - technical, institutional, and cultural. I've constructed an holistic model (called the Environmental Trilogy) for environmental management, encompassing at least these three perspectives and their interrelationships. In this paper, I outline the model and report the results of a working session facilitated at Virginia Tech in Blacksburg, Virginia, in October 1991, involving three representatives from each of the technical, institutional, and cultural perspectives. The institutional members of this group were people who understand institutional effects, rather than those who represent institutions. The working group discussed and analyzed the technical-institutional, technical-cultural, and institutional-cultural inter- relationships of the environmental trilogy. The goals of the working group were to put structure on the environmental trilogy model, to facilitate the definition of variables, and explore relationships between and among variables. The working group members are to continue studying issues and components, perspectives, connections, and cause-and-effect in the models and report back to interested parties. The outcome is projected to be a more holistic, integrated view of the environment. (author)

Technical Meeting on Fast Reactors and Related Fuel Cycle Facilities with Improved Economic Characteristics. Working Material

International Nuclear Information System (INIS)

2013-01-01

The objectives of the meeting were: - To identify the main issues and technical features that affect capital and energy production costs of fast reactors and related fuel cycle facilities; - To present fast reactor concepts and designs with enhanced economic characteristics, as well as innovative technical solutions (components, subsystems, etc.) that have the potential to reduce the capital costs of fast reactors and related fuel cycle facilities; - To present energy models and advanced tools for the cost assessment of innovative fast reactors and associated nuclear fuel cycles; - To discuss the results of studies and on-going R&D activities that address cost reduction and the future economic competitiveness of fast reactors; and - To identify research and technology development needs in the field, also in view of new IAEA initiatives to help and support Member States in improving the economic competitiveness of fast reactors and associated nuclear fuel cycles

Technical memory 2007

International Nuclear Information System (INIS)

2009-01-01

The technical memory 2007 of the Nuclear Regulatory Authority (ARN) of the Argentine Republic, compiles the papers published in the subject on radiation protection and nuclear safety, and presented in journals, technical reports, congress or meetings of these specialities by personnel of the mentioned institution during 2007. In this edition the documents are presented on: environmental protection; transport of radioactive materials; regulations; research reactors and nuclear power plants; biological radiation effects; therapeutic uses of ionizing radiation and radioprotection of patients; internal dosimetry; physical dosimetry; knowledge management; radioactive waste management [es

ARIES tokamak reactor study

International Nuclear Information System (INIS)

Steiner, D.; Embrechts, M.

1990-07-01

This is a status report on technical progress relative to the tasks identified for the fifth year of Grant No. FG02-85-ER52118. The ARIES tokamak reactor study is a multi-institutional effort to develop several visions of the tokamak as an attractive fusion reactor with enhanced economic, safety, and environmental features. The ARIES study is being coordinated by UCLA and involves a number of institutions, including RPI. The RPI group has been pursuing the following areas of research in the context of the ARIES-I design effort: MHD equilibrium and stability analyses; plasma-edge modeling and blanket materials issues. Progress in these areas is summarized herein

Decommissioning of the research reactors at the Russian Research Centre Kurchatov Institute

International Nuclear Information System (INIS)

Ponomarev-Stepnoy, N.N.; Ryantsev, E.P.; Kolyadin, V.I.; Kucharkin, N.E.; Melkov, E.S.; Gorlinsky, Yu.E.; Kyznetsova, T.I.; Bulkin, B.K.

2002-01-01

The Kurchatov Institute is the largest research center of Russia in the field of nuclear science and engineering. It comprises more than 10 research institutes and scientific-technological complexes carrying out research work in the field of safe development of atomic engineering, controlled thermonuclear fusion, and plasma physics, nuclear physics and elementary particle physics, research reactors, radiation materials technology, solid state physics and superconductivity, molecular and chemical physics, and also perspective know-how's, information science and ecology. This report is basically devoted to the decommissioning of the research reactor installations, in particular to the reactor MR because of the volume and complexity of actions involved. (author)

Advanced Fast Reactor - 100 (AFR-100) Report for the Technical Review Panel

Energy Technology Data Exchange (ETDEWEB)

Grandy, Christopher [Argonne National Lab. (ANL), Argonne, IL (United States); Sienicki, James J. [Argonne National Lab. (ANL), Argonne, IL (United States); Moisseytsev, Anton [Argonne National Lab. (ANL), Argonne, IL (United States); Krajtl, Lubomir [Argonne National Lab. (ANL), Argonne, IL (United States); Farmer, Mitchell T. [Argonne National Lab. (ANL), Argonne, IL (United States); Kim, Taek K. [Argonne National Lab. (ANL), Argonne, IL (United States); Middleton, B. [Argonne National Lab. (ANL), Argonne, IL (United States)

2014-06-04

This report is written to provide an overview of the Advanced Fast Reactor-100 in the requested format for a DOE technical review panel. This report was prepared with information that is responsive to the DOE Request for Information, DE-SOL-0003674 Advanced Reactor Concepts, dated February 27, 2012 from DOE’s Office of Nuclear Energy, Office of Nuclear Reactor Technologies. The document consists of two main sections. The first section is a summary of the AFR-100 design including the innovations that are incorporated into the design. The second section contains a series of tables that respond to the various questions requested of the reactor design team from the subject DOE RFI.

Status of the RA research reactor decommissioning project

International Nuclear Information System (INIS)

Ljubenov, V.; Nikolic, D.; Pesic, M.; Milosevic, M.; Kostic, Lj.; Steljic, M.; Sotic, O.; Antic, D. . E-mail address of corresponding author: vladan@vin.bg.ac.yu; Ljubenov, V.)

2005-01-01

The 6.5 MW heavy water RA research reactor at the VINCA Institute of Nuclear Sciences operated from 1959 to 1984. After 18 years of extended shutdown in 2002 it was decided that the reactor shutdown should be final. Preliminary decommissioning activities have been initiated by the end of 2002 under the Technical Co-operation Programme of the International Atomic Energy Agency. The objective of the project is to implement safe, timely and cost-effective decommissioning of the RA reactor up to unrestricted use of the site. Decommissioning project is closely related to two other projects: Safe Removal of the RA Reactor Spent Nuclear Fuel and Radioactive Waste Management in VINCA Institute. The main phases of the project include preparation of the detailed decommissioning plan, radiological characterization of the reactor site, dismantling and removal of the reactor components and structures, decontamination, final radiological site survey and the documentation of all the activities in order to obtain the approval for unrestricted use of the facility site. In this paper a review of the activities related to the preparation and realization of the RA reactor decommissioning project is given. Status of the project's organizational and technical aspects as for July 2004 are presented and plans for the forthcoming phases of the project realization are outlined. (author)

Technical Meeting on Fast Reactors and Related Fuel Cycle Facilities with Improved Economic Characteristics. Working Material

International Nuclear Information System (INIS)

2013-01-01

In recent years, engineering oriented work, rather than basic research and development (R&D), has led to significant progress in improving the economics of innovative fast reactors and associated fuel cycle facilities, while maintaining and even enhancing the safety features of these systems. Optimization of plant size and layout, more compact designs, reduction of the amount of plant materials and the building volumes, higher operating temperatures to attain higher generating efficiencies, improvement of load factor, extended core lifetimes, high fuel burnup, etc. are good examples of achievements to date that have improved the economics of fast neutron systems. The IAEA, through its Technical Working Group on Fast Reactors (TWG-FR) and Technical Working Group on Nuclear Fuel Cycle Options and Spent Fuel Management (TWG-NFCO), devotes many of its initiatives to encouraging technical cooperation and promoting common research and technology development projects among Member States with fast reactor and advanced fuel cycle development programmes, with the general aim of catalysing and accelerating technology advances in these fields. In particular the theme of fast reactor deployment, scenarios and economics has been largely debated during the recent IAEA International Conference on Fast Reactors and Related Fuel Cycles: Safe Technologies and Sustainable Scenarios, held in Paris in March 2013. Several papers presented at this conference discussed the economics of fast reactors from different national and regional perspectives, including business cases, investment scenarios, funding mechanisms and design options that offer significant capital and energy production cost reductions. This Technical Meeting on Fast Reactors and Related Fuel Cycle Facilities with Improved Economic Characteristics addresses Member States’ expressed need for information exchange in the field, with the aim of identifying the main open issues and launching possible initiatives to help and

Technical meeting on progress in managing, and limiting the consequences of events exceeding the design basis

International Nuclear Information System (INIS)

Fabian, H.

2004-01-01

The Technical Groups on 'Reactor Safety' and 'Thermodynamics and Fluid Dynamics' of the Kerntechnische Gesellschaft e.V. organized a joint technical meeting on 'Progress in Managing, and Limiting the Consequences of, Events Exceeding the Design Basis' at the FTU Training Center of the Karlsruhe Research Center. The topic chosen, the papers presented, the presenters, and the non-technical part of the program met with lively interest on the part of institutions in the nuclear field. These were the objectives of the technical meeting: - Establishing a forum for communicating relevant topics. - In-depth discussion of the main topic, i.e. the advanced development of reactor safety, research in the field, and its application, in twenty selected papers presented by speakers from different institutions. - Presentation of topical work in a nuclear technology institution, the Karlsruhe Research Center. (orig.) [de

Assessment of the Technical Maturity of Generation IV Concepts for Test or Demonstration Reactor Applications, Revision 2

Energy Technology Data Exchange (ETDEWEB)

Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-10-01

The United States Department of Energy (DOE) commissioned a study the suitability of different advanced reactor concepts to support materials irradiations (i.e. a test reactor) or to demonstrate an advanced power plant/fuel cycle concept (demonstration reactor). As part of the study, an assessment of the technical maturity of the individual concepts was undertaken to see which, if any, can support near-term deployment. A Working Group composed of the authors of this document performed the maturity assessment using the Technical Readiness Levels as defined in DOE’s Technology Readiness Guide . One representative design was selected for assessment from of each of the six Generation-IV reactor types: gas-cooled fast reactor (GFR), lead-cooled fast reactor (LFR), molten salt reactor (MSR), supercritical water-cooled reactor (SCWR), sodium-cooled fast reactor (SFR), and very high temperature reactor (VHTR). Background information was obtained from previous detailed evaluations such as the Generation-IV Roadmap but other technical references were also used including consultations with concept proponents and subject matter experts. Outside of Generation IV activity in which the US is a party, non-U.S. experience or data sources were generally not factored into the evaluations as one cannot assume that this data is easily available or of sufficient quality to be used for licensing a US facility. The Working Group established the scope of the assessment (which systems and subsystems needed to be considered), adapted a specific technology readiness scale, and scored each system through discussions designed to achieve internal consistency across concepts. In general, the Working Group sought to determine which of the reactor options have sufficient maturity to serve either the test or demonstration reactor missions.

KfK, Institute for Reactor Components. Results of research and development activities in 1989

International Nuclear Information System (INIS)

1990-03-01

R and D activities at IRB (Institut fuer Reaktorbauelemente - Institute for Reactor Components) are dedicated to thermodynamics and fluid dynamics. Emphasis is on the design of nuclear reactor and fusion reactor components. Environmental engineering was added recently. Most activities are applications-oriented. Fundamental investigations focus on energy research and energy technology. The activities are carried out in the framework of different projects (PKF/nuclear fusion, PSF/nuclear safety, PSU/pollution control). Points of main effort are the development of basic liquid-metal-cooled blanket solutions, investigations on natural convection in reactor tanks, and the cooling properties of future containments for pressurized water reactors in the case of nuclear fusion accidents. (orig./GL) [de

Integral design concepts of advanced water cooled reactors. Proceedings of a technical committee meeting

International Nuclear Information System (INIS)

1997-11-01

Under the sub-programme on non-electrical applications of advanced reactors, the International Atomic Energy Agency has been providing a worldwide forum for exchange of information on integral reactor concepts. Two Technical Committee meetings were held in 1994 and 1995 on the subject where state-of-the-art developments were presented. Efforts are continuing for the development of advanced nuclear reactors of both evolutionary and innovative design, for electricity, co-generation and heat applications. While single purpose reactors for electricity generation may require small and medium sizes under certain conditions, reactors for heat applications and co-generation would be necessary in the small and medium range and need to be located closer to the load centres. The integral design approach to the development of advanced light water reactors has received special attention over the past few years. Several designs are in the detailed design stage, some are under construction, one prototype is in operation. A need has been felt for guidance on a number of issues, ranging from design objectives to the assessment methodology needed to show how integral designs can meet these objectives, and also to identify their advantages and problem areas. The technical document addresses the current status of the design, safety and operational issues of integral reactors and recommends areas for future development

Report of researches by common utilization of facilities in Kyoto University Research Reactor Institute, first half of fiscal year 1981

International Nuclear Information System (INIS)

1983-01-01

The technical report of the Kyoto University Research Reactor Institute is published any time to immediately report on the results of the functional tests of various experimental facilities, the test results for the products made for trial, radiation control, the situation of waste treatment, the data required for research and experiment such as the reports of study meetings, the conspicuous results obtained amid researches, new processes, and the discussion on other papers and reports. In this report, the title, the names of reporters and the summary of 57 researches carried out by the common utilization of the facilities in the Kyoto University Research Reactor Institute are collected. The themes of the researches are such as neutron radiography using a research reactor, measurement of Zr/Hf ratio in zirconium, interstitial germanium atoms in thermal neutron irradiation study, measurement of induced radioactivity due to neutrons in Nagasaki and Hiroshima atomic bombings, properties of semiconductor electrons in radiation study, induction of mutation in crops by neutron irradiation and utilization for breeding, thermal fluorescence mechanism of alkali halide and MgO single crystals, atomic configuration in PZT rhombohedron phase, modulated structure of Cu-Co alloys, excitation of nuclei by positron annihilation and others. (Kako, I.)

Report of researches by common utilization of facilities in Kyoto University Research Reactor Institute, latter half of fiscal year 1981

International Nuclear Information System (INIS)

1983-01-01

The technical report of the Kyoto University Research Reactor Institute is published any time to immediately report on the results of the functional tests of various experimental facilities, the test results for the products made for trial, radiation control, the situation of waste treatment, the data required for research and experiment such as the reports of study meetings, the conspicuous results obtained amid researches, new processes, and the discussion on other papers and reports. In this report, the title, the names of reporters and the summary of 61 researches carried out by the common utilization of the facilities in the Kyoto University Research Reactor Institute are collected. The themes of the researches are such as radioactivation analysis of trace elements in rocks and minerals, anodic oxidation films of GaAs and structure, measurement of yield of uranium isotopes produced by reactor neutron irradiation of thorium, geochemical study of trace elements in hydrosphere by radio-activation analysis, various diseases and variation of elements in rat furs, Moessbauer spectroscopic study of gold compounds with singular coupling by Au-197, measurement of grass-eating quantity and rate of digestion of cows using Au and Eu, sickness biochemical study of trace elements in hair samples of patients and others. (Kako, I.)

Proceedings of the 18th technical meeting on nuclear reactor and radiation for KURRI engineers and the 9th technical official group section 5 meeting in Kyoto University

International Nuclear Information System (INIS)

2010-03-01

This report is a summary of 18th Technical Meeting on Nuclear Reactor and Radiation for KURRI Engineers in Kyoto University. This was also the 9th meeting for technical official group section 5 (nuclear and radiation) in Kyoto University. In the workshop, three special lectures held were: (1) 'On Border Between Subcritical and Supercritical', (2) 'Memories of Nuclear Power Plant Management for 40 Years', and (3) 'Introduction of Technical Office in Faculty of Engineering, Kyoto University'. The technical presentations held were: (1) 'Radiation Background Study of Specialty Products in Senshu Region', (2) 'Introduction of Radioactivation Analysis at KUR', (3) 'Consideration of Critical Approach Method for KUR Low-Enrichment Fuel Reactor Core Using SRAC', (4) 'Evaluation of Temperature Coefficient of KUR Low-Enrichment Fuel Reactor Core Using SRAC'. As training for technical staffs in Technical Office, we visited the facility in Ashiu Research Forest. An introduction of this facility and the comments from the participants were included in this report. (S.K.)

Technical, economic and institutional aspects of regional spent fuel storage facilities

International Nuclear Information System (INIS)

2005-11-01

A particular challenge facing countries with small nuclear programmes is the preparation for extended interim storage and then disposal of their spent nuclear fuel. The costs and complications of providing for away-from-reactor storage facilities and/or geological repositories for relatively small amounts of spent fuel may be prohibitively high, motivating interest in regional solutions. This publication addresses the technical, economic and institutional aspects of regional spent fuel storage facilities (RSFSF) and is based on the results of a series of meetings on this topic with participants from IAEA Member States. Topics discussed include safety criteria and standards, safeguards and physical protection, fuel acceptance criteria, long term stability of systems and stored fuel, selection of site, infrastructure aspects, storage technology, licensing, operations, transport, decommissioning, as well as research and development. Furthermore the publication comprises economic, financial and institutional considerations including organizations and legal aspects followed by political and public acceptance and ethical considerations. Approaches and processes for implementation are discussed, as well as the overall benefits and risks of implementing a regional facility. It is illustrated that implementing a RSFSF facility would involve simultaneously addressing a wide range of diverse challenges. The appendix to this report tabulates the numerous issues that have been touched upon in the study. It appears, however, from the discussions that the challenges can in principle be met; the RSFSF concept is technically feasible and potentially economically viable. The technical committees producing this report did not identify any obvious institutional deficiencies that would prevent completion of such a project. Storing spent fuel in a few safe, reliable, secure facilities could enhance safeguards, physical protection and non-proliferation benefits. The committee also

Safety operation of training reactor VR-1

International Nuclear Information System (INIS)

Matejka, K.

2001-01-01

There are three nuclear research reactors in the Czech Republic in operation now: light water reactor LVR-15, maximum reactor power 10 MW t , owner and operator Nuclear Research Institute Rez; light water zero power reactor LR-0, maximum reactor power 5 kW t , owner and operator Nuclear Research Institute Rez and training reactor VR-1 Sparrow, maximum reactor power 5 kW t , owner and operate Faculty of Nuclear Sciences and Physical Engineering, CTU in Prague. The training reactor VR-1 Vrabec 'Sparrow', operated at the Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University in Prague, was started up on December 3, 1990. Particularly it is designed for training the students of Czech universities, preparing the experts for the Czech nuclear programme, as well as for certain research work, and for information programmes in the nuclear programme, as well as for certain research work, and for information programmes in sphere of using the nuclear energy (public relations). (author)

Research reactor RB, technical characteristics and experimental possibilities; Zbornik radova, Konferencija o koriscenju nuklearnih reaktora u Jugoslaviji

Energy Technology Data Exchange (ETDEWEB)

Sotic, O; Vranic, S [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

1978-05-15

Nuclear research reactor RB tn the Nuclear Engineering Laboratory at the Institute of Nuclear Sciences 'Boris Kidric' in Vinca is the first reactor system built in Yugoslavia in 1958. In this report, the basic technical characteristics of this reactor are described, as well as the experimental possibilities it offers to the users. Its relatively simple construction and flexibility enables direct measurements of a series of physical parameters, and the absence of the biological protection shield makes it very useful for Various biological and other irradiations and dosimetric measurements Where strong neutron source is required. (author) Istrazivacki nuklearni reaktor RB u Laboratoriji za nuklearnu energetiku i tehnicku fiziku Instituta za nuklearne nauke 'Boris Kidric' u Vinci je prvi reaktorski sistem izgradjen u Jugooslaviji 1958. godine. U ovom radu opisane su osnovne tehnicke karakteristike tog reaktora, kao i mogucnosti za izvodjenje eksperimenata koje on pruza korisnicima. Njegova relativno jednostavna konstrukcija i fleksibilnost omogucavaju da se na njemu izvrse direktna merenja niza fizickih parametara, a s druge strane odsustvo bioloskog zastitnog omotaca cini ga veoma pogodnim za razna bioloska i druga ozracivanja, a takodje i dozimetrijska merenja gde se zahteva snazan izvor neutrona. (author)

Performance Evaluation of Indian Technical Institutions Using PROMETHEE-GAIA Approach

Science.gov (United States)

Ranjan, Rajeev; Chakraborty, Shankar

2015-01-01

It has now become an important issue to evaluate the performance of technical institutions to develop better research and enrich the existing teaching processes. The results of such performance appraisal would serve as a reference point for decisions to choose a particular institution, hire manpower, and provide financial support for the…

IAEA Technical Meeting on Innovative Heat Exchanger and Steam Generator Designs for Fast Reactors. Working Material

International Nuclear Information System (INIS)

2011-01-01

The IAEA, within the framework of its Nuclear Energy Department’s Technical Working Group on Fast Reactors (TWG-FR), assists Member States activities in fast reactors technology development areas by providing an umbrella for information exchange [topical Technical Meetings (TMs), Workshops and large Conferences] and collaborative R&D [Coordinated Research Projects (CRPs)]. The Technical meeting on “Innovative Heat Exchanger and Steam Generator Designs for Fast Reactors” was held from 21 – 22 December 2011 in Vienna, addressing Member States’ expressed needs of information exchange in the field of advanced fast reactor design features, with particular attention to innovative heat exchangers and steam generators. The Objective of the TM is to provide a global forum for in-depth information exchange and discussion on the most advanced concepts of heat exchangers and steam generators for fast reactors. More specifically, the objectives are: · Review of the status of advanced fast reactor development activities with special emphasis on design and performance of heat exchangers and steam generators; · Discuss requirements for innovative heat exchangers and steam generators; · Present results of studies and conceptual designs for innovative heat exchangers and steam generators; · Provide recommendations for international collaboration under the IAEA aegis. The meeting agenda of the meeting is in Annex I

Future prospects of reactor Ra at VINCA institute

International Nuclear Information System (INIS)

Davidovic, M.; Babic-Stojic, B.; Dobrijevic, R.

1997-01-01

Reactor RA at Nuclear Research Institute Vinca belongs to a group of the medium thermal neutron flux reactors, according to classification at end of nineties. At the beginning reactor RA has been used as a powerful source of neutrons and gamma-quanta for various experiments (interaction of neutrons and gamma-quanta with materials) and for production of artificial radioactive materials for commercial use. Very successful utilization of this neutron spectrum has been in its use for structural studies of crystal materials and liquid metals, for magnetic structure studies of various magnetic materials, as well as, dynamic properties of ferro magnetics, ferroelectrics, etc. This kind of spectrometers still exist at reactor RA and with an improved detection system could be used again if reactor starts functioning. Besides this, a part of activity was devoted to construction of neutron guide tubes for thermal neutrons and this could also be accomplished relatively easy in the future. A part of the activities of the reactor should in the future be devoted to the training of students in the field of solid state physics and nuclear physics. Particular attention will be paid to the use of established technologies in production of radioactive isotopes and a new class of isotopes for custom use will be developed as well as highly commercial and prospective products (silicon doping, radiography, etc.). (author)

Technical water system of the Reactor - Design description, operation regime and manipulation

International Nuclear Information System (INIS)

Badrljica, R.

1984-05-01

Technical water, reactor secondary coolant system is made of four parts: pumping station on the Danube; sedimentation facility in the village Vinca; coolant, heat exchangers and other elements within the RA reactor building; leading outer pipes and the stream Mlaka. All the four parts are connected to form a functional entirety, and each of them connected to cooling heat exchangers forms a partial functional system which enable the whole system to fulfill its fundamental task [sr

Characterization and testing of materials for nuclear reactors. Proceedings of a technical meeting

International Nuclear Information System (INIS)

2007-03-01

Nuclear techniques in general and neutrons based methods in particular have played and will continue to play an important role in research in materials science and technology. Today the world is looking at nuclear fission and nuclear fusion as the main sources of energy supply for the future. Research reactors have played a key role in the development of nuclear technology. A materials development programme will thus play a major role in the design and development of new nuclear power plants, for the extension of the life of operating reactors as well as for fusion reactors. Against this background, the IAEA had organized a Technical Meeting on Development, Characterization and Testing of Materials - With Special Reference to the Energy Sector under the activity on specific applications of research reactors. The meeting was held in Vienna, May 29- June 2, 2006. There was also participation by experts in techniques, complementary to neutrons. The participants for the technical meeting were experts in the utilization of nuclear techniques namely the high flux and medium flux research reactors, fusion research and positron annihilation. They presented the design, development and utilization of the facilities at their respective centres for materials characterization with main focus on materials for nuclear energy, both fission and fusion. In core irradiation of materials, development of instrument for residual stress measurement in large and / or irradiated specimen, neutron radiography for inspection of irradiated fuel, work on oxide dispersion strengthened (ODS) steels and SiC composites, relevant to future power systems were cited as application of nuclear techniques in fission reactors. The use of neutron scattering for helium bubbles in steel, application of positron annihilation to study helium bubbles in Cu, Ti-stabilized stainless steel and voidswelling studies etc. show that these techniques have an important role in the development of materials for energy

Lenin nuclear reactor research institute in the tenth five-year plan

International Nuclear Information System (INIS)

Tsykanov, V.A.; Kulov, E.V.

1980-01-01

Main tasks and research results of Lenin Nuclear Reactor Reseach Institute in the 10-th Five-Year Plan are considered. Main research achievements are noted in nuclear power, radiation material testing, accumulation of transuranium elements and investigation of their physicochemical properties at VK-50, BOR-60, SM-2, RBT-6 and MIR reactor plants and in material testing laboratories

Meeting of the Technical Working Group on Fast Reactors (TWG-FR) (39th annual meeting). Working material

Energy Technology Data Exchange (ETDEWEB)

NONE

2007-07-01

The 39th Annual Meeting of the Technical Working Group on Fast Reactors (TWG FR) was held from 15-19 May 2006 in Beijing, China, at the invitation of the China Institute of Atomic Energy (CIAEA). The meeting was attended by TWG-FR Members and Advisers from the following Member States (MS): Belgium (observer), Brazil, China, France, Germany, India, Italy, Japan, the Republic of Kazakhstan, the Republic of Korea, the Russian Federation, Sweden (observer), the United Kingdom, and the United States. Belarus, Switzerland, the European Commission, and OECD/NEA were unable to participate. Moreover, Prof. Carlo Rubbia, CERN director general emeritus, participated, upon IAEA invitation, in the meeting as distinguished scientist and IAEA expert. Mr. S.C. Chetal, from India (Indira Gandhi Centre for Atomic Research, IGCAR), was appointed chairman. The objectives of the meeting were to: - Exchange information on the national programmes on Fast Reactors (FR) and Accelerator Driven Systems (ADS); - Review the progress since the 38th TWG-FR Annual Meeting, including the status of the actions; - Consider meeting arrangements for 2006 and 2007; - Reviewed the Agency's ongoing information exchange and co-ordinated research activities in the technical fields relevant to the TWG-FR (FRs and ADS), as well as co-ordination of the TWG-FR's activities with other organizations; - Discuss future joint activities in view of the Agency's Programme and Budget Cycle 2008-2009 (and beyond)

Status of Technical and Vocational Education in Rural Institutions in ...

African Journals Online (AJOL)

This study was aimed at assessing the state of technical and vocational education in rural institutions in Delta State. Three research questions guided the study. A sample size of fifty (50) principals from 50 rural institutions in Delta State was used for the study. Data were collected using a questionnaire. The study reveals that ...

Assuring PSA technical adequacy for new advanced light water reactor designs

International Nuclear Information System (INIS)

Lutz, R.J.; Detar, H.L.; Schneider, R.E.

2012-01-01

The Probabilistic Safety Assessment (PSA) for an Advanced Light Water Reactor (ALWR) must exhibit a high level of technical adequacy, or technical quality, in order to be used as a reliable tool for making risk informed decisions concerning design and eventual operation of the plant. During the design phase, decisions on some design features may use the PSA as an input. Also, the PSA may be used as input to other operational decisions during plant design and construction including the development of procedures, development of technical specification limiting conditions for operation and scheduling of preventive maintenance activities. For the existing fleet of light water reactors (LWRs), PSA technical adequacy can be judged from wide ranging acceptance criteria such as the PRA (Probabilistic Risk Assessment) Standard in the United States of America that was developed jointly by the American Society of Mechanical Engineers (ASME) and the American Nuclear Society (ANS). However, the requirements for PRA technical adequacy in this PRA Standard assumes that the plant is built and has operation experience. Some of the requirements cannot be met for ALWRs in the design or construction phase and with no operational history. Key elements of a high level of technical adequacy include procedures, operator interviews, plant walk-downs and equipment reliability histories. The ability to include these key elements into the ALWR PSA to improve technical adequacy will progress as the ALWR progresses from the design stage through the construction stage and finally to the fuel load / pre-operational stage. As the technical adequacy becomes more robust, more confidence can be placed on risk-informed decisions that are made with the PSA. To assist in using the PSA as input to design and operational decisions in the design and construction stages of an ALWR, an addition to the ASME/ANS PRA Standard is being developed. The intent of this addition to the Standard is to provide

Conceptual designs of advanced fast reactor. Proceedings of a technical committee meeting

International Nuclear Information System (INIS)

1996-10-01

A Technical Committee meeting (TCM) was held on Conceptual Designs of Advanced Fast Power Reactors to review the lessons learned from the construction and operation of demonstration and near-commercial size plants. This TCM focused on design and development of advanced fast reactors and identified methodologies to evaluate the economic competitiveness and reliability of advanced projects. The Member States which participated in the TCM were at different stages of LMFR development. The Russian Federation, Japan and India had prototype and/or experimental LMFRs and continue with mature R and D programmes. China, the Republic of Korea and Brazil were at the beginning of LMFR development. Therefore the aims of the TCM were to obtain technical descriptions of different design approaches for experimental, prototype, demonstration and commercial LMFRs, and to describe the engineering judgements made in developing the design approaches. Refs, figs, tabs

Conceptual designs of advanced fast reactor. Proceedings of a technical committee meeting

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-10-01

A Technical Committee meeting (TCM) was held on Conceptual Designs of Advanced Fast Power Reactors to review the lessons learned from the construction and operation of demonstration and near-commercial size plants. This TCM focused on design and development of advanced fast reactors and identified methodologies to evaluate the economic competitiveness and reliability of advanced projects. The Member States which participated in the TCM were at different stages of LMFR development. The Russian Federation, Japan and India had prototype and/or experimental LMFRs and continue with mature R and D programmes. China, the Republic of Korea and Brazil were at the beginning of LMFR development. Therefore the aims of the TCM were to obtain technical descriptions of different design approaches for experimental, prototype, demonstration and commercial LMFRs, and to describe the engineering judgements made in developing the design approaches. Refs, figs, tabs.

The Physikalisch-Technische Bundesanstalt PTB (physical-technical Federal institution) - research institute and highest technical authority

International Nuclear Information System (INIS)

Klages, H.

1976-01-01

The PTB Braunschweig and Berlin is a Federal institution for the natural sciences and engineering and the highest technical authority for measurements. It is subject to the directions of the Federal Ministry for Economic Affairs. Its main tasks are representation, maintenance and development of physical units and, in connection with this, research, examinations, and granting permissions for calibration measuring equipment, as well as examinations of building types and permissions. The types of measuring equipment are represented. Many examinations are carried out on a voluntary basis. The advisory activities and the PTB's publications are also reported on. An organizational plan informs of the structure of the PTB. (orig.) [de

Research reactor instrumentation and control technology. Report of a technical committee meeting

International Nuclear Information System (INIS)

1997-10-01

The majority of research reactors operating today were put into operation 20 years ago, and some of them underwent modifications, upgrading and refurbishing since their construction to meet the requirements for higher neutron fluxes. However, a few of these ageing research reactors are still operating with their original instrumentation and control systems (I and C) which are important for reactor safety to guard against abnormal occurrences and reactor control involving startup, shutdown and power regulation. Worn and obsolete I and C systems cause operational problems as well as difficulties in obtaining replacement parts. In addition, satisfying the stringent safety conditions laid out by the nuclear regulatory bodies requires the modernization of research reactors I and C systems and integration of additional instrumentation units to the reactor. In order to clarify these issues and to provide some guidance to reactor operators on state-of-art technology and future trends for the I and C systems for research reactors, a Technical Committee Meeting on Technology and Trends for Research Reactor Instrumentation and Controls was held in Ljubljana, Slovenia, from 4 to 8 December 1995. This publication summarizes the discussions and recommendations resulting from that meeting. This is expected to benefit the research reactor operators planning I and C improvements. Refs, figs, tabs

Decommissioning strategy for the 'RA' research nuclear reactor at the 'Vinca' Institute

International Nuclear Information System (INIS)

Matausek, M.V.

2000-01-01

Adopting the global strategy for decommissioning of the research reactor RA at the Vinca Institute and preliminary planning of particular activities is necessary independently on the decision of the future status of this reactor, namely even in the case that it is decided to complete the modernization and to use the reactor again. In this paper the global decommissioning strategy for the RA reactor is proposed, as well as the optimal time schedule of particular activities, based on the relevant experiences from other countries (author) [sr

In-service inspection of nuclear reactor vessels and steam generators. Results and evolution of the technics

International Nuclear Information System (INIS)

Rapin, Michel; Saglio, Robert.

1978-01-01

Methods and original technics have been developed by the CEA for inspection of the primary coolant circuit of PWR. Multifrequency Eddy currents for inspection of steam generators tubes gudgeons and bolts; focussed ultrasonics to test all the welds of the reactor vessel and its cover of mixed welds of tanks and steam generators, pressurizer welds and gudgeons from the inside; gamma radiography of vessel mixed welds, televisual examination of the stainless steel lining of the reactor vessel and its cover. Use of these technics is made with specific automatic machines designed either for inspection of steam generator tubes or for complete inspection of the vessel. Several reactors were inspected with these devices [fr

String Theory clues for the low-$\\ell$ CMB ?

CERN Document Server

Kitazawa, N.

2015-05-29

"Brane Supersymmetry Breaking" is a peculiar string-scale mechanism that can unpair Bose and Fermi excitations in orientifold models. It results from the simultaneous presence, in the vacuum, of collections of D-branes and orientifolds that are not mutually BPS, and is closely tied to the scale of string excitations. It also leaves behind, for a mixing of dilaton and internal breathing mode, an exponential potential that is just too steep for a scalar to emerge from the initial singularity while descending it. As a result, in this class of models the scalar can generically bounce off the exponential wall, and this dynamics brings along, in the power spectrum, an infrared depression typically followed by a pre-inflationary peak. We elaborate on a possible link between this type of bounce and the low-$\\ell$ end of the CMB angular power spectrum. For the first 32 multipoles, one can reach a 50 % reduction in $\\chi^{\\,2}$ with respect to the standard $\\Lambda$CDM setting.

Results of 1989/90 research and development activities at KfK Institute for Reactor Components

International Nuclear Information System (INIS)

1991-03-01

R and D activities at IRB (Institut fuer Reaktorbauelemente - Institute for Reactor Components) are dedicated to thermodynamics and fluid dynamics. Emphasis is on the design of nuclear reactor and fusion reactor components. Environmental engineering was added recently. Most activities are applications-oriented. Fundamental investigations focus on energy research and energy technology. The activities are carried out in the framework of different projects (PKF/nuclear fusion, PSF/nuclear safety, PSU/pollution control). Points of main effort are the development of basic liquid-metal-cooled blanket solutions, investigations on natural convection in reactor ranks, and the cooling properties of future containments for pressurized water reactors in the case of nuclear fusion accidents. (orig./GL) [de

Development of TRITN : In-house technical information network of Technical Research Institute in The Chugoku Electric Power Co., Inc.

Science.gov (United States)

Ebara, Hisao

With the increase in information amount, progress in information network, diversified technology and change in in-house organization the Technical Research Institute has necessitated to shift the conventional technical database processing by use of personal computers into new system promised for future use. It enables to; provide classification system which meets the needs of the institute; execute distributed processing using workstations; make the low cost processing using commercial softwares; make networks using LAN, WAN personal computer communication, and provide primary materials (text data) on facsimiles using optical disks. Such features are those of technical information database system "TRITN" which is available to any persons and at any time. This paper describes how TRITN was developed and the system outline.

Sodium-cooled reactors, objectives, achieved technical state and development trends

International Nuclear Information System (INIS)

Wolff, U.

1988-01-01

The use of fossil fuels to cover the future world-wide energy demand alone would rapidly deplete these ressources, especially oil and gas. Today's knowledge suggests the enhanced exploitation of solar energy, nuclear fusion and the application of uranium in sodium-cooled breeder reactors as the alternative energies offering a great potential. The sodium-cooled reactor outdistances the other options in terms of development. Its technical feasibility and safe operation have been verified and its profitability appears to be possible when using today's technology. The verification of its profitability while maintaining a high safety level is the overriding task for the future. The paper discusses corresponding activities in the USA, the USSR, Japan and Western Europe. (orig.) [de

Technical and institutional capacity in local organisations to manage ...

African Journals Online (AJOL)

Technical and institutional capacity in local organisations to manage decentralised forest resources in Uganda. ... Southern Forests: a Journal of Forest Science ... to implement decentralised forest governance exists in local organisations through partnerships with other actors in the productive use of the available resources.

Commercial products and services of research reactors. Proceedings of a technical meeting

International Nuclear Information System (INIS)

2013-07-01

Although the number of operational research reactors is steadily decreasing, more than half of those that remain are greatly underutilized and, in most cases, underfunded. To continue to play a key role in the development of peaceful uses of nuclear technology, the remaining research reactors will need to provide useful products and services to private, national and regional customers, in some cases with adequate revenue generation for reliable, safe and secure facility management and operation. In the light of declining governmental financial support and the need for improved physical security and conversion to low enriched uranium (LEU) fuel, many research reactors have been challenged to generate income to offset increasing operational and maintenance costs. The renewed interest in nuclear power (and therefore in nuclear education and training), the global expansion of diagnostic and therapeutic nuclear medicine, and the extensive use of semiconductors in electronics and in other areas have created new opportunities for research reactors, prominent among them, markets for products and services in regions and countries without such facilities. It is clear that such initiatives towards greater self-reliance will need to address such aspects as market surveys, marketing and business plans, and cost of delivery services. It will also be important to better inform present and future potential end users of research reactor services of the capabilities and products that can be provided. This publication is a compilation of material from an IAEA technical meeting on “Commercial Products and Services of Research Reactors”, held in Vienna, Austria, from 28 June to 2 July 2010. The overall objective of the meeting was to exchange information on good practices and to provide concrete examples, in technical presentations and brainstorming discussions, to promote and facilitate the development of commercial applications of research reactors. The meeting also aimed to

Progress report on research and development in 1991, Institute of Neutron Physics and Reactor Engineering, KfK

International Nuclear Information System (INIS)

1992-03-01

Progress report on research and development in 1991 Institute of Neutron Physics and Reactor Engineering. The Institute of Neutron Physics and Reactor Engineering is concerned with research work in the field of nuclear engineering related to the safety of fast and thermal reactors as well as with specific problems of fusion reactor technology. Under the project of nuclear safety research, the Institute works on concepts designed to drastically improve reactor safety. Apart from that, methods to estimate and minimize the radiological consequences of reactor accidents are developed. Under the fusion technology project, the Institute deals with neutron physics and technological questions of the breeding blanket. Basic research covers technico-physical questions of the interaction between light ion radiation of a high energy density and matter. In addition and to a small extent, questions of employing hydrogen in the transport area are studied. For all these tasks it is indispensable to use up-to-date data processing methods and equipment, from the highest capacity computer to the integrated minicomputer system. (orig./DG) [de

Annual activities report of Brazilian Aerospace Technical Center -CTA/IEAv - 1989

International Nuclear Information System (INIS)

1989-01-01

This document reports the research activities on nuclear physics and reactors physics and engineering in the Brazilian Aerospace Technical Center/Advanced Studies Institute, Sao Paulo State, in the year of 1989

Report of the Office of Nuclear Reactor Regulation technical assistance task force

International Nuclear Information System (INIS)

1981-11-01

In 1981, the Director of the Office of Nuclear Reactor Regulation (NRR) of the Nuclear Regulatory Commission (NRC) chartered a task force to assess the office program of technical assistance and to recommend improvements. The task force divided the technical assistance program into four areas, and the practices in each area were assessed through a series of surveys of staff, management, and contractor personnel. The task force placed emphasis in its interview and assessment process on the problem areas that exist in the technical assistance program. The report thus reflects a weight on the faults found as a result of the inquiries made. The four major areas of technical assistance contracting studied were program planning, program management and execution, program control and management information systems, and program administration and coordination

A window on⋯ the national food institute, technical university of Denmark

DEFF Research Database (Denmark)

Wegener, Henrik Caspar

2011-01-01

The National Food Institute is an institute of the Technical University of Denmark. The Institute has a staff of 400, out of which approximately 275 hold an academic degree. It is divided into five Divisions; Chemistry, Toxicology, Microbiology, Nutrition, Industrial Food Research, and a Manageme...... composition database. The results of the research activities forms part of the evidence-base used to give advice to national and international authorities....

Evolutionary computing in Nuclear Engineering Institute/CNEN-Brazil

International Nuclear Information System (INIS)

Pereira, Claudio M.N.A.; Lapa, Celso M.F.; Lapa, Nelbia da Silva; Mol, Antonio C.

2000-01-01

This paper aims to discuss the importance of evolutionary computation (CE) for nuclear engineering and the development of this area in the Instituto de Engenharia Nuclear (IEN) at the last years. Are describe, briefly, the applications realized in this institute by the technical group of CE. For example: nuclear reactor core design optimization, preventive maintenance scheduling optimizing and nuclear reactor transient identifications. It is also shown a novel computational tool to implementation of genetic algorithm that was development in this institute and applied in those works. Some results were presents and the gains obtained with the evolutionary computation were discussing. (author)

Report of researches by common utilization of facilities in Kyoto University Research Reactor Institute, latter half of fiscal year 1982

International Nuclear Information System (INIS)

1983-01-01

The technical report of the Kyoto University Research Reactor Institute is published any time to immediately report on the results of the functional tests of various experimental facilities, the test results for the products made for trial, radiation control, the situation of waste treatment, the data required for research and experiment such as the reports of study meetings, the conspicuous results obtained amid researches, new processes, and the discussion on other papers and reports. In this report, the title, the names of reporters and the summary of 65 researches carried out by the common utilization of the facilities in the Kyoto University Research Reactor Institute are collected. The themes of the researches are such as Moessbauer spectroscopic study of ferrocene and its derivative iodides by I-129, decomposition of cadmium telluride during heat treatment, element distribution in resource living things and environmental substances produced in northern ocean, radioactivation analysis of trace elements in blood of tumor-bearing animals, radioactivation analysis of noble metal elements in geochemical samples, relaxation phenomena by gamma-gamma perturbation angle correlation, separation of components in Allende meteorite and their radioactivation analysis, measurement of cross section of Pa-231 (n, gamma) reaction and others. (Kako, I.)

Report of researches by common utilization of facilities in Kyoto University Research Reactor Institute, first half of fiscal year 1982

International Nuclear Information System (INIS)

1983-01-01

The technical report of the Kyoto University Research Reactor Institute is published any time to immediately report on the results of the functional tests of various experimental facilities, the test results for the products made for trial, radiation control, the situation of waste treatment, the data required for research and experiment such as the reports of study meetings, the conspicuous results obtained amid researches, new processes, and the discussion on other papers and reports. In this report, the title, the names of reporters and the summary of 47 researches carried out by the common utilization of the facilities in the Kyoto University Research Reactor Institute are collected. The themes of the researches are such as diffusion of impurities ion-implanted in silicon into natural oxide films, origin of igneous rocks by trace element distribution study, element distribution in black ore and its accompanying rocks and origin of black ore, reprocessing of molten salt fuel of thorium group, forerunning martensite transformation of Fe-Pt invar alloy, change of nucleic acid component to recoil tritium at cryogenic temperature, gamma irradiation effect of KC1 containing Pb 2+ , radiation effect on cadmium halide crystals and impurity metallic ions and others. (Kako, I.)

Utilization of nuclear research reactors

International Nuclear Information System (INIS)

1980-01-01

Full text: Report on an IAEA interregional training course, Budapest, Hungary, 5-30 November 1979. The course was attended by 19 participants from 16 Member States. Among the 28 training courses which the International Atomic Energy Agency organized within its 1979 programme of technical assistance was the Interregional Training Course on the Utilization of Nuclear Research Reactors. This course was held at the Nuclear Training Reactor (a low-power pool-type reactor) of the Technical University, Budapest, Hungary, from 5 to 30 November 1979 and it was complemented by a one-week Study Tour to the Nuclear Research Centre in Rossendorf near Dresden, German Democratic Republic. The training course was very successful, with 19 participants attending from 16 Member States - Bangladesh, Bolivia, Czechoslovakia, Ecuador, Egypt, India, Iraq, Korean Democratic People's Republic, Morocco, Peru, Philippines, Spain, Thailand, Turkey, Vietnam and Yugoslavia. Selected invited lecturers were recruited from the USA and Finland, as well as local scientists from Hungarian institutions. During the past two decades or so, many research reactors have been put into operation around the world, and the demand for well qualified personnel to run and fully utilize these facilities has increased accordingly. Several developing countries have already acquired small- and medium-size research reactors mainly for isotope production, research in various fields, and training, while others are presently at different stages of planning and installation. Through different sources of information, such as requests to the IAEA for fellowship awards and experts, it became apparent that many research reactors and their associated facilities are not being utilized to their full potential in many of the developing countries. One reason for this is the lack of a sufficient number of trained professionals who are well acquainted with all the capabilities that a research reactor can offer, both in research and

Experience in management of aging of research reactors in the Vinca Institute

International Nuclear Information System (INIS)

Pesic, M.; Cupac, S.; Vukadin, Z.

1997-03-01

This paper describes history, the current status and experimental possibilities of research reactors in the VINCA Institute, and summarises annual reports on their utilisation and maintenance. Operating problems as consequences of the ageing of the reactors' equipment and materials, including funding aspects and influence of changing of the nuclear programme in the country are discussed. (author)

Current status of light water reactor and Hitachi's technical improvements for BWR

International Nuclear Information System (INIS)

Miki, Minoru; Ohki, Arahiko.

1984-01-01

Gradual technical improvements in Japan over the years has improved the reliability of light water reactors, and has achieved the highest capacity factor level in the world. Commercial operation of Fukushima 2-2 (1,100 MW) of the Tokyo Electric Power Co. was started in February, 1984, as the first standardized BWR base plant, ushering in a new age of domestic light water reactor technology. The ABWR (1,300 MW class) has been developed as Japan's next generation light water reactor, with construction aimed at the latter half of the 1980's. Hitachi's extensive efforts range from key nuclear equipment to various related robots, directed at improving safety, reliability, and the capacity factor, while reducing radiation exposure. This paper presents an outline of Hitachi's participation in the light water reactor's improvement and standardization, and the current status of our role in the international cooperation plan for the ABWR. (author)

Competence Model and Modern Trends of Development of the Russian Institute of Technical Customer

Directory of Open Access Journals (Sweden)

Mishlanova Marina

2017-01-01

Full Text Available Article considers modern maintenance and development of the management actor by the investment-construction projects of the technical customer. Urgent problems of the formation of Institute of the technical customer establishment are allocated. Elementary competence model is presented: based competences of technical customer, model of the primary competence, example of the operational level of the model. Analysis of the development of the Institute of the technical customer was performed: compliance with current realities of investment-construction activities, improvement of contractual relations, compliance with international standards, state participation, creation of the single technical customer. Necessity of development of competence models for the urgent justification of professional standards is assessed. The possibility of modeling of the competencies and functions of technical customer in approach to the FIDIC-model was revealed. Possibility of usage of the competence model of the technical customer on the stage of building in terms of public-private partnership. Results show the direction for further researches.

Las Conferencias Lowell de Kuhn: Un estudio crítico

Directory of Open Access Journals (Sweden)

Juan Vicente Mayoral

2013-09-01

Full Text Available Algunas interpretaciones recientes de la obra de Thomas Kuhn subrayan su contribución a la preservación de aspectos centrales del positivismo lógico, como resultado de un desconocimiento de la corriente (y, más en general, de la filosofía de la ciencia de la época y de una crítica anti-positivista que, también por ese motivo, era superficial. En este artículo ofrezco argumentos en contra de dicha tesis partiendo de un texto inédito de Kuhn: sus conferencias en el Instituto Lowell de Boston en 1951, The Quest for Physical Theory, que fueron un primer intento de exhibir el punto de vista más tarde ofrecido en The Structure of Scientific Revolutions.

NRC review of Electric Power Research Institute's advanced light water reactor utility requirements document. Passive plant designs, chapter 1, project number 669

International Nuclear Information System (INIS)

1994-08-01

The Electric Power Research Institute (EPRI) is preparing a compendium of technical requirements, referred to as the open-quotes Advanced Light Water Reactor [ALWR] Utility Requirements Documentclose quotes, that is acceptable to the design of an ALWR power plant. When completed, this document is intended to be a comprehensive statement of utility requirements for the design, construction, and performance of an ALWR power plant for the 1990s and beyond. The Requirements Document consists of three volumes. Volume 1, open-quotes ALWR Policy and Summary of Top-Tier Requirementsclose quotes, is a management-level synopsis of the Requirements Document, including the design objectives and philosophy, the overall physical configuration and features of a future nuclear plant design, and the steps necessary to take the proposed ALWR design criteria beyond the conceptual design state to a completed, functioning power plant. Volume II consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant [approximately 1350 megawatts-electric (MWe)]. Volume III contains utility design requirements for nuclear plants for which passive features will be used in their designs (approximately 600 MWe). In April 1992, the staff of the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, issued Volume 1 and Volume 2 (Parts 1 and 2) of its safety evaluation report (SER) to document the results of its review of Volumes 1 and 2 of the Requirements Document. Volume 1, open-quotes NRC Review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document - Program Summaryclose quotes, provided a discussion of the overall purpose and scope of the Requirements Document, the background of the staff's review, the review approach used by the staff, and a summary of the policy and technical issues raised by the staff during its review

Balancing human and technical reliability in the design of advanced nuclear reactors

International Nuclear Information System (INIS)

Papin, Bernard

2011-01-01

Highlights: ► Human factors exigencies are often overseen during the early design phases of NPP. ► Optimization of reactors safety is only based on technical reliability considerations. ► The search for more technical reliability often leads to more system complexity. ► System complexity is a major contributor to the operator's poor performance. ► Our method enables to assess plant complexity and it's impact on human performance. - Abstract: The strong influence of human factors (HF) on the safety of nuclear facilities is nowadays recognised and the designers are now enforced to consider HF requirements in the design of new facilities. Yet, this consideration of human factors requirements is still more or less restricted to the latest phases of the projects, essentially for the design of human-system interfaces (HSI's) and control rooms, although the design options influencing at most the human performance in operation are indeed fixed during the very early phases of the new reactors projects. The main reason of this late consideration of HF is that there exist few methods and models for anticipating the influence of fundamental design options on the future performance of operation teams. This paper describes a set of new tools permitting (i) determination of the impact of the fundamental process design options on the future activity of the operation teams and (ii) assessment of the influence of these operational constraints on teams performance. These tools are intended to guide the design of future 4th generation (GEN4) reactors, within the frame of a global risk-informed design approach, considering technical and human reliability exigencies in a balanced way.

Lowell Public Library, Final Performance Report for Library Services and Construction Act (LSCA) Title VI, Library Literacy Program.

Science.gov (United States)

Maravilla, Virginia

The Lowell Public Library (Indiana) Adult Literacy Program expanded literacy efforts of the library and its volunteer tutors by increasing the program enrollment numbers of the functionally illiterate English-speaking, English as a Second Language (ESL), migrant workers, and Basic Math students; assisted students in achieving their stated goals in…

Annual report on operation, utilization and technical development of research reactors and hot laboratory

International Nuclear Information System (INIS)

1990-09-01

This report describes the activities of the Department of Research Reactor Operation in fiscal year of 1989. It also presents some technical topics on the reactor operation and utilization in details. The Department is responsible for operation of the research reactors, JRR-2 and JRR-4, and the Hot Laboratory. The research reactor JRR-3 was reconstructed to enhance the performance for utilization. The first criticality was achieved on March 22, 1989, and it subsequently went into operation. In connection with the reactor operation, the various research and development activities in the area of fuel management, water chemistry, radiation monitoring and material irradiation have been made. In the Hot Laboratory, post-irradiation examinations of fuels and materials have been carried out along with the development of related techniques. (author)

Tritium release during inspection of reactor 'RA' at 'Vinca' institute

International Nuclear Information System (INIS)

Sipka, V.; Miljevic, N.; Grsic, Z.; Todorovic, D.; Radenkovic, M.

1997-01-01

Tritium content in daily precipitation, atmospheric water vapor inside of the reactor hall and around 'Vinca' Institute as well as in soil up to 800 m distance was monitoring during the regular inspection of the fuel channels. Tritium activity in the reactor hall air moisture was in the range from 0.022 to 6.7 MBq/m 3 . Tritium content in soil moisture between 12.7 and 530.9 Bq/l indicate a certain contamination due to tritium release in the environment, depending on the depth and distance from the place of release (author) [sr

Annual report of Department of Research Reactor and Tandem Accelerator, JFY2011. Operation, utilization and technical development of JRR-3, JRR-4, NSRR and tandem accelerator

International Nuclear Information System (INIS)

Ishii, Tetsuro; Nakamura, Kiyoshi; Kawamata, Satoshi; Ishikuro, Yasuhiro; Kawashima, Kazuhito; Kabumoto, Hiroshi; Nakamura, Takemi; Tamura, Itaru; Kawasaki, Sayuri; Sataka, Masao

2013-03-01

The Department of Research Reactors and Tandem Accelerator is in charge of the operation, utilization and technical development of JRR-3(Japan Research Reactor No.3), JRR-4(Japan Research Reactor No.4), NSRR(Nuclear Safety Research Reactor) and Tandem Accelerator. This annual report describes a summary of activities of services and technical developments carried out in the period between April 1, 2011 and March 31, 2012. The activities were categorized into six service/development fields: (1) Recovery from the Great East Japan Earthquake, (2) Operation and maintenance of research reactors and tandem accelerator, (3) Utilization of research reactors and tandem accelerator, (4) Upgrading of utilization techniques of research reactors and tandem accelerator, (5) Safety administration for research reactors and tandem accelerator, (6) International cooperation. Also contained are lists of publications, meetings, granted permissions on lows and regulations concerning atomic energy, number of staff members dispatched to Fukushima for the technical assistance, commendation, outcomes in service and technical developments and so on. (author)

Innovative and practical technical development of nuclear energy. Efforts on proposal and recruitment type technical development of nuclear energy

International Nuclear Information System (INIS)

Matsui, Kazuaki; Shioiri, Akio; Hamada, Jun; Kanagawa, Takashi; Mori, Yukihide; Kouno, Koji

2003-01-01

In technical development of nuclear energy conceiving a view on energy environment problem at the 21st Century, technical development on innovative nuclear energy system as well as next generation LWR is an important subject. Even in Japan, on the 'Long-term program for research, development and utilization of nuclear energy (LPRNE)' summarized by the Atomic Energy Commission, investigation on R and Ds of innovative reactors under cooperation of government, industrial field, and universities is required. In the Energy Generalized Engineering Institute, by receiving a subsidy from the Ministry of Economy and Industry since 2000, a proposal recruitment business on innovative and practical technical development of nuclear energy has been carried out. Here were introduced hopeful and unique five themes out of them applied to the recruitment, such as a super-critical pressure water cooling reactor (SCPR), an integrated modular LWR (IMR): technical development for practice, technical development on general purpose boiling transitional analysis method, technical development on direct extraction of U and Pu from consumed fuels based on super-DIREX reprocessing method, and material transfer forecasting in natural barriers at landfill disposal of radioactive wastes. (G.K.)

Technical specification: Mixed-oxide pellets for the light-water reactor irradiation demonstration test

International Nuclear Information System (INIS)

Cowell, B.S.

1997-06-01

This technical specification is a Level 2 Document as defined in the Fissile Materials Disposition Program Light-Water Reactor Mixed-oxide Fuel Irradiation Test Project Plan. It is patterned after the pellet specification that was prepared by Atomic Energy of Canada, Limited, for use by Los Alamos National Laboratory in fabrication of the test fuel for the Parallex Project, adjusted as necessary to reflect the differences between the Canadian uranium-deuterium reactor and light-water reactor fuels. This specification and the associated engineering drawing are to be utilized only for preparation of test fuel as outlined in the accompanying Request for Quotation and for additional testing as directed by Oak Ridge National Laboratory or the Department of Energy

Advanced reactors: A retrospective

International Nuclear Information System (INIS)

Starr, C.

1989-01-01

The objectives for nuclear power have always emphasized competitive costs, reliability, and public safety. During its initial two decades, the nuclear reactor program was enthusiastically and generously supported by the public, government, and industry. In the subsequent decades this external support was substantially eroded by the growing public fears of catastrophic accidents, poor economic performance of many nuclear plants, regulatory constraints, and a plethora of engineering issues disclosed by plant operations. The technical and institutional histories are discussed with particular relevance to their influence on the framework for future development of the several proposed advance reactors

Triennial technical report - 1986, 1987, 1988 - Instituto de Engenharia Nuclear (IEN) -Dept. of Reactors (DERE)

International Nuclear Information System (INIS)

1989-01-01

The research activities developed during the period 1986, 1987 and 1988 by the Reactor Department of Brazilian Nuclear Energy Commission (CNEN-DERE) are summarized. The principal aim of the Department of Reactors is concerned to the study and development of fast reactors and research thermal reactors. The DERE also assists the CNEN in the areas related to analysis of power reactor structure; to teach Reactor Physics and Engineering at the University, and professional training to the Nuclear Engineering Institute. To develop its research activity the DERE has three big facilities: Argonauta reactor, CTS-1 sodium circuit, and water circuit. (M.I.)

Catalogue and classification of technical safety rules for light-water reactors and reprocessing plants

International Nuclear Information System (INIS)

Bloser, M.; Fichtner, N.; Neider, R.

1975-08-01

This report on the cataloguing and classification of technical rules for land-based light-water reactors and reprocessing plants contains a list of classified rules. The reasons for the classification system used are given and discussed

Technical committee meeting on evaluation of radioactive materials release and sodium fires in fast reactors

International Nuclear Information System (INIS)

1996-01-01

The objectives of the Technical Committee Meeting was to review the activities of research on radioactive materials release and sodium fires in fast reactors in each of the participating countries. It covered: out-of-pile experiments and analysis codes on source term; in-pile experiments on source term; core disruptive accidents; sodium leak experience in liquid metal fast reactors; evaluation of sodium fire; and aerosol behaviour

New control and safety rod unit for the training reactor of the Dresden Technical University

International Nuclear Information System (INIS)

Adam, E.; Schab, J.; Knorr, J.

1983-01-01

The extension of the experimental training of students at the training reactor AKR of the Dresden Technical University requires the reconstruction of the reactor with a new control and safety rod unit. The specific conditions at the AKR led to a new variant. Results of preliminary experiments, design and mode of operation of the first unit as well as hitherto gained operation experiences are presented. (author)

Technical committee meeting on evaluation of radioactive materials release and sodium fires in fast reactors

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-07-01

The objectives of the Technical Committee Meeting was to review the activities of research on radioactive materials release and sodium fires in fast reactors in each of the participating countries. It covered: out-of-pile experiments and analysis codes on source term; in-pile experiments on source term; core disruptive accidents; sodium leak experience in liquid metal fast reactors; evaluation of sodium fire; and aerosol behaviour.

The emergency organization of the Federal Institute for Reactor Research

International Nuclear Information System (INIS)

Brunner, H.; Huerlimann, T.

1977-01-01

The organization and means of the emergency organization of the Federal Institute for Reactor Research (EIR), the eldest and largest nuclear installation in Switzerland, are described. It consists of a central command group and the following emergency teams: fire brigade, radiation protection, first aid, control, operating teams (reactors, hot laboratory etc.). The radiation protection team is formed by the Health Physics Division and is discussed in detail. A description of the alarm system and the first actions to be taken in case of an emergency is given. The importance of frequent and well-planned exercises and of radio communication between the teams and the command group is stressed and the emergency training programme of the EIR School for Radiation Protection, operated by the Health Physics Division, is presented. A fortunate lack of incidents at EIR is partly compensated for by experience gained from emergency team assistance operations during incidents outside the Institute. (author)

NRC review of Electric Power Research Institute's advanced light water reactor utility requirements document. Passive plant designs, chapters 2-13, project number 669

International Nuclear Information System (INIS)

1994-08-01

The Electric Power Research Institute (EPRI) is preparing a compendium of technical requirements, referred to as the open-quotes Advanced Light Water Reactor [ALWR] Utility Requirements Documentclose quotes, that is acceptable to the design of an ALWR power plant. When completed, this document is intended to be a comprehensive statement of utility requirements for the design, construction, and performance of an ALWR power plant for the 1990s and beyond. The Requirements Document consists of three volumes. Volume I, open-quotes ALWR Policy and Summary of Top-Tier Requirementsclose quotes, is a management-level synopsis of the Requirements Document, including the design objectives and philosophy, the overall physical configuration and features of a future nuclear plant design, and the steps necessary to take the proposed ALWR design criteria beyond the conceptual design state to a completed, functioning power plant. Volume II consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant [approximately 1350 megawatts-electric (MWe)]. Volume III contains utility design requirements for nuclear plants for which passive features will be used in their designs (approximately 600 MWe). In April 1992, the staff of the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, issued Volume 1 and Volume 2 (Parts 1 and 2) of its safety evaluation report (SER) to document the results of its review of Volumes 1 and 2 of the Requirements Document. Volume 1, open-quotes NRC Review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document - Program Summaryclose quotes, provided a discussion of the overall purpose and scope of the Requirements Document, the background of the staff's review, the review approach used by the staff, and a summary of the policy and technical issues raised by the staff during its review

Technical Memory 2008. Nuclear Regulatory Authority

International Nuclear Information System (INIS)

2011-01-01

The technical memory 2008 of the Nuclear Regulatory Authority of Argentine Republic, compile the papers published in the subject on radiation protection and nuclear safety, and presented in journals, technical reports, congress or meetings of these specialties by personnel of the mentioned institution during 2008. In this edition the documents are presented on: environmental protection; transport of radioactive materials; regulations; research reactors and nuclear power plants; biological radiation effects; therapeutic uses of ionizing radiation and radioprotection of patients; internal dosimetry; physical dosimetry; knowledge management; radioactive waste management. [es

AREVA Technical Days (ATD) session 3: operations of the reactors and services division, technical and economic aspects

International Nuclear Information System (INIS)

2003-01-01

These technical days organized by the Areva Group aims to explain the group activities in a technological and economic point of view, to provide an outlook of worldwide energy trends and challenges and to present each of their businesses in a synthetic manner. This third session deals with the reactors technologies basics, the EPR and SWR 1000 issues and outlook, the nuclear systems of the future, the business opportunities and business models. (A.L.B.)

Presentation of the High-Flux Reactor of the Institut Laue-Langevin

International Nuclear Information System (INIS)

Guyon, H.

2006-01-01

Full text of publication follows: The High-Flux Reactor (HFR) of the Institut Laue-Langevin is the world's most intense source of neutrons for fundamental research. Thanks to its extremely compact core, which is made up of a single fuel element, the HFR is capable of producing a neutron flux of up to 1.5.10 15 n.cm -2 .s -1 with a moderate power output of 58 MW. Its heavy water reflector thermalizes these neutrons, giving them a wave length of the order of one angstrom. They then become an excellent tool for exploring the atomic structure of matter. In order to provide a full neutron spectrum, the reactor is also equipped with a hot source (a block of graphite heated to 2000 deg. C) and two cold sources (a volume of liquid deuterium at 25 K). All the reactor's components can be replaced and adapted in order to keep pace with both changing scientific needs and changing safety requirements. For example, in 1992 the reactor block was replaced, a second cold source was installed in 1985, and the beam tubes are replaced at regularly intervals and are also occasionally modified. In the same way, the reactor's civil engineering structures are currently being reinforced in order to comply with the reassessment of the reference earthquake spectra. Finally, the Institut Laue-Langevin's reactor is equipped with three solid containment barriers: - the fuel cladding: during the 35 years the reactor has been in operation, a cladding failure has never been detected; - the leak-tight primary cooling system: this is partly submerged in a pool which provides radiological shielding; - the double-wall containment: an overpressure of air is maintained between the inner reinforced concrete wall and the outer metal wall. The High-Flux Reactor is therefore all set to provide the scientific community with top quality service for the next 20 years to come, on a site which: - is home to the brightest synchrotron in the world (ESRF); - benefits from the microbiology expertise of the EMBL

Presentation of the High-Flux Reactor of the Institut Laue-Langevin

Energy Technology Data Exchange (ETDEWEB)

Guyon, H. [Institut Laue-Langevin, Grenoble (France)

2006-07-01

Full text of publication follows: The High-Flux Reactor (HFR) of the Institut Laue-Langevin is the world's most intense source of neutrons for fundamental research. Thanks to its extremely compact core, which is made up of a single fuel element, the HFR is capable of producing a neutron flux of up to 1.5.10{sup 15} n.cm{sup -2}.s{sup -1} with a moderate power output of 58 MW. Its heavy water reflector thermalizes these neutrons, giving them a wave length of the order of one angstrom. They then become an excellent tool for exploring the atomic structure of matter. In order to provide a full neutron spectrum, the reactor is also equipped with a hot source (a block of graphite heated to 2000 deg. C) and two cold sources (a volume of liquid deuterium at 25 K). All the reactor's components can be replaced and adapted in order to keep pace with both changing scientific needs and changing safety requirements. For example, in 1992 the reactor block was replaced, a second cold source was installed in 1985, and the beam tubes are replaced at regularly intervals and are also occasionally modified. In the same way, the reactor's civil engineering structures are currently being reinforced in order to comply with the reassessment of the reference earthquake spectra. Finally, the Institut Laue-Langevin's reactor is equipped with three solid containment barriers: - the fuel cladding: during the 35 years the reactor has been in operation, a cladding failure has never been detected; - the leak-tight primary cooling system: this is partly submerged in a pool which provides radiological shielding; - the double-wall containment: an overpressure of air is maintained between the inner reinforced concrete wall and the outer metal wall. The High-Flux Reactor is therefore all set to provide the scientific community with top quality service for the next 20 years to come, on a site which: - is home to the brightest synchrotron in the world (ESRF); - benefits from the

The research reactors their contribution to the reactors physics

International Nuclear Information System (INIS)

Barral, J.C.; Zaetta, A.; Johner, J.; Mathoniere, G.

2000-01-01

The 19 october 2000, the french society of nuclear energy organized a day on the research reactors. This associated report of the technical session, reactors physics, is presented in two parts. The first part deals with the annual meeting and groups general papers on the pressurized water reactors, the fast neutrons reactors and the fusion reactors industry. The second part presents more technical papers about the research programs, critical models, irradiation reactors (OSIRIS and Jules Horowitz) and computing tools. (A.L.B.)

Results of research and development activities in 1983 of the Institut fuer Neutronenphysik und Reaktortechnik

International Nuclear Information System (INIS)

1984-02-01

The Institute for Neutron Physics and Reactor Technology treats research problems of nuclear engineering mainly those that are related to the development of sodium-cooled fast breeder reactors and fusion reactor technology. The activities are in approximately equal parts of an experimental and theoretical nature. A great part of the research activities is performed in co-operation with other institutes and industrial groups in the framework of projects. For the Fast Breeder Reactor Project the Institute works on reactor physical design and safety problems by the core of large-scale fast breeder reactors. Questions concerning the consequences of accidents in lightwater reactors upon the environment and the population are treated as part of the Nuclear Safety project. The Institute contributes to the Reprocessing Project with theoretical investigations on the physics of the fuel cycle and by developing control devices for a reprocessing plant. In the framework of the Fusion Projekt the Institute is concerned with neutron physical and technological questions of the the breeder blanket. Independent institutional work concerns technical and physical questions on the interaction between light ion beams of a high energy density and substances. Apart from this, investigations are carried out on the application of nuclear energy in the field of traffic with hydrogen as energy carrier. (RW) [de

Annual report of Department of Research Reactor and Tandem Accelerator, JFY2007. Operation, utilization and technical development of JRR-3, JRR-4, NSRR and tandem accelerator

International Nuclear Information System (INIS)

Miyazaki, Osamu; Awa, Yasuaki; Isaka, Koji; Kutsukake, Kenichi; Komeda, Masao; Shibata, Ko; Hiyama, Kazuhisa; Suzuki, Mayu; Sone, Takuya; Ohuchi, Tomoaki; Terakado, Yuichi; Sataka, Masao

2009-06-01

The Department of Research Reactors and Tandem Accelerator is in charge of the operation, utilization and technical development of JRR-3(Japan Research Reactor-3), JRR-4(Japan Research Reactor-4), NSRR(Nuclear Safety Research Reactor) and Tandem Accelerator. This annual report describes a summary of activities of services and technical developments carried out in the period between April 1, 2007 and March 31, 2008. The activities were categorized into five service/development fields: (1) Operation and maintenance of research reactors and tandem accelerator. (2) Utilization of research reactors and tandem accelerator. (3) Upgrading of utilization techniques of research reactors and tandem accelerator. (4) Safety administration for research reactors and tandem accelerator. (5) International cooperation. Also contained are lists of publications, meetings, granted permissions on lows and regulations concerning atomic energy, commendation, plans and outcomes in service and technical developments and so on. (author)

Annual report of Department of Research Reactor and Tandem Accelerator, JFY2010. Operation, utilization and technical development of JRR-3, JRR-4, NSRR and Tandem Accelerator

International Nuclear Information System (INIS)

Ishii, Tetsuro; Nakamura, Kiyoshi; Kawamata, Satoshi; Yamada, Yusuke; Kawashima, Kazuhiro; Asozu, Takuhiro; Nakamura, Takemi; Arai, Masaji; Yoshinari, Shuji; Sataka, Masao

2012-03-01

The Department of Research Reactors and Tandem Accelerator is in charge of the operation, utilization and technical development of JRR-3(Japan Research Reactor No.3), JRR-4(Japan Research Reactor No.4), NSRR(Nuclear Safety Research Reactor) and Tandem Accelerator. This annual report describes a summary of activities of services and technical developments carried out in the period between April 1, 2010 and March 31, 2011. The activities were categorized into five service/development fields: (1) Operation and maintenance of research reactors and tandem accelerator, (2) Utilization of research reactors and tandem accelerator, (3) Upgrading of utilization techniques of research reactors and tandem accelerator, (4) Safety administration for research reactors and tandem accelerator, (5) International cooperation. Also contained are lists of publications, meetings, granted permissions on lows and regulations concerning atomic energy, commendation, outcomes in service and technical developments and so on. (author)

General meeting. Technical reunion: the numerical and experimental simulation applied to the Reactor Physics

International Nuclear Information System (INIS)

2001-10-01

The SFEN (French Society on Nuclear Energy), organized the 18 october 2001 at Paris, a technical day on the numerical and experimental simulation, applied to the reactor Physics. Nine aspects were discussed, giving a state of the art in the domain:the french nuclear park; the future technology; the controlled thermonuclear fusion; the new organizations and their implications on the research and development programs; Framatome-ANP markets and industrial code packages; reactor core simulation at high temperature; software architecture; SALOME; DESCARTES. (A.L.B.)

Important functions and development ideas of the library and information work in scientific and technical research institutes

International Nuclear Information System (INIS)

Ye Qingsen

2010-01-01

The library and information branch is an important constituent part of a scientific and technic research institute, and is a technically assistant department which is set for scientific and technic research and production. The investigations show that the achievements by the library and information departments are marked in scientific and technic research institutes, and the library and information works play important roles for the existence, development and innovation of the institute in past years. But, the present conditions and statuses are dropped behind, and more problems existed in the library and information departments as a whole. The development ideas are proposed for the library and information work by analyzing the characteristics of nowadays knowledge service, knowledge economy and network information time. (authors)

Mirror power reactor magnet coil system: a technically and economically feasible design

International Nuclear Information System (INIS)

Peterson, M.A.

1977-01-01

The design and preliminary engineering analysis of a ''Yin Yang'' coil system utilizing several original design concepts to achieve technical and economic feasibility will be presented. The design analysis is begun with a general description of the constraints and prerequisites which define the problem of designing a satisfactory coil system for a mirror power reactor. This description includes a discussion of the coil conductor geometry required by plasma physics considerations, and also a description of the magnitude and direction of the magnetic force system distributed over the conductor geometry. In addition, the important design constraints which all mirror coil system designs must satisfy if they are to successfully interface with the other reactor components are reviewed. After considering the basic constraints that Yin Yong coil systems must be developed around, a survey of the various design concepts that were developed and explored in search of a satisfactory coil system design is discussed. From this extensive preliminary investigation of potential coil system configurations, a coil system design was developed which appears to offer by far the best combination of technical and economic feasibility of any other coil system design developed thus far

Safety considerations of new critical assembly for the Research Reactor Institute, Kyoto University

International Nuclear Information System (INIS)

Umeda, Iwao; Matsuoka, Naomi; Harada, Yoshihiko; Miyamoto, Keiji; Kanazawa, Takashi

1975-01-01

The new critical assembly type of nuclear reactor having three cores for the first time in the world was completed successfully at the Research Reactor Institute of Kyoto University in autumn of 1974. It is called KUCA (Kyoto University Critical Assembly). Safety of the critical assembly was considered sufficiently in consequence of discussions between the researchers of the institute and the design group of our company, and then many bright ideas were created through the discussions. This paper is described the new safety design of main equipments - oil pressure type center core drive mechanism, removable water overflow mechanism, core division mechanism, control rod drive mechansim, protection instrumentation system and interlock key system - for the critical assembly. (author)

IAEA Technical Meeting on Innovative Heat Exchanger and Steam Generator Designs for Fast Reactors. Presentations

International Nuclear Information System (INIS)

2011-01-01

concepts incorporating innovative systems and components, as well as advanced fuel and fuel cycle technologies. In particular, innovative heat exchangers and steam generators are key to significanly reduce the capital cost of the NSSS of the fast reactors. The IAEA, within the framework of its Nuclear Energy Department’s Technical Working Group on Fast Reactors (TWG-FR), assists Member States activities in these technology development areas by providing an umbrella for information exchange [topical Technical Meetings (TMs), Workshops and large Conferences] and collaborative R&D [Coordinated Research Projects (CRPs)]. This topical TM is addressing Member States’ expressed information exchange needs in the field of advanced fast reactor design features, with particular attention to innovative heat exchangers and steam generators

Characterization and management of radioactive sodium and other reactor components as input data for the decommissioning of liquid metal-cooled fast reactors. A compilation of data produced of data produced by members of the IAEA technical working group on fast reactors (TWG-FR) at two consultancies and one technical committee meeting. Working material

International Nuclear Information System (INIS)

2002-01-01

A number of liquid metal cooled fast reactors (LMFRs) are in operation and, some have already been shut down; other reactors will reach the end of their design lifetime in a few years and become candidates for decommissioning. It is unfortunate that little consideration was devoted to decommissioning of reactors at the plant design and construction stage. It is with this focus that the Technical Working Group on Fast Reactors (TWGFR) recommended that the IAEA organize the exchange of information on LMFRs decommissioning technology. It was pointed out that the decommissioning of small sodium-cooled reactors has shown that there are two basic differences between thermal and fast reactors decommissioning: on the one side, the treatment and disposal of radioactive sodium coolant, and on the other side, the management of reactor components, for which the structural materials are activated in depth by fast neutrons. To this end, a Technical Committee Meeting on Sodium Removal and Disposal from LMFRs in Normal Operation and in the framework of Decommissioning (Aix-en-Provence, France, November 1997) and two Consultancies on Decommissioning of the Kazakh BN-350 LMFR (Vienna, Austria, October 1996; Obninsk, Russian Federation, February 1998) were convened by the IAEA. These Meetings brought together a group of experts from France, Russia, Kazakhstan, the UK, and the USA to exchange information on, and to review current technical knowledge and experience in the management of radioactive coolant and reactor components following closing of LMFRs, as well as their design features and operating experience relevant for decommissioning procedures. The report provides general and detailed information on activation characteristics of the primary coolant; treatment and disposal of the spent sodium; removal of the residual sodium deposits and decontamination; the activation characteristics of the reactor components and the management of the latter. The recurring theme is finding

Pharmaceutical compounds in Merrimack River water used for public supply, Lowell, Massachusetts, 2008-09

Science.gov (United States)

Massey, Andrew J.; Waldron, Marcus C.

2011-01-01

This report presents results of a study conducted by the U.S. Geological Survey (USGS), in cooperation with the Massachusetts Department of Environmental Protection, to determine the occurrence of 14 commonly used human-health pharmaceutical compounds and fecal-indicator bacteria in Merrimack River water used as a drinking-water source by 135,000 residents in eastern Massachusetts. The study was designed to complement the USGS National Water-Quality Assessment Program's Source Water-Quality Assessment, which identifies patterns of occurrence of 280 primarily unregulated organic wastewater contaminants in source water used by community water systems and determines whether these patterns also occur in treated drinking water prior to distribution. The study involved periodic collection and analysis of raw Merrimack River water and treated drinking water over the course of 1 year. Water samples were collected periodically without regard to flow regime or antecedent weather conditions at the Lowell Regional Water Utility's Merrimack River intake upstream from Lowell, Mass. The same parcel of water was then sampled as finished water following treatment. Despite the presence of many potential sources of contamination in the drinking-water source area, only 2 of the 14 pharmaceutical analytes were detected at reportable concentrations in the source-water samples, and these occurred in only one set of periodic samples. Acetaminophen, a nonprescription analgesic, and caffeine were detected in the September source-water samples at concentrations of 0.084 and 0.068 micrograms per liter, respectively. Three other compounds-carbamazepine, an antiepileptic; cotinine, a metabolite of nicotine; and diphenhydramine, a nonprescription antihistamine-were detected in source-water samples, but at concentrations too low to be reliably quantified. None of the 14 pharmaceuticals was found in the finished water at a reportable concentration, defined as two times the long-term detection

Karlsruhe Nuclear Research Center, Institute of Neutron Physics and Reactor Engineering. Progress report on research and development work in 1993

International Nuclear Information System (INIS)

1994-03-01

The Institute of Neutron Physics and Reactor Engineering is concerned with research work in the field of nuclear engineering related to the safety of thermal reactors as well as with specific problems of fusion reactor technology. Under the project of nuclear safety research, the Institute works on concepts designed to drastically improve reactor safety. Apart from that, methods to estimate and minimize the radiological consequences of reactor accidents are developed. Under the fusion technology project, the Institute deals with neutron physics and technological questions of the breeding blanket. Basic research covers technico-physical questions of the interaction between light ion radiation of a high energy density and matter. In addition and to a small extent, questions of employing hydrogen in the transport area are studied. (orig.) [de

Nuclear reactor and materials science research: Technical report, May 1, 1985-September 30, 1986

International Nuclear Information System (INIS)

1987-01-01

Throughout the 17-month period of its grant, May 1, 1985-September 30, 1986, the MIT Research Reactor (MITR-II) was operated in support of research and academic programs in the physical and life sciences and in related engineering fields. The reactor was operated 4115 hours during FY 1986 and for 6080 hours during the entire 17-month period, an average of 82 hours per week. Utilization of the reactor during that period may be classified as follows: neutron beam tube research; nuclear materials research and development; radiochemistry and trace analysis; nuclear medicine; radiation health physics; computer control of reactors; dose reduction in nuclear power reactors; reactor irradiations and services for groups outside MIT; MIT Research Reactor. Data on the above utilization for FY 1986 show that the MIT Nuclear Reactor Laboratory (NRL) engaged in joint activities with nine academic departments and interdepartmental laboratories at MIT, the Charles Stark Draper Laboratory in Cambridge, and 22 other universities and nonprofit research institutions, such as teaching hospitals

Annual report of Department of Research Reactor and Tandem Accelerator, JFY2012. Operation, utilization and technical development of JRR-3, JRR-4, NSRR, Tandem Accelerator and RI Production Facility

International Nuclear Information System (INIS)

Murayama, Yoji; Ishii, Tetsuro; Nakamura, Kiyoshi; Uno, Yuki; Ishikuro, Yasuhiro; Kawashima, Kazuhito; Ishizaki, Nobuhiro; Matsumura, Taichi; Nagahori, Kazuhisa; Odauchi, Shouji; Maruo, Takeshi

2014-03-01

The Department of Research Reactor and Tandem Accelerator is in charge of the operation, utilization and technical development of JRR-3(Japan Research Reactor No.3), JRR-4(Japan Research Reactor No.4), NSRR(Nuclear Safety Research Reactor), Tandem Accelerator and RI Production Facility. This annual report describes a summary of activities of services and technical developments carried out in the period between April 1, 2012 and March 31, 2013. The activities were categorized into five service/development fields: (1) Operation and maintenance of research reactors and tandem accelerator, (2) Utilization of research reactors and tandem accelerator, (3) Upgrading of utilization techniques of research reactors and tandem accelerator, (4) Safety administration for department of research reactor and tandem accelerator, (5) International cooperation. Also contained are lists of publications, meetings, granted permissions on laws and regulations concerning atomic energy, number of staff members dispatched to Fukushima for the technical assistance, outcomes in service and technical developments and so on. (author)

Some technical constraints on possible Tokamak machines from next generation to reactor size

International Nuclear Information System (INIS)

Knobloch, A.

1975-11-01

A simplified consistent scaling of possible Tokamak reactors is set up in the power range of 0.1 - 10 GW. The influence of some important parameters on the scaling is shown and the role of some technical constraints is discussed. The scaling is evaluated for the two cases of a circular and a strongly elongated plasma section. (orig.) [de

Control and instrumentation system of the Zero Power Reactor at IEA, Sao Paulo (Brazil)

International Nuclear Information System (INIS)

Peluso, M.A.V.; Matsuda, K.; Hukai, R.

1974-01-01

The control and instrumentation system of the Zero Power Reactor at the IEA (Institute of Atomic Energy - Sao Paulo, Brazil) is described. Technical specifications of the main items of equipment are presented in a general way. Information is also given on the connection between the system described and the electrical supply system of the IEA reactor physics laboratory [pt

Integrating ethical and technical considerations in the energy debate

International Nuclear Information System (INIS)

Pickering, G.W.

1980-01-01

A study being carried out by the author on analyzing and integrating ethical and technical considerations in formulating an energy policy for the United States, with special emphasis on nuclear issues, is reviewed. Beginning with an overview of the historical development of the American nuclear program and the emergence of an organized politcal opposition to that program, seven outstanding issues are identified: institutional adequacy, radiation, reactor safety, waste disposal, economics and reliability, international consequences, and alternative energy sources, and an attempt is made to identify the range of dispute in relation to each. A framework is proposed for analyzing these issues by introducing guidelines for defining a social justice problem. Suggestions are made about how to get morally serious about such issues, and this framework is applied to four issues: radiation, reactor safety, waste disposal, and institutional adequacy. Ultimately it is a political choice whether or not to include nuclear power in a nations's energy policy. It is important, however, that such choices be made using the best technical and ethical assessments of the consequences. (LL)

The Marketability of Technical Graduates from Higher Educational Institutions (HEIs) Offering Technical and Vocational Education and Training (TVET): A Case from Malaysia

Science.gov (United States)

Rajadurai, Jegatheesan; Sapuan, Noraina Mazuin; Daud, Salina; Abidin, Nurazariah

2018-01-01

Technical, Vocational Education and Training has been viewed as a means of developing a nation. The marketability of technical graduates is reliant on whether these graduates possess the attributes demanded by their respective industries. Hence, this study aims to investigate the gap between the key attributes of Higher Education Institutions'…

Accident Tolerant Fuel Concepts for Light Water Reactors. Proceedings of a Technical Meeting

International Nuclear Information System (INIS)

2016-06-01

Nuclear fuel is a highly complex material that has been subject to continuous development over the past 40 years and has reached a stage where it can be safely and reliably irradiated up to 65 GWd/tU in commercial nuclear reactors. During this time, there have been many improvements to the original designs and materials used. However, the basic design of uranium oxide fuel pellets clad with zirconium alloy tubing has remained the fuel choice for the vast majority of commercial nuclear power plants. Severe accidents, such as those at the Three Mile Island and Fukushima Daiichi have shown that under such extreme conditions, nuclear fuel will fail and the high temperature reactions between zirconoi alloys and water will lead to the generation of hydrogen, with the potential for explosions to occur, daming the plant further. Recognizing that the current fuel designs are vulnerable to severe accident conditions, tehre is renewed interesst in alternative fuel designs that would be more resistant to fuel failure and hydrogen production. Such new fuel designs will need to be compatible with existing fuel and reactor systems if they are to be utilized in the current reactor fleet and in current new build designs, but there is also the possibility of new designs for new reactor systems. This publication provides a record of the Technical Meeting on Accident Tolerant Fuel Concepts for Light Water Reactors, held at Oak Ridge National Laboratories (ORNL), United States of America, 13-16 October 2014, to consider the early stages of research and development into accident tolerant fuel. There were 45 participants from 10 countries taking part in the meeting, with 32 papers organized into 7 sessions, of which 27 are included in this publication. This meeting is part of a wider investigation into such designs, and it is anticipated that further Technical Meetings and research programmes will be undertaken in this field

Reactor safety-a New Approach

International Nuclear Information System (INIS)

Machiels, A.J.; Marston, T.U.; Taylor, J.J.

1993-01-01

Since 1982, the U.S. utilities have been leading to an industry-wide effort to establish a technical foundation for the design of the next generation of light water reactors (LWRs) in the United States. Since 1985, the utility initiative has been effected through a major technical program managed by the Electric Power Research Institute (EPRI): the Advanced Light Water Reactor (ALWR) Program. In addition to the U.S. utility leadership and sponsorship, the ALWR Program has also greatly benefitted from the participation and sponsorship of numerous international utility companies and from the close cooperation with the U.S. Department of Energy (DOE). One of the main goals of the ALWR Program has been to develop a comprehensive set of design requirements for the advanced LWRs. The Utility Requirement Document (URD) defines the technical basis for improved and standardized future LWR designs. The URD covers the entire plant up to the grid interface. Therefore, it is the basis for an integrated plant design, i.e., nuclear steam supply system and balance of plant. It emphasizes those areas which are most important to the objective of achieving an advanced LWR that is excellent with respect to safety, performance, constructibility, and economics. There are numerous basic design policies underlying the ALWR URD. Of particular interest is the treatment of reactor safety

Primary coolant pipe rupture event in liquid metal cooled reactors. Proceedings of a technical meeting

International Nuclear Information System (INIS)

2004-08-01

In liquid-metal cooled fast reactors (LMFR) the primary coolant pipes (PCP) connect the primary coolant pumps to the grid plate. A rupture in one of these pipes could cause significant loss of coolant flow to the core with severe consequences. In loop type reactors, all primary pipelines are provided with double envelopes and inter-space coolant leak monitoring systems that permit leak detection before break. Thus, the PCP rupture event can be placed in the beyond design basis event (BDBE) category. Such an arrangement is difficult to incorporate for pool type reactors, and hence it could be argued that the PCP rupture event needs to be analysed in detail as a design basis event (DBE, category 4 event). However, the primary coolant pipes are made of ductile austenitic stainless steel material and operate at temperatures of the cold pool and at comparatively low pressures. For such low stressed piping with negligible creep and embrittlement effects, it is of interest to discuss under what design provisions, for pool type reactors, the guillotine rupture of PCP could be placed in the BDBE category. The topical Technical Meeting (TM) on Primary Coolant Pipe Rupture Event in Liquid Metal Cooled Reactors (Indira Gandhi Centre for Atomic Research, Kalpakkam, India, 13-17 January 2003) was called to enable the specialists to present the philosophy and analyses applied on this topic in the various Member States for different LMFRs. The scope of the technical meeting was to provide a global forum for information exchange on the philosophy applied in the various participating Member States and the analyses performed for different LMFRs with regard to the primary coolant pipe rupture event. More specifically, the objectives of the technical meeting were to review the safety philosophy for the PCP rupture event in pool type LMFR, to assess the structural reliability of the PCP and the probability of rupture under different conditions (with/without in-service inspection), to

Technical survey of decommissioning of commercial power reactors

International Nuclear Information System (INIS)

Nakamura, Masahide

2003-01-01

The technical survey of decommissioning of commercial power reactors had been carried out from 1982 to 2003. The investigation items are scenarios, procedures, simplification and recycling. On the scenarios, the case studies on the decommissioning steps (1983 to 1984), evaluation of the prior conditions of case studies (1994 to 1998), evaluation of rationalization of the scenarios of decommissioning steps (1999 to 2001) and evaluation of the effects of investigation of clearance level (1999 to 2002) are described. Procedures (1985 to 1996) and simplification (1985 to 1987) of decommissioning are investigated. On the recycling, survey on recycle of waste produced by the decommissioning step (1985 to 1993) and recycle of demolition waste (1997 to 2002) are reported. Recycle of radioactive waste has to be controlled under lows. (S.Y.)

Critical technical issues and evaluation and comparison studies for inertial fusion energy reactors

International Nuclear Information System (INIS)

Abdou, M.A.; Ying, A.Y.; Tillack, M.S.; Ghoniem, N.M.; Waganer, L.M.; Driemeyer, D.E.; Linford, G.J.; Drake, D.J.

1994-01-01

The critical issues, evaluation and comparison of two inertial fusion energy (IFE) reactor design concepts developed in the Prometheus studies are presented. The objectives were (1) to identify and characterize the critical issues and the R and D required to solve them, and (2) to establish a sound basis for future IFE technical and programmatic decisions by evaluating and comparing the different design concepts. Quantitative evaluation and comparison of the two design options have been made with special focus on physics feasibility, engineering feasibility, economics, safety and environment, and research and development (R and D) requirements. Two key conclusions are made based on the overall evaluation analysis: (1) The heavy-ion driven reactors appear to have an overall advantage over laser-driven reactors; and: (2) The differences in scores are not large and future results of R and D could change the overall ranking of the two IFE concepts

AN IMPECUNIOUS PILOT: BRITISH BUSINESS CULTURE AND THE CASE OF LOWELL YEREX, 1933-1946

Directory of Open Access Journals (Sweden)

Erik Benson

2003-01-01

Full Text Available This article examines the relationship between a New Zealand-born entrepreneur, Lowell Yerex, and the British government. It will give particular attention to the role of British business culture in shaping the government’s response to Yerex. During the 1930s and l940s, Yerex built an important airline network in the Caribbean Basin. A British subject, he had an affinity for the empire. However, the British rebuffed his repeated efforts to form an alliance. While they had political, economic and strategic reasons for doing so, the British also manifested a disdain for Yerex because he did not measure up to their business ideals.

Technical meeting (TM) to 'Review of national programmes on fast reactors and accelerator driven systems (ADS)'. Technical Working Group on Fast Reactors (TWG-FR) (37th annual meeting). Working material

Energy Technology Data Exchange (ETDEWEB)

NONE

2004-07-01

The objectives of the 37th Annual Meeting of the Technical Working Group on Fast Reactors, were to: 1) exchange information on the national programmes on Fast Reactors (FR) and Accelerator Driven Systems (ADS); 2) review the progress since the 36th TWG-FR Annual Meeting, including the status of the actions; 3) consider meeting arrangements for 2004 and 2005; 4) review the Agency's co-ordinated research activities in the field of FRs and ADS, as well as co-ordination of the TWG-FR's activities with other organizations. The participants made presentations on the status of the respective national programmes on FR and ADS development. A summary of the highlights for the period since the 36th TWG-FR Annual Meeting is included in this proceedings. Annex IV contains the Review of National Programs on Fast Reactors and Accelerator Driven Systems (ADS), and the TWG-FR Activity Report for the Period May 2003-April 2004.

The substantive core of introduction of the saving production into educational process of the higher technical educational institution

Directory of Open Access Journals (Sweden)

Світлана Станіславівна Єрмакова

2016-03-01

Full Text Available Problem and topicality of research consist in the study of theoretical and methodological problems at using the base of saving production in the higher educational institutions taking into account the special features of its activity, specificity of problems and tasks.Aim of the article was the grounding of the system of development of the higher technical education on the principles of economy. The problems consist in disclosure and substantiation of the elements of development of the higher technical education system taking into account the principles of saving production; determination of the possible use of saving production instruments in higher education sphere especially in the modern educational institution.Methods of research were the following: methods of analysis, prognostication, diagnostics, observation, pedagogical experiment.The concretized essence of saving use of the higher technical educational institution resources embodied in experimental monitoring technology of professional training of the future specialists in the higher technological educational institution on the base of idea of saving production. There was proved pedagogical expediency of advanced training of the future specialists in the process of their professional training in the higher technical educational institution on the base of saving production at its professional activity. There were substantiated: the lean-technology of professional training in the higher technical educational institution in the context of idea of continuous education based on principles of saving “production of knowledge” is the system of continuous monitoring of educational process effectiveness and of the saving use of its possibilities in the higher technical educational institution as the dominant of the future engineers training that is directed on the permanent revelation and neutralization of educational losses for the development of professionalism and creation at the expense

Technical committee meeting on material-coolant interactions and material movement and relocation in liquid metal fast reactors

International Nuclear Information System (INIS)

1994-01-01

The Technical Committee Meeting on Material-Coolant Interactions and Material Movement and Relocation in Liquid Metal Fast Reactors was sponsored by the International Working Group on Fast Reactors (IWGFR), International Atomic Energy Agency (IAEA) and hosted by PNC, on behalf of the Japanese government. A broad range of technical subjects was discussed in the TCM, covering entire aspects of material motion and interactions relevant to the safety of LMFRs. Recent achievement and current status in research and development in this area were presented including European out-of-pile test of molten material movement and relocation; molten material-sodium interaction; molten fuel-coolant interaction; core disruptive accidents; sodium boiling; post accident material relocation, heat removal and relevant experiments already performed or planned

Technical committee meeting on material-coolant interactions and material movement and relocation in liquid metal fast reactors

Energy Technology Data Exchange (ETDEWEB)

NONE

1994-07-01

The Technical Committee Meeting on Material-Coolant Interactions and Material Movement and Relocation in Liquid Metal Fast Reactors was sponsored by the International Working Group on Fast Reactors (IWGFR), International Atomic Energy Agency (IAEA) and hosted by PNC, on behalf of the Japanese government. A broad range of technical subjects was discussed in the TCM, covering entire aspects of material motion and interactions relevant to the safety of LMFRs. Recent achievement and current status in research and development in this area were presented including European out-of-pile test of molten material movement and relocation; molten material-sodium interaction; molten fuel-coolant interaction; core disruptive accidents; sodium boiling; post accident material relocation, heat removal and relevant experiments already performed or planned.

Project Experiences in Research Reactor Ageing Management, Modernization and Refurbishment. Report of a Technical Meeting on Research Reactor Ageing Management, Modernization and Refurbishment

International Nuclear Information System (INIS)

2014-08-01

performance, or provide new products and services to existing or potential users and customers, as well as share information on the technical details of the work involved, will allow other organizations contemplating similar work to consider and better understand their own challenges. Such understanding can help to optimize future planning in terms of budget, schedule and resource expectations, and to enable informed decision making. The IAEA is working to systematically collect existing knowledge on research reactor ageing management, modernization and refurbishment to share within the community of reactor owners, operators and oversight authorities. To this end, experts from the research reactor community gathered at IAEA headquarters from 10–14 October 2011 for the Technical Meeting on Research Reactor Ageing, Modernization and Refurbishment. The outcomes of the meeting and the submissions of the participants are documented in this publication

Project Experiences in Research Reactor Ageing Management, Modernization and Refurbishment. Report of a Technical Meeting on Research Reactor Ageing Management, Modernization and Refurbishment

Energy Technology Data Exchange (ETDEWEB)

NONE

2014-08-15

performance, or provide new products and services to existing or potential users and customers, as well as share information on the technical details of the work involved, will allow other organizations contemplating similar work to consider and better understand their own challenges. Such understanding can help to optimize future planning in terms of budget, schedule and resource expectations, and to enable informed decision making. The IAEA is working to systematically collect existing knowledge on research reactor ageing management, modernization and refurbishment to share within the community of reactor owners, operators and oversight authorities. To this end, experts from the research reactor community gathered at IAEA headquarters from 10–14 October 2011 for the Technical Meeting on Research Reactor Ageing, Modernization and Refurbishment. The outcomes of the meeting and the submissions of the participants are documented in this publication.

Research Institute for Technical Careers

Science.gov (United States)

Glenn, Ronald L.

1996-01-01

The NASA research grant to Wilberforce University enabled us to establish the Research Institute for Technical Careers (RITC) in order to improve the teaching of science and engineering at Wilberforce. The major components of the research grant are infrastructure development, establishment of the Wilberforce Intensive Summer Experience (WISE), and Joint Research Collaborations with NASA Scientists. (A) Infrastructure Development. The NASA grant has enabled us to improve the standard of our chemistry laboratory and establish the electronics, design, and robotics laboratories. These laboratories have significantly improved the level of instruction at Wilberforce University. (B) Wilberforce Intensive Summer Experience (WISE). The WISE program is a science and engineering bridge program for prefreshman students. It is an intensive academic experience designed to strengthen students' knowledge in mathematics, science, engineering, computing skills, and writing. (C) Joint Collaboration. Another feature of the grant is research collaborations between NASA Scientists and Wilberforce University Scientists. These collaborations have enabled our faculty and students to conduct research at NASA Lewis during the summer and publish research findings in various journals and scientific proceedings.

Economic evaluation of technical, environmental and institutional barriers on biomass residue collection cost in California

International Nuclear Information System (INIS)

Sethi, P.; Tiangco, V.; Dee, V.; Simons, G.; Lee, Y.; Yomogida, D.

1999-01-01

This paper deals with the economic impacts of the technical, environmental, and institutional barriers in the production, harvesting, and processing of various biomass fuels in California. For each biomass fuel, a base case scenario was developed to characterize the procedures of harvesting, processing and transporting. An economic model was utilized to project the biomass production costs supplied to direct-combustion power plants for each technical, environmental, and institutional barrier. These results will enable the California Energy Commission to identify the most significant barriers to economical biomass energy production (production, harvesting, and processing). (author)

Technical assessment: An independent safety assessment of Department of Energy nuclear reactor facilities

International Nuclear Information System (INIS)

Brodsky, R.S.

1981-02-01

Inherent in the design of DOE reactors under review are many features which provide significant protection against the likelihood of TMI-type accidents. In addition, other features in the design or operating characteristics would tend to limit or reduce the consequences of the accident. Some of these features were discussed earlier in this report. However, some of the events included within the TMI accident sequence contain technical implications for the DOE reactors. These implications were reviewed by this Assessment Team, and the results of this review are reported in this and the following sections of this report. It is also important to reemphasize that as a result of this review, no major TMI-related safety issues have been identified that would indicate that these DOE reactors cannot be operated in a safe manner. Rather, the findings of this report, by nature, generally reemphasize and support ongoing DOE efforts and identify areas for additional improvements

AFRRI (Armed Forces Radiobiology Research Institute) reports, April, May, June 1987. Technical report

Energy Technology Data Exchange (ETDEWEB)

1987-07-01

This document is a collection of reprinted technical reports. Partial contents include: Effect of ionizing radiation on prostaglandins and gastric secretion in rhesus monkeys; Characterization of rat prothymocyte with monoclonal antibodies recognizing rat lymphocyte membrane antigenic determinants; Effects of subdiaphragmatic vagotomy on the acquisition of a radiation-induced condition taste aversion; Ethanol-induced taste aversions; Lack of involvement of acetaldehyde and the area postrema; Dose and time relationships of the radioprotector WR-2721 on locomotor activity in mice; Purification and analysis of rat hematopoietic stem cells by flow cytometry, Plasma histamine and catecholamine levels during hypotension induced by morphine and compound 48/80; Effects of ionizing radiation on hippocampal excitability, Tumor necrosis factor/cachectin is a less-potent inducer of serum amyloid A synthesis than interleukin 1, Protection of mice against fission-neutron irradiation by WR-2721 or WR-151327, Induction of colony-stimulating factor in vivo by recombinant interleukin 1 a and recombinant tumor necrosis factor alpha; 16,16-Dimethyl prostaglandin E2 increases survival in mice following irradiation, Selenium pretreatment enhances the radioprotective effect and reduces the lethal toxicity of WR-2721; Rat phantom depth dose studies in electron, x-ray, gamma-ray, and reactor-radiation fields; Wall attenuation and scatter characteristics of ionization chambers at Armed Forces Radiobiology Research Institute.

Brief history of the reactor physics activities at ICN Pitesti

International Nuclear Information System (INIS)

Dumitrache, I.

2004-01-01

The Institute was established 33 years ago, in April 1971. Several specialists from the Institute for Atomic Physics - Bucharest came at the new research entity and the reactor physics activities had a successful start. One can identify three distinct periods: 1971-1980, the Bucharest years, 1980-1996, solving critical problems years and 1977-present (2004), technical support years. The first period is usually seen as a training one. This is only partially true. Most of the physicists came from University in 1971 and 1972 years. A significant number of them were trained abroad, in France, Germany, Italy, USA, Canada etc., usually under IAEA Vienna fellowships. The work was really pleasant and the progress was exciting. Unfortunately, the main task (to design a thermal reactor and a fast reactor, both for research activities) was, probably, much too difficult from the technical point of view and, in addition, required an unrealistic economic effort. In the Fall of the 1976 year, most of the reactor physicists were removed from Bucharest to Pitesti. One year later, all the remaining specialists were concentrated in Pitesti. The dual core TRIGA reactors were commissioned in the last months of the 1979 year. The CYBER 720 mainframe computer was available in December 1980. Between 1980 and 1992 years, practically all the Romanian activities related to reactor physics were performed in Pitesti, Mioveni compound. The details related to critical problems will be presented in the paper. We mention here four of the problems that have a significant impact even today, namely: -Final dimensioning of the adjuster rods for the Cernavoda NPP, Unit 2. The rods were manufactured in USA and Canada, using the AECL design and the final dimensions have been specified by ICN Pitesti; -Use of the LEU fuel in TRIGA-SSR Reactor, instead of the original HEU fuel; -Design of the irradiation experiments in TRIGA cores, in order to provide the required conditions during the test, according to

Radiation Protection Programme in the Venezuelan Institute of Scientific Research

International Nuclear Information System (INIS)

Lopez Barrios, A.; Diaz, J.; Gonzalez, E.; Lea, D.

1979-01-01

The paper describes the main activities of the radiation protection service in the IVIC concerning: 1. the operation of nuclear reactor RV-1, 2. utilization of radioisotopes, 3. radioactive waste management, 4. personal dosimetry and control of internal contamination, 5. environmental radioactivity and 6. technical assistance to other units of the Institute. (author)

Superphenix: technical and scientific achievements

International Nuclear Information System (INIS)

Guidez, Joel; Prele, Gerard

2016-04-01

In this book, the authors propose a synthesis of technical and scientific achievements related to the design, fabrication and eleven-year operation of Superphenix, the most powerful fast breeder reactor ever built and operated. They had the opportunity to use various and important archives maintained by the different involved institutions, actors and companies, such as the CEA with its MADONA database, AREVA and EDF. They address all the different fields: construction, chemistry, exploitation, handling, small and large components, materials, fuel manufacturing, environmental assessment, thermal hydraulics, the sodium-water reaction, sodium fires, the release of residual power, in-service inspection, and dismantling operations. Moreover, a chapter addresses design studies for Superphenix 2 and for the European Fast Reactor (EFR) which should be the successors of Superphenix

Investigations of the reactivity temperature coefficient of the Dresden Technical University training and research reactor

International Nuclear Information System (INIS)

Adam, E.; Knorr, J.

1982-01-01

Approximate formulas are derived for determining the temperature coefficient of reactivity of the training and research reactor (AKR) of the Dresden Technical University. Values calculated on the basis of these approximations show good agreement with experimentally obtained results, thus confirming the applicability of the formulas to simple systems

Annual activities report of Brazilian Aerospace Technical Center -CTA/IEAv - 1989; Relatorio anual de atividades - CTA/IEAv - 1989

Energy Technology Data Exchange (ETDEWEB)

NONE

1990-12-31

This document reports the research activities on nuclear physics and reactors physics and engineering in the Brazilian Aerospace Technical Center/Advanced Studies Institute, Sao Paulo State, in the year of 1989.

Annual activities report of Brazilian Aerospace Technical Center -CTA/IEAv - 1989; Relatorio anual de atividades - CTA/IEAv - 1989

Energy Technology Data Exchange (ETDEWEB)

NONE

1989-12-31

This document reports the research activities on nuclear physics and reactors physics and engineering in the Brazilian Aerospace Technical Center/Advanced Studies Institute, Sao Paulo State, in the year of 1989.

Technical Basis for Physical Fidelity of NRC Control Room Training Simulators for Advanced Reactors

Energy Technology Data Exchange (ETDEWEB)

Minsk, Brian S.; Branch, Kristi M.; Bates, Edward K.; Mitchell, Mark R.; Gore, Bryan F.; Faris, Drury K.

2009-10-09

The objective of this study is to determine how simulator physical fidelity influences the effectiveness of training the regulatory personnel responsible for examination and oversight of operating personnel and inspection of technical systems at nuclear power reactors. It seeks to contribute to the U.S. Nuclear Regulatory Commission’s (NRC’s) understanding of the physical fidelity requirements of training simulators. The goal of the study is to provide an analytic framework, data, and analyses that inform NRC decisions about the physical fidelity requirements of the simulators it will need to train its staff for assignment at advanced reactors. These staff are expected to come from increasingly diverse educational and experiential backgrounds.

An Assessment of Workplace Skills Acquired by Students of Vocational and Technical Education Institutions

Science.gov (United States)

Bakar, Ab Rahim; Mohamed, Shamsiah; Hamzah, Ramlah

2013-01-01

This study was performed to identify the employability skills of technical students from the Industrial Training Institutes (ITI) and Indigenous People's Trust Council (MARA) Skills Training Institutes (IKM) in Malaysia. The study sample consisted of 850 final year trainees of IKM and ITI. The sample was chosen by a random sampling procedure from…

Scientific and technical production of IPEN - Nuclear and Energetic Research Institute, SP, Brazil. 1997-1999

International Nuclear Information System (INIS)

2001-01-01

This document reports the general activities results of technical and scientific research production of the Institute for Energetic and Nuclear Researches, IPEN, Brazil, during the year of 1997-1999, listing journal articles, scientific events (complete texts, communications, abstracts and panels), thesis and dissertations, books, technical and scientific reports

Light Water Reactor Sustainability Program Reactor Safety Technologies Pathway Technical Program Plan

Energy Technology Data Exchange (ETDEWEB)

Corradini, M. L. [Univ. of Wisconsin, Madison, WI (United States); Peko, D. [US Dept. of Energy, Washington, DC (United States); Farmer, M. [Argonne National Lab. (ANL), Argonne, IL (United States); Rempe, J. [Rempe and Associates LLC, Idaho Falls, ID (United States); Humrickhouse, P. [Idaho National Lab. (INL), Idaho Falls, ID (United States); O' Brien, J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Robb, K. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gauntt, R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Osborn, D. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2016-06-01

In the aftermath of the March 2011 multi-unit accident at the Fukushima Daiichi nuclear power plant (Fukushima), the nuclear community has been reassessing certain safety assumptions about nuclear reactor plant design, operations and emergency actions, particularly with respect to extreme events that might occur and that are beyond each plant’s current design basis. Because of our significant domestic investment in nuclear reactor technology (99 operating reactors in the fleet of commercial LWRs with five under construction), the United States has been a major leader internationally in these activities. The U.S. nuclear industry is voluntarily pursuing a number of additional safety initiatives. The NRC continues to evaluate and, where deemed appropriate, establish new requirements for ensuring adequate protection of public health and safety in the occurrence of low probability events at nuclear plants; (e.g., mitigation strategies for beyond design basis events initiated by external events like seismic or flooding initiators). The DOE has also played a major role in the U.S. response to the Fukushima accident. Initially, DOE worked with the Japanese and the international community to help develop a more complete understanding of the Fukushima accident progression and its consequences, and to respond to various safety concerns emerging from uncertainties about the nature of and the effects from the accident. DOE R&D activities are focused on providing scientific and technical insights, data, analyses methods that ultimately support industry efforts to enhance safety. These activities are expected to further enhance the safety performance of currently operating U.S. nuclear power plants as well as better characterize the safety performance of future U.S. plants. In pursuing this area of R&D, DOE recognizes that the commercial nuclear industry is ultimately responsible for the safe operation of licensed nuclear facilities. As such, industry is considered the primary

Light Water Reactor Sustainability Program: Reactor Safety Technologies Pathway Technical Program Plan

Energy Technology Data Exchange (ETDEWEB)

Corradini, M. L. [Univ. of Wisconsin, Madison, WI (United States)

2015-06-01

In the aftermath of the March 2011 multi-unit accident at the Fukushima Daiichi nuclear power plant (Fukushima), the nuclear community has been reassessing certain safety assumptions about nuclear reactor plant design, operations and emergency actions, particularly with respect to extreme events that might occur and that are beyond each plant’s current design basis. Because of our significant domestic investment in nuclear reactor technology (99 operating reactors in the fleet of commercial LWRs with five under construction), the United States has been a major leader internationally in these activities. The U.S. nuclear industry is voluntarily pursuing a number of additional safety initiatives. The NRC continues to evaluate and, where deemed appropriate, establish new requirements for ensuring adequate protection of public health and safety in the occurrence of low probability events at nuclear plants; (e.g., mitigation strategies for beyond design basis events initiated by external events like seismic or flooding initiators). The DOE has also played a major role in the U.S. response to the Fukushima accident. Initially, DOE worked with the Japanese and the international community to help develop a more complete understanding of the Fukushima accident progression and its consequences, and to respond to various safety concerns emerging from uncertainties about the nature of and the effects from the accident. DOE R&D activities are focused on providing scientific and technical insights, data, analyses methods that ultimately support industry efforts to enhance safety. These activities are expected to further enhance the safety performance of currently operating U.S. nuclear power plants as well as better characterize the safety performance of future U.S. plants. In pursuing this area of R&D, DOE recognizes that the commercial nuclear industry is ultimately responsible for the safe operation of licensed nuclear facilities. As such, industry is considered the primary

Removal of spent fuel from the TVR reactor for reprocessing and proposals for the RA reactor spent fuel handling

International Nuclear Information System (INIS)

Volkov, E.B.; Konev, V.N.; Shvedov, O.V.; Bulkin, S.Yu; Sokolov, A.V.

2002-01-01

The 2,5 MW heavy-water moderated and cooled research reactor TVR was located at the Moscow Institute for Theoretical and Experimental Physics site. In 1990 the final batch of spent nuclear fuel (SNF) from the TVR reactor was transported for reprocessing to Production Association (PA) 'Mayak'. This transportation of the SNF was a part of TVR reactor decommissioning. The special technology and equipment was developed in order to fulfill the preparation of TVR SNF for transportation. The design of the TVR reactor and the fuel elements used are similar to the design and fuel elements of the RA reactor. Two different ways of RA spent fuel elements for transportation to reprocessing plant are considered: in aluminum barrels, and in additional cans. The experience and equipment used for the preparing TVR fuel elements for transportation can help the staff of RA reactor to find the optimal way for these technical operations. (author)

Technical feasibility study of 60 MWe fast reactor concept: RAPID

International Nuclear Information System (INIS)

Kambe, Mitsuru; Ueda, Nobuyuki; Uotani, Masaki

1993-01-01

A study has been performed on the passive safety features and technical feasibility of an inherently safe 60 MWe fast reactor concept RAPID to meet various power requirements in Japan. The system dynamic analyses on the UTOP and ULOF transients revealed that the enhanced reactivity feedback derived from an annular core configuration and the integrated fuel assembly provides a high margin of self-protection. Structural integrity of the integrated fuel assembly has also been confirmed. The following innovative key technologies have been demonstrated; Lithium Injection Modules (LIM) for ultimate shutdown, Lithium Expansion Modulus (LEM) for inherent reactivity feedback and Void Leading Channel (VLC) for the sodium void worth reduction. (author)

Power Reactor Embrittlement Data Base

International Nuclear Information System (INIS)

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

1990-01-01

Regulatory and research evaluations of embrittlement predication models and of pressure vessel integrity can be greatly expedited by the use of a well-designed, computerized data base. The Power Reactor Embrittlement Data Base (PR-EDB) is such a comprehensive collection of data for US commercial nuclear reactors. The Nuclear Regulatory Commission (NRC) has provided financial support, and the Electric Power Research Institute (EPRI) has provided technical assistance in the quality assurance (QA) of the data to establish an industry-wide data base that will be maintained and updated on a long-term basis. Successful applications of the data base to several of NRC's evaluations have received favorable response and support for its continuation. The future direction of the data base has been designed to include the test reactor and other types of data of interest to the regulators and the researchers. 1 ref

Education and training at the Rensselaer Polytechnic Institute reactor critical facility

International Nuclear Information System (INIS)

Harris, D.R.

1989-01-01

The Rensselaer Polytechnic Institute (RPI) Reactor Critical Facility (RCF) has provided hands-on education and training for RPI and other students for almost a quarter of a century. The RCF was built in the 1950s by the American Locomotive Company (ALCO) as a critical facility in which to carry out experiments in support of the Army Package power Reactor (APPR) program. A number of APPRs were built and operated. In the middle 1960s, ALCO went out of business and provided the facility to RPI. Since that time, RPI has operated the RCF primarily in a teaching mode in the nuclear engineering department, although limited amounts of reactor research, activation analysis, and reactivity assays have been carried out as well. Recently, a U.S. Department of Energy (DOE) upgrade program supported refueling the RCF with 4.81 wt% enriched UO 2 high-density pellets clad in stainless steel rods. The use of these SPERT (F1) fuel rods in the RCF provided a cost-effective approach to conversion from high-enrichment bombgrade fuel to low-enrichment fuel. More important, however, is the fact that the new fuel is of current interest for light water power reactors with extended lifetime fuel. Thus, not only are critical reactor experiments being carried out on the fuel but, more importantly, the quality of the education and training has been enhanced

Report on operation, utilization and technical development of research reactors and hot laboratory

International Nuclear Information System (INIS)

1980-03-01

Activities of the Division of Research Reactor Operation in fiscal 1978 are described. The division is responsible for operation and maintenance of JRR-2, JRR-3, JRR-4 and Hot Laboratory. In the above connection, various other works are performed, including technical management of fuel and coolant, radiation control, irradiation technique, etc. In Hot Laboratory, postirradiation examinations of fuels and materials are made, and also development of examination methods. (author)

Report on operation utilization and technical development of research reactors and hot laboratory

International Nuclear Information System (INIS)

1982-03-01

Activities of the Division of Research Reactor Operation in fiscal 1980 are described. The division is responsible for operation and maintenance of JRR-2, JRR-3, JRR-4 and Hot Laboratory. In the above connection, various other works are performed, including technical management of fuel and coolant, radiation control, irradiation technique, etc. In Hot Laboratory, postirradiation examinations of fuels and materials are made, and also development of examination methods. (author)

Report on operation, utilization and technical development of Research Reactors and Hot Laboratory

International Nuclear Information System (INIS)

1984-10-01

Activities of the Division of Research Reactor Operation in fiscal 1981 are described. The division is responsible for operation and maintenance of JRR-2, JRR-3, JRR-4 and Hot Laboratory. In the above connection, various other works are performed, including technical management of fuel and coolant, radiation control, irradiation technique, etc. In Hot Laboratory, postirradiation examinations of fuels and materials are made, and also development of examination methods. (author)

Western Research Institute: Annual technical progress report, October 1987--September 1988

Energy Technology Data Exchange (ETDEWEB)

1988-12-01

This report describes the technical progress made by the Western Research Institute of the University of Wyoming Research Institute of the University of Wyoming Research Corporation on work performed for the period October 1, 1987 through September 30, 1988. This research involves five resource areas: oil shale, tar sand, underground coal gasification, advanced process technology, and advanced fuels research. Under the terms of the cooperative agreement, an annual project plan has been approved by DOE. The work reported herein reflects the implementation of the research in the plan and follows the structure used therein. 49 refs., 32 figs., 87 tabs.

Creep-fatigue damage rules for advanced fast reactor design. Proceedings of a technical committee meeting

International Nuclear Information System (INIS)

1997-03-01

The IAEA, following the recommendations of the International Working Group on Fast Reactors, convened a Technical Committee Meeting on Creep-Fatigue Damage Rules to be used in Fast Reactor Design. The objective of the meeting was to review developments in design rules for creep-fatigue conditions and to identify any areas in which further work would be desirable. The meeting was hosted by AEA Technology, Risley, and held in Manchester, United Kingdom, 11-13 June 1996. It was attended by experts from the European Commission, France, India, Japan, the Republic of Korea, the Russian Federation and the United Kingdom. Refs, figs, tabs

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

International Nuclear Information System (INIS)

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

2000-01-01

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

ENSI's view on technical safety for the long term operation of reactors 1 and 2 in the Beznau nuclear power plant

International Nuclear Information System (INIS)

2010-11-01

The reactors 1 and 2 of the Beznau nuclear power plant (KKB) are operated since about 40 years. For an operation beyond the design period of 40 years the Swiss Federal Nuclear Safety Inspectorate (ENSI) demands the evidence to be brought that the design limits of the safety relevant components will not be reached during the extended operation period. In 2008 the license holder of KKB delivered the requested documentation on material ageing on the basis of deterministic as well as probabilistic safety analyses and concluded that both reactors can be safely operated beyond 40 years. Thanks to continuous additional outfits, both reactors are in good condition from the point of view of technical safety. With a view to the extension of operation beyond 40 years, KKB already applied the necessary measures regarding technics, finances and personnel in order to keep the present technical level. Since 1991 KKB has analysed and checked components that are difficult to replace. From the evidence presented, ENSI concluded that both reactors are able to be operated up to 60 years long, however with two restrictions for reactor 1 because there the material used for the reactor pressure vessel (RPV) suffered more neutron brittleness than in reactor 2. In addition, reactor 1 is much more affected by ageing phenomena than reactor 2, but, according to neutron fluence calculations, the limiting criteria will not be reached even after 60 years of operation. Some corrosion damages were noted at the lower part of the RPV due to water containing boron acid; they are more pronounced in reactor 1 than in reactor 2. Even though the calculations done by KKB are very conservative, they show that also in the long term the operation limiting criteria about the mechanical resistance of the RPV are never reached. ENSI concludes that the safety design of both KKB reactors ensures safe control of the design basis accidents. Both reactors were continuously fitted with new equipment. With the planed

Special issue on ageing management and long term operation of light water reactors

International Nuclear Information System (INIS)

Kashima, Koichi; Kanno, Masanori; Kimura, Itsurou

2008-01-01

Ageing management and long term operation of light water reactors became important in Japan and relevant research on physical and materials ageing has been carried out among organizations of government, academia and industry and establishment of technical standards and guideline based on the results is under way. The Japan Energy Policy Institute (JEPI) issued a special number discussing this theme, which consisted of ten reports of experts describing these activities. Main topics were ageing evaluation of reactor components due to neutron irradiation embrittlement and stress corrosion cracking, regulatory evaluation of deteriorations due to ageing with technical information basis, technology development in the area of inspection/monitoring, ageing evaluation and preventive maintenance/repairs, nuclear power plant life management of electric utilities, and advancement of reactor maintenance and inspection. (T. Tanaka)

Exploring the Deployment Potential of Small Modular Reactors

Science.gov (United States)

Abdulla, Ahmed Y.

This thesis reports the results of several investigations into the viability of an emergent technology. Due to the lack of data in such cases, and the sensitivity surrounding nuclear power, exploring the potential of small modular reactors (SMRs) proved challenging. Moreover, these reactors come in a wide range of sizes and can employ a number of technologies, which made investigating the category as a whole difficult. We started by looking at a subset of SMRs that were the most promising candidates for near to mid-term deployment: integral light water SMRs. We conducted a technically detailed elicitation of expert assessments of their capital costs and construction duration, focusing on five reactor deployment scenarios that involved a large reactor and two light water SMRs. Consistent with the uncertainty introduced by past cost overruns and construction delays, median estimates of the cost of new large plants varied by more than a factor of 2.5. Expert judgments about likely SMR costs displayed an even wider range. There was consensus that an SMR plant's construction duration would be shorter than a large reactor's. Experts identified more affordable unit cost, factory fabrication, and shorter construction schedules as factors that may make light water SMRs economically viable, though these reactors do not constitute a paradigm shift when it comes to nuclear power's safety and security. Using these expert assessments of cost and construction duration, we calculated levelized cost of electricity values for four of the five scenarios. For the large plant, median levelized cost estimates ranged from 56 to 120 per MWh. Median estimates of levelized cost ranged from 77 to 240 per MWh for a 45MWe SMR, and from 65 to 120 per MWh for a 225MWe unit. We concluded that controlling construction duration is important, though not as important a factor in the analysis as capital cost, and, given the price of electricity in some parts of the U.S., it is possible to construct an

Advanced Small Modular Reactor (SMR) Probabilistic Risk Assessment (PRA) Technical Exchange Meeting

Energy Technology Data Exchange (ETDEWEB)

Smith, Curtis [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2013-09-01

During FY13, the INL developed an advanced SMR PRA framework which has been described in the report Small Modular Reactor (SMR) Probabilistic Risk Assessment (PRA) Detailed Technical Framework Specification, INL/EXT-13-28974 (April 2013). In this framework, the various areas are considered: Probabilistic models to provide information specific to advanced SMRs Representation of specific SMR design issues such as having co-located modules and passive safety features Use of modern open-source and readily available analysis methods Internal and external events resulting in impacts to safety All-hazards considerations Methods to support the identification of design vulnerabilities Mechanistic and probabilistic data needs to support modeling and tools In order to describe this framework more fully and obtain feedback on the proposed approaches, the INL hosted a technical exchange meeting during August 2013. This report describes the outcomes of that meeting.

Uncertainty Quantification in the Reliability and Risk Assessment of Generation IV Reactors: Final Scientific/Technical Report

International Nuclear Information System (INIS)

Vierow, Karen; Aldemir, Tunc

2009-01-01

The project entitled, 'Uncertainty Quantification in the Reliability and Risk Assessment of Generation IV Reactors', was conducted as a DOE NERI project collaboration between Texas A and M University and The Ohio State University between March 2006 and June 2009. The overall goal of the proposed project was to develop practical approaches and tools by which dynamic reliability and risk assessment techniques can be used to augment the uncertainty quantification process in probabilistic risk assessment (PRA) methods and PRA applications for Generation IV reactors. This report is the Final Scientific/Technical Report summarizing the project.

Critical technical issues and evaluation and comparison studies for inertial fusion energy reactors

Energy Technology Data Exchange (ETDEWEB)

Abdou, M.A. (Mechanical, Aerospace and Nuclear Engineering Dept., Univ. of California, Los Angeles, CA (United States)); Ying, A.Y. (Mechanical, Aerospace and Nuclear Engineering Dept., Univ. of California, Los Angeles, CA (United States)); Tillack, M.S. (Mechanical, Aerospace and Nuclear Engineering Dept., Univ. of California, Los Angeles, CA (United States)); Ghoniem, N.M. (Mechanical, Aerospace and Nuclear Engineering Dept., Univ. of California, Los Angeles, CA (United States)); Waganer, L.M. (McDonnell Douglas Aerospace, St. Louis, MI (United States)); Driemeyer, D.E. (McDonnell Douglas Aerospace, St. Louis, MI (United States)); Linford, G.J. (TRW Space and Electronics Div., Redondo Beach, CA (United States)); Drake, D.J.

1994-01-01

Two inertial fusion energy (IFE) reactor design concepts developed in the Prometheus studies were evaluated. Objectives were to identify and characterize critical issues and the R and D required to resolve them, and to establish a sound basis for future IFE technical and programmatic decisions. Each critical issue contains several key physics and engineering issues associated with major reactor components and impacts key aspects of feasibility, safety, and economic potential of IFE reactors. Generic critical issues center around: demonstration of moderate gain at low driver energy, feasibility of direct drive targets, feasibility of indirect drive targets for heavy ions, feasibility of indirect drive targets for lasers, cost reduction strategies for heavy ion drivers, demonstration of higher overall laser driver efficiency, tritium self-sufficiency in IFE reactors, cavity clearing at IFE pulse repetition rates, performance/reliability/lifetime of final laser optics, viability of liquid metal film for first wall protection, fabricability/reliability/lifetime of SiC composite structures, validation of radiation shielding requirements, design tools, and nuclear data, reliability and lifetime of laser and heavy ion drivers, demonstration of large-scale non-linear optical laser driver architecture, demonstration of cost effective KrF amplifiers, and demonstration of low cost, high volume target production techniques. Quantitative evaluation and comparison of the two design options have been made with special focus on physics feasibility, engineering feasibility, economics, safety and environment, and research and development (R and D) requirements. Two key conclusions are made based on the overall evaluation analysis. The heavy-ion driven reactors appear to have an overall advantage over laser-driven reactors.

Industrial structure at research reactor suppliers

International Nuclear Information System (INIS)

Roegler, H.-J.; Bogusch, E.; Friebe, T.

2001-01-01

Due to the recent joining of the forces of Framatome S. A. from France and the Nuclear Division of Siemens AG Power Generation (KWU) from Germany to a Joint Venture named Framatome Advanced Nuclear Power S.A.S., the issue of the necessary and of the optimal industrial structure for nuclear projects as a research reactor is, was discussed internally often and intensively. That discussion took place also in the other technical fields such as Services for NPPs but also in the field of interest here, i. e. Research Reactors. In summarizing the statements of this presentation one can about state that: Research Reactors are easier to build than NPPs, but not standardised; Research Reactors need a wide spectrum of skills and experiences; to design and build Research Reactors needs an experienced team especially in terms of management and interfaces; Research Reactors need background from built reference plants more than from operating plants; Research Reactors need knowledge of suitable experienced subsuppliers. Two more essential conclusions as industry involved in constructing and upgrading research reactors are: Research Reactors by far are more than a suitable core that generates a high neutron flux; every institution that designs and builds a Research Reactor lacks quality or causes safety problems, damages the reputation of the entire community

Impressions from a visit by the ASN of the Laue Langevin Institute research reactor in Grenoble

International Nuclear Information System (INIS)

Nifenecker, H.

2011-01-01

After having recalled some specific characteristics of the Laue Langevin Institute research reactor (fuel type, cooling system, power, fuel management, fuel storage pool), the author reports the examination of the emergency procedures and of the reactor maintenance. He describes two exercises which respectively simulated the occurrence of an earthquake and that of a flooding due to a dam breaching

Technical Bases to Consider for Performance and Demonstration Testing of Space Fission Reactors

International Nuclear Information System (INIS)

Hixson, Laurie L.; Houts, Michael G.; Clement, Steven D.

2004-01-01

Performance and demonstration testing are critical to the success of a space fission reactor program. However, the type and extent to which testing of space reactors should be performed has been a point of discussion within the industry for many years. With regard to full power ground nuclear tests, questions such as 'Do the benefits outweigh the risks? Are there equivalent alternatives? Can a test facility be constructed (or modified) in a reasonable amount of time? Will the test article accurately represent the flight system? Are the costs too restrictive?' have been debated for decades. There are obvious benefits of full power ground nuclear testing such as obtaining systems integrated reliability data on a full-scale, complete end-to-end system. But these benefits come at some programmatic risk. In addition, this type of testing does not address safety related issues. This paper will discuss and assess these and other technical considerations essential in deciding which type of performance and demonstration testing to conduct on space fission reactor systems. (authors)

Use of the VR-1 ''Vrabec'' training reactor

International Nuclear Information System (INIS)

Matejka, K.; Kolros, A.; Krops, S.; Polach, S.; Sklenka, L.

1994-01-01

An overview is presented of the extent and ways of using the VR-1 training reactor, which is operated by the Faculty of Nuclear Science and Physical Engineering, Czech Technical University in Prague. A list and the characteristics of 16 problems developed for teaching purposes is given, and the 14 faculties and 2 research institutes participating in the teaching activities are listed. The reactor is used in the education and training of nuclear scientists and engineers. The instrumentation, experimental, handling and operating tools, as well as documentation and texts relating to the reactor are described. The following examples of the teaching activities are included: a guided visit to the operating reactor site, reactor dynamics study and delayed neutron measurement, training course, and the basic criticality experiment. Nuclear safety aspects (hypothetical accidents, quality control and system qualification demonstration, safety culture) are stressed during the education. The reactor department is involved in international cooperation projects. (J.B.). 3 refs

Forty-Sixth Meeting of the Technical Working Group on Fast Reactors (TWG-FR). Working Material

International Nuclear Information System (INIS)

2013-01-01

The objectives of the meeting were to: • Review the current status and the progress since the 45th TWG-FR meeting of FR and ADS technology development activities in IAEA Member States; • Review the activities (past, present and planned) of the IAEA’s project 1.1.5.3, “Support for fast reactor research, technology development and deployment” to ensure that they remain relevant to the needs of Member States; • Provide the experts group with updates to advise the IAEA on FR and ADS activities, including on proposals for relevant studies and reviews; • Serve as a means for exchanging information on national and international FR and ADS programmes; • Review the main achievements and outcomes of the “International Conference on Fast Reactors and Related Fuel Cycle: Safe Technologies and Sustainable Scenarios – FR13”, held on 4 – 7 March 2013 in Paris, France; • Promote the exchange of technical information by proposing topics for, and assisting in the organization of, IAEA Workshops and Technical Meetings for 2014-2015 and further, and • Review the IAEA’s concluded, on-going and planned coordinated research projects (CRPs) in the technical fields relevant to the TWG-FR (FRs and ADS), as well as coordination of the TWG-FR’s activities with other organizations and international initiatives (GIF, INPRO, NEA, Euratom, etc.)

The RA nuclear research reactor at VINCA Institute as an engineering and scientific challenge

International Nuclear Information System (INIS)

Mesarovic, M.

1997-01-01

The RA nuclear research at the Vinca Institute of Nuclear Sciences is the largest nuclear research facility in Yugoslavia and belongs to that generation of research reactors which have had an important contribution to nuclear technology development. As these older reactors were generally not built to specific nuclear standards, new safety systems had to be installed at the RA reactor for a renewal of its operating licence in 1984 and it was shut down, after 25 years of operation. Although all the required and several additional systems were built for the restart of the RA reactor, a disruption of foreign delivery of new control equipment caused its conversion to a 'dormant' facility, and it is still out of operation. Therefore, the future status of the RA reactor presents an engineering and scientific challenge to the engineers and scientists from Yugoslavia and other countries that may be interested to participate. To attract their attention on the subject, principal features of the RA reactor and its present status are described in detail, based on a recent engineering economic and safety evaluation. A comparative review of the world research reactors is also presented.(author)

Technical issues in fusion reactors

International Nuclear Information System (INIS)

Rohatgi, V.K.; Vijayan, T.

1989-01-01

In this paper the issues in fusion reactor technology are examined. Rapid progress in fusion technology research in recent years can be attributed to the advances in various technologies. The commercial generation of fusion power greatly depends on the evolution and improvements in these technologies. With better understanding of plasma physics, fusion reactor designs are becoming more and more realistic and comprehensive. It is now possible to compare various concepts within the framework of established technologies. The technological issues needing better understanding and solutions to problem areas are identified. Various instabilities and energy losses are major problem areas. Extensive developments in reactor-relevant advanced materials, compact and powerful superconducting magnets, high-power systems, and plasma heating drivers need to be undertaken and emphasized

Education program at the Massachusetts Institute of Technology research reactor for pre-college science teachers

International Nuclear Information System (INIS)

Hopkins, G.R.; Fecych, W.; Harling, O.K.

1989-01-01

A Pre-College Science Teacher (PCST) Seminar program has been in place at the Massachusetts Institute of Technology (MIT) Nuclear Reactor Laboratory for 4 yr. The purpose of the PCST program is to educate teachers in nuclear technology and to show teachers, and through them the community, the types of activities performed at research reactors. This paper describes the background, content, and results of the MIT PCST program

Reactor Sharing Program

International Nuclear Information System (INIS)

Tehan, Terry

2002-01-01

Support utilization of the RINSC reactor for student and faculty instructions and research. The Department of Energy award has provided financial assistance during the period 9/29/1995 to 5/31/2001 to support the utilization of the Rhode Island Nuclear Science Center (RINSC) reactor for student and faculty instruction and research by non-reactor owning educational institutions within approximately 300 miles of Narragansett, Rhode Island. Through the reactor sharing program, the RINSC (including the reactor and analytical laboratories) provided reactor services and laboratory space that were not available to the other universities and colleges in the region. As an example of services provided to the users: Counting equipment, laboratory space, pneumatic and in-pool irradiations, demonstrations of sample counting and analysis, reactor tours and lectures. Funding from the Reactor Sharing Program has provided the RINSC to expand student tours and demonstration programs that emphasized our long history of providing these types of services to the universities and colleges in the area. The funding have also helped defray the cost of the technical assistance that the staff has routinely provided to schools, individuals and researchers who have called on the RINSC for resolution of problems relating to nuclear science. The reactor has been featured in a Public Broadcasting System documentary on Pollution in the Arctic and how a University of Rhode Island Professor used Neutron Activation Analysis conducted at the RINSC to discover the sources of the ''Arctic Haze''. The RINSC was also featured by local television on Earth Day for its role in environmental monitoring

KfK, Institute for Reactor Development. Results of research and development activities in 1990

International Nuclear Information System (INIS)

1991-03-01

R and D activities at IRE (Institut fuer Reaktorentwicklung-Institute for Reactor Development) are dedicated to power engineering and handling processes in the framework of the point-of-main-effort projects 'nuclear fusion', 'solid-state and materials research', 'handling', 'nuclear safety' and 'miscellaneous research'. Nuclear fusion contributions deal with special vacuum system design problems, heavy-duty design and materials selection and safety aspects. Specimens from 30 different carbon-based materials were subjected to comparative tests in a plasma spraying plant. In the framework of solid-state and materials research activities a viscoplastic model for mechanical component analysis was investigated for its compatibility with elementary physical laws under complex loads. The handling project was dedicated to the specific requirements of nuclear fusion with regard to JET and NET uses, the development of system solutions for flexible industrial techniques, and to standardization. A study on the control system of a work station for NET remote handling tasks was completed. In the framework of the nuclear safety project one currently investigates the dynamics of fast reactors under failure conditions, the possible propagation of local cooling failures in the reactor core, and core monitoring problems. (orig./GL) [de

Technologies for gas cooled reactor decommissioning, fuel storage and waste disposal. Proceedings of a technical committee meeting

International Nuclear Information System (INIS)

1998-09-01

Gas cooled reactors (GCRs) and other graphite moderated reactors have been important part of the world's nuclear programme for the past four decades. The wide diversity in status of this very wide spectrum of plants from initial design to decommissioning was a major consideration of the International Working group on Gas Cooled Reactors which recommended IAEA to convene a Technical Committee Meeting dealing with GCR decommissioning, including spent fuel storage and radiological waste disposal. This Proceedings includes papers 25 papers presented at the Meeting in three sessions entitled: Status of Plant Decommissioning Programmes; Fuels Storage Status and Programmes; waste Disposal and decontamination Practices. Each paper is described here by a separate abstract

Research on neutron radiography in Research Reactor Institute, Kyoto University and activities related to it

International Nuclear Information System (INIS)

Fujine, Shigenori; Yoneda, Kenji

1994-01-01

The research on neutron radiography in Research Reactor Institute, Kyoto University was begun in 1974 using the E-2 experimental hole which was designed for neutron irradiation. It was reconstructed for the excellent performance as neutron radiography facility by fixing aluminum plugs, a collimator and so on. The research activities thereafter are briefly described. In 1989, the cold neutron facility was installed in the graphite thermal neutron facility, and the experiment on cold neutron radiography became feasible. The reactor in Kyoto University is of the thermal output of 5 MW, and is put to the joint utilization by universities and research institutes in whole Japan. The experimental items carried out so far are enumerated. At present, the main subjects of research are the development of the standard for establishing image evaluation method, the analysis of gas-liquid two-phase flow, the construction of the data base for the literatures and images of neutron radiography, the application of cold neutron radiography, the development of the imaging method using fast neutrons and so on. The thermal neutron radiography and the cold neutron radiography facilities of Kyoto University research reactor are described. The research and activities at Kyoto University research reactor and the investigation of problems are reported. (K.I.) 56 refs

Technical Basis for Water Chemistry Control of IGSCC in Boiling Water Reactors

Science.gov (United States)

Gordon, Barry; Garcia, Susan

Boiling water reactors (BWRs) operate with very high purity water. However, even the utilization of near theoretical conductivity water cannot prevent intergranular stress corrosion cracking (IGSCC) of sensitized stainless steel, wrought nickel alloys and nickel weld metals under oxygenated conditions. IGSCC can be further accelerated by the presence of certain impurities dissolved in the coolant. The goal of this paper is to present the technical basis for controlling various impurities under both oxygenated, i.e., normal water chemistry (NWC) and deoxygenated, i.e., hydrogen water chemistry (HWC) conditions for mitigation of IGSCC. More specifically, the effects of typical BWR ionic impurities (e.g., sulfate, chloride, nitrate, borate, phosphate, etc.) on IGSCC propensities in both NWC and HWC environments will be discussed. The technical basis for zinc addition to the BWR coolant will also provided along with an in-plant example of the most severe water chemistry transient to date.

Report on research and development work 1981 of the Institut fuer Reaktorentwicklung

International Nuclear Information System (INIS)

1982-02-01

The Institute for Reactor Development works chiefly on technical problems as a function of the development and safety evidence of nuclear installations. The tasks that are for a greater part performed together with other internal and external institutions, are both theoretical- and experimental-natured and concentrate mainly on problems out of the reactor safety. The works are in the main concentrated in the 'Fast Breeder Project' (PSB) and in the 'Nuclear Safety Project' (PNS). The PSB-contributions to design and safety of sodium-cooled breeding reactors comprise above all investigations on the thermohydraulic and neutronkinetic behaviour of the core as a result of incidents with very low probability to occur, as well as on the effects of local cooling failures in the reactor core in particular consideration of possible failure propagation mechanisms and of the detectability of such processes. The works in the field of safety of lightwater reactors concern in particular investigations on the dynamic load of component parts of reactor pressure vessels related with the HDR-blowdown-tests and containment structures, as well as investigations on the fuel rod behaviour in the loss of coolant incident. (orig./RW) [de

Nuclear reactor development in China for non-electrical applications

International Nuclear Information System (INIS)

Sun Yuliang; Zhong Daxin; Dong Duo; Xu Yuanhui

1998-01-01

In parallel to its vigorous program of nuclear power generation, China has attached great importance to the development of nuclear reactors for non-electrical applications. The Institute of Nuclear Energy Technology (INET) in Beijing has been developing technologies of the water-cooled heating reactor and the modular high temperature gas-cooled reactor. In 1989, a 5 MW water cooled test reactor was erected. Currently, an industrial demonstration nuclear heating plant is being projected. Feasibility studies are being made of sea-water desalination using the INET developed nuclear heating reactor as heat source. Also, a 10 MW high temperature gas-cooled test reactor is being constructed at INET in the framework of China's national high-tech program. The paper gives an overview of China's energy market situation. With respect to China's technology development of high temperature gas-cooled reactors and water cooled heating reactors, the paper describes some general requirements on the technical development, reviews the national programs and activities, describes briefly the design and safety features of the reactor concepts, discusses aspects of application potentials. (author)

Establishment of the International Power Institute. Final technical report

Energy Technology Data Exchange (ETDEWEB)

Julius E. Coles

2000-08-04

The International Power Institute, in collaboration with American industries, seeks to address technical, political, economic and cultural issues of developing countries in the interest of facilitating profitable transactions in power related infrastructure projects. IPI works with universities, governments and commercial organizations to render project-specific recommendations for private-sector investment considerations. IPI also established the following goals: Facilitate electric power infrastructure transactions between developing countries and the US power industry; Collaborate with developing countries to identify development strategies to achieve energy stability; and Encourage market driven solutions and work collaboratively with other international trade energy, technology and banking organizations.

Forty-Fourth Meeting of the Technical Working Group on Fast Reactors (TWG-FR). Working Material

International Nuclear Information System (INIS)

2011-01-01

The objectives of the meeting were to: - Exchange information on the national programmes on Fast Reactors (FR) and Accelerator Driven Systems (ADS); - Review the progress since the 43rd TWG-FR Annual Meeting, including the status of the actions; - Consider topical technical meeting arrangements for 2012-2013, as well as review FR-related activities included in the IAEA Project&Budget (P&B) biennium 2012-2013; - Review the IAEA’s ongoing information exchange and coordinated research projects in the technical fields relevant to the TWG-FR (FRs and ADS), as well as coordination of the TWG-FR’s activities with other organizations and international initiatives (GIF, INPRO, NEA, ESNII, etc.)

Forty-Second Meeting of the Technical Working Group on Fast Reactors (TWG-FR). Working Material

International Nuclear Information System (INIS)

2009-01-01

The objectives of the meeting were to: - Exchange information on the national programmes on Fast Reactors (FR) and Accelerator Driven Systems (ADS); - Review the progress since the 41st TWG-FR Annual Meeting, including the status of the actions; - Consider topical technical meetings meeting arrangements for 2009, 2010, 2011 and beyond; - Review the IAEA’s ongoing information exchange and coordinated research activities in the technical fields relevant to the TWG-FR (FRs and ADS), as well as coordination of the TWG-FR’s activities with other organizations and international initiatives; - Discuss future joint activities in view of IAEA’s Programme and Budget Cycles beyond 2010-2011

The TITAN Reversed-Field Pinch fusion reactor study

International Nuclear Information System (INIS)

1988-03-01

The TITAN Reversed-Field Pinch (RFP) fusion reactor study is a multi-institutional research effort to determine the technical feasibility and key developmental issues of an RFP fusion reactor, especially at high power density, and to determine the potential economics, operations, safety, and environmental features of high-mass-power-density fusion systems. The TITAN conceptual designs are DT burning, 1000 MWe power reactors based on the RFP confinement concept. The designs are compact, have a high neutron wall loading of 18 MW/m 2 and a mass power density of 700 kWe/tonne. The inherent characteristics of the RFP confinement concept make fusion reactors with such a high mass power density possible. Two different detailed designs have emerged: the TITAN-I lithium-vanadium design, incorporating the integrated-blanket-coil concept; and the TITAN-II aqueous loop-in-pool design with ferritic steel structure. This report contains a collection of 16 papers on the results of the TITAN study which were presented at the International Symposium on Fusion Nuclear Technology. This collection describes the TITAN research effort, and specifically the TITAN-I and TITAN-II designs, summarizing the major results, the key technical issues, and the central conclusions and recommendations. Overall, the basic conclusions are that high-mass power-density fusion reactors appear to be technically feasible even with neutron wall loadings up to 20 MW/m 2 ; that single-piece maintenance of the FPC is possible and advantageous; that the economics of the reactor is enhanced by its compactness; and the safety and environmental features need not to be sacrificed in high-power-density designs. The fact that two design approaches have emerged, and others may also be possible, in some sense indicates the robustness of the general findings

Monte Carlo Analysis of the Accelerator-Driven System at Kyoto University Research Reactor Institute

Directory of Open Access Journals (Sweden)

Wonkyeong Kim

2016-04-01

Full Text Available An accelerator-driven system consists of a subcritical reactor and a controllable external neutron source. The reactor in an accelerator-driven system can sustain fission reactions in a subcritical state using an external neutron source, which is an intrinsic safety feature of the system. The system can provide efficient transmutations of nuclear wastes such as minor actinides and long-lived fission products and generate electricity. Recently at Kyoto University Research Reactor Institute (KURRI; Kyoto, Japan, a series of reactor physics experiments was conducted with the Kyoto University Critical Assembly and a Cockcroft–Walton type accelerator, which generates the external neutron source by deuterium–tritium reactions. In this paper, neutronic analyses of a series of experiments have been re-estimated by using the latest Monte Carlo code and nuclear data libraries. This feasibility study is presented through the comparison of Monte Carlo simulation results with measurements.

Monte Carlo analysis of the accelerator-driven system at Kyoto University Research Reactor Institute

Energy Technology Data Exchange (ETDEWEB)

Kim, Won Kyeong; Lee, Deok Jung [Nuclear Engineering Division, Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of); Lee, Hyun Chul [VHTR Technology Development Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Pyeon, Cheol Ho [Nuclear Engineering Science Division, Kyoto University Research Reactor Institute, Osaka (Japan); Shin, Ho Cheol [Core and Fuel Analysis Group, Korea Hydro and Nuclear Power Central Research Institute, Daejeon (Korea, Republic of)

2016-04-15

An accelerator-driven system consists of a subcritical reactor and a controllable external neutron source. The reactor in an accelerator-driven system can sustain fission reactions in a subcritical state using an external neutron source, which is an intrinsic safety feature of the system. The system can provide efficient transmutations of nuclear wastes such as minor actinides and long-lived fission products and generate electricity. Recently at Kyoto University Research Reactor Institute (KURRI; Kyoto, Japan), a series of reactor physics experiments was conducted with the Kyoto University Critical Assembly and a Cockcroft-Walton type accelerator, which generates the external neutron source by deuterium-tritium reactions. In this paper, neutronic analyses of a series of experiments have been re-estimated by using the latest Monte Carlo code and nuclear data libraries. This feasibility study is presented through the comparison of Monte Carlo simulation results with measurements.

Management of radioactive liquid and solid wastes at the Research Reactor Institute, Kyoto University, (3)

International Nuclear Information System (INIS)

Tsutsui, Tenson; Shimoura, K.; Koyama, A.

1977-11-01

In this report, the management of radioactive liquid and solid wastes at the Research Reactor Institute, Kyoto University during past 6 years, from April in 1971 to March in 1977 are reviewed. (auth.)

Technical report on natural evaporation system for radioactive liquid waste treatment arising from TRIGA research reactors' decontamination and decommissioning activities

International Nuclear Information System (INIS)

Moon, J. S.; Jung, K. J.; Baek, S. T.; Jung, U. S.; Park, S. K.; Jung, K. H.

1999-01-01

This technical report described that radioactive liquid waste treatment for dismantling/decontamination of TRIGA Mark research reactor in Seoul. That is, we try safety treatment of operation radioactive liquid waste during of operating TRIGA Mark research reactor and dismantling radioactive liquid waste during R and D of research reactor hereafter, and by utilizing of new natural evaporation facility with describing design criteria of new natural evaporation facility. Therefore, this technical report described the quantity of present radioactive liquid waste and dismantling radioactive liquid waste hereafter, analysis the status of radial-rays/radioactivity, and also treatment method of this radioactive liquid waste. Also, we derived the method that the safeguard of outskirts environment and the cost down of radioactive liquid waste treatment by minimize of the radioactive liquid waste quantities, through-out design/operation of new natural evaporation facility for treatment of operation radioactive liquid waste and dismantling radioactive liquid waste. (author). 6 refs., 12 tabs., 5 figs

Computer aided training in nuclear power engineering at the Gdansk Technical University

International Nuclear Information System (INIS)

Marecki, J.; Duzinkiewicz, K.; Kosmowski, K.T.

1993-01-01

The Faculty of Electrical Engineering of the Gdansk Technical University has organized post-graduate studies in nuclear power engineering in cooperation with the Institute of Nuclear Research at Swierk since 1973. Post-graduate courses in nuclear power plant construction and design were organized twice. Between 1986 and 1990, prototype software was developed for aiding lectures, self-teaching and knowledge testing in the following fields: 1) dynamics and control of nuclear reactors; 2) simulators of nuclear power plant basic systems (reactor, steam generator, steam turbine, and synchronous generator). (Z.S.) 2 refs

Vinca institute nuclear decommissioning program - Establishment and initialisation

International Nuclear Information System (INIS)

Pesic, M.; Subotic, K.; Ljubenov, V.; Sotic, O.

2003-01-01

Present conditions in The Vinca Institute of Nuclear Sciences related to the nuclear and radiation safety, as result of ambitious nuclear program in the former Yugoslavia and strong economic crisis during the previous decade, have to be improved as soon as possible. RA research reactor, which extended shutdown stage took almost 18 years, spent nuclear fuel from the RA operation in the water pools within the reactor building and inadequate storage facilities for the low and intermediate radioactive wastes at the Vinca site are the main safety problems that have to be solved. To solve the problems mentioned above, a new 'Vinca Nuclear Decommissioning (VIND) Program' is initiated in the Vinca Institute during 2002. The Program team is assembled from about 60 experts from the Institute and relevant organisations. The Program, known also as the G reen Vinca , will be supported, besides the government funding and expected donation from foreign institutions, by experts' help from the IAEA. The necessary equipment will be obtained through the technical assistance from the IAEA. Close co-operation of the team members with experts and relevant companies from nuclear developed countries is expected. (author)

The interplay of institutions, actors and technologies in socio-technical systems - An analysis of transformations in the Australian urban water sector

NARCIS (Netherlands)

Fuenfschilling, Lea; Truffer, Bernhard

2016-01-01

Literature on socio-technical transitions has primarily emphasized the co-determination of institutions and technologies. In this paper, we want to focus on how actors play a mediating role between these two pillars of a socio-technical system. By introducing the theoretical concept of institutional

Advanced reactor development: The LMR integral fast reactor program at Argonne

International Nuclear Information System (INIS)

Till, C.E.

1990-01-01

Reactor technology for the 21st Century must develop with characteristics that can now be seen to be important for the future, quite different from the things when the fundamental materials and design choices for present reactors were made in the 1950s. Argonne National Laboratory, since 1984, has been developing the Integral Fast Reactor (IFR). This paper will describe the way in which this new reactor concept came about; the technical, public acceptance, and environmental issues that are addressed by the IFR; the technical progress that has been made; and our expectations for this program in the near term. 3 figs

Decommissioning of research reactors: Evolution, state of the art, open issues

International Nuclear Information System (INIS)

2006-01-01

Many research reactors throughout the world date from the original nuclear research programmes in Member States. Consequently, dozens of old research reactors are candidates for near term decommissioning in parallel with progressive ageing and technical and economic obsolescence. Many of them are located in countries/institutions that, although familiar with the operation and management of their reactors, do not necessarily have adequate expertise and technologies for planning and implementing state of the art decommissioning projects. It is felt that IAEA reports may contribute to the awareness of technologies and know-how already tested successfully elsewhere. This report addresses a subject area that was dealt with earlier by two IAEA publications, namely, Planning and Management for the Decommissioning of Research Reactors and Other Small Nuclear Facilities (Technical Reports Series No. 351) and Decommissioning Techniques for Research Reactors (Technical Reports Series No. 373). This publication updates those reports in view of the technological progress, experience gained and the progressive ageing of research reactors, many of which have already reached the permanent shutdown stage and should be decommissioned soon. It is intended to contribute to the systematic coverage of the entire range of activities that have been addressed by the IAEA's decommissioning work in past years. The perspective of the report is historical, in that relevant issues are identified as solved, pending, or emerging. Much of the information provided in this report will also be of use for the decommissioning of nuclear power plants and other nuclear facilities. A Technical Committee Meeting on this subject was held in Vienna from 17 to 21 May 2004, at which the participants reviewed a draft report written by consultants from Canada, Germany, Israel, the Russian Federation and the United Kingdom

Case study on the use of PSA methods: Assessment of technical specifications for the reactor protection system instrumentation

International Nuclear Information System (INIS)

1992-10-01

This case study presents a methodology for the probabilistic evaluation of alternative plant technical specifications regarding system surveillance frequencies and out-of-service times. The methodology is applied to the reactor protection systems of a 4 loop BWR-RESAR-3S type nuclear power plant. The effect of the statistical characteristics of the system on the relative comparison of various sets of technical specifications is examined through sensitivity studies and an uncertainty analysis. Refs, figs and tabs

A program to improve educational qualifications of reactor site technical personnel

International Nuclear Information System (INIS)

Christenson, J.M.; Eckart, L.E.

1982-01-01

The authors describe the planning and execution of a program that meets all of the Institute for Nuclear Power Operations (INPO) recommendations and Nuclear Regulatory Commission (NRC) requirements for Shift Technical Advisor (STA) education. They recall and comment these recommendations and requirements, the classification categories of the prospective candidates, indicate the courses proposed by the education program, comment the implementation of the STA program plan

Establishment of regulatory framework for the development reactor licensing

International Nuclear Information System (INIS)

Jo, Jong C.; Yune, Young G.; Kim, Woong S.; Ahn, Sang K.; Kim, In G.; Kim, Hho J.

2003-01-01

With a trend that various types of advanced reactor designs are currently under development worldwide, the Korea Atomic Energy Research Institute has been developing an advanced reactor called ' System-integrated Modular Advanced Reactor (SMART)', which is a small sized integral type pressurized water reactor with a rated thermal power of 330 MW. To demonstrate the safety and the performance of the SMART reactor design, the SMART Research and Development Center has embarked to build a scaled-down pilot plant of SMART, called 'SMART-P' with a rated thermal power of 65 MW. In preparation for the forthcoming applications for both construction permit and operating license of SMART-P in the near future, the Korea Institute of Nuclear Safety is developing a new regulatory framework for licensing review of such a development reactor, which covers establishment of licensing process, identification and resolution of technical and safety issues, development of regulatory evaluation or verification-purpose computer codes and analytical methods, and establishment of design-specific, general design and operating criteria, regulations, and associated regulatory guides. This paper presents the current activities for establishing a regulatory framework for the licensing of a research and development reactor. Discussions are made on the SMART-P development program, the current Korean regulatory framework for reactor licensing, the SMART-P licensing-related issues, and the approach and strategy for developing an effective regulatory framework for the SMART-P licensing

International Conference on Physics and Technology of Reactors and Applications

International Nuclear Information System (INIS)

2007-01-01

The first international conference on physics and technology of reactors and applications (PHYTRA 1) which took place in Marrakech (Morocco) from 14 to 16 March 2007, was designed to bring together scientists, teachers and students from universities, research centres and industry and other institutions to exchange knowledge and to discuss ideas and future issues. The programmes of the PHYTRA 1 conference covers a wide variety topics, the conference was organised in three plenary sessions, ten oral technical sessions and two poster sessions. The plenary sessions covers the following topics : The prospects of nuclear energy, The situation of nuclear sciences and energy in Morocco and Africa, and the new development in reactor physics and reactor design [fr

Catalogue and classification of technical safety standards, rules and regulations for nuclear power reactors and nuclear fuel cycle facilities

International Nuclear Information System (INIS)

Fichtner, N.; Becker, K.; Bashir, M.

1977-01-01

The present report is an up-dated version of the report 'Catalogue and Classification of Technical Safety Rules for Light-water Reactors and Reprocessing Plants' edited under code No EUR 5362e, August 1975. Like the first version of the report, it constitutes a catalogue and classification of standards, rules and regulations on land-based nuclear power reactors and fuel cycle facilities. The reasons for the classification system used are given and discussed

Reactor vessel pressure transient protection for pressurized water reactors

International Nuclear Information System (INIS)

Zech, G.

1978-09-01

During the past few years the NRC has been studying the issue of protection of the reactor pressure vessels at Pressurized Water Reactors (PWRs) from transients when the vessels are at a relatively low temperature. This effort was prompted by concerns related to the safety margins available to vessel damage as a result of such events. Nuclear Reactor Regulation Category A Technical Activity No. A-26 was established to set forth the NRC plan for resolution of the generic aspects of this safety issue. The purpose of the report is to document the completion of this generic technical activity

Technical problems in case of utilizing uranium of medium enrichment for a research reactor

International Nuclear Information System (INIS)

Kanda, Keiji; Shibata, Shun-ichi

1979-01-01

Usually, highly enriched uranium of 90 - 93% is used for research reactors, but the US government proposed the strong policy to use low enriched uranium of the uranium of medium enrichment in unavoidable case from the viewpoint of the resistance to nuclear proliferation in November, 1977. This policy is naturally applied to Japan also. The export of highly enriched uranium will be permitted only when the President approves it after the technical and economical evaluations by the government. The Kyoto University high flux reactor has the features which are not seen in other research reactors, such as medical irradiation, and it is hard to attain the objectives of researches unless HEU is used. The application for the export of HEU was accepted in February, 1978. The nuclear characteristics of the KUHFR when medium or low enriched uranium is used, the criticality experiment in the KUCA using the uranium of medium enrichment, and the burning test on the uranium fuel plates of medium enrichment are described. The research project to lower the degree of enrichment in the fuel for research and test reactors is expected to be continued down to less than 20%. The MEU of 45% enrichment will be actually used in 1983. (Kako, I.)

San Juanico Hybrid System Technical and Institutional Assessment: Preprint

Energy Technology Data Exchange (ETDEWEB)

Corbus, D.; Newcomb, C.; Yewdall, Z.

2004-07-01

San Juanico is a fishing village of approximately 120 homes in the Municipality of Comondu, Baja California. In April, 1999, a hybrid power system was installed in San Juanico to provide 24-hour power, which was not previously available. Before the installation of the hybrid power system, a field study was conducted to characterize the electrical usage and institutional and social framework of San Juanico. One year after the installation of the hybrid power system a''post-electrification'' study was performed to document the changes that had occurred after the installation. In December of 2003, NREL visited the site to conduct a technical assessment of the system.

Reactor materials research as an effective instrument of nuclear reactor perfection

International Nuclear Information System (INIS)

Baryshnikov, M.

2006-01-01

The work is devoted to reactor materiology, as to the practical tool of nuclear reactor development. The work is illustrated with concrete examples from activity experience of the appropriate division of the Russian Research Centre Kurchatov Institute - Institute of Reactor Materials Research and Radiation Nanotechnologies. Besides the description of some modern potentials of the mentioned institute is given. (author)

Forty-Fifth Annual Meeting of the Technical Working Group on Fast Reactors (TWG-FR). Presentations

International Nuclear Information System (INIS)

2012-01-01

The objectives of the meeting were to: • Exchange information on the national programmes on Fast Reactors (FR) and Accelerator Driven Systems (ADS); • Review the progress since the 44 th TWG-FR Annual Meeting, including the status of the actions; • Consider topical technical meeting arrangements for 2012-2013, as well as review FR-related activities included in the IAEA Programme & Budget (P&B) biennium 2012-2013; • Review the IAEA’s concluded, on-going and planned coordinated research projects in the technical fields relevant to the TWG-FR (FRs and ADS), as well as coordination of the TWG-FR’s activities with other organizations and international initiatives (GIF, INPRO, NEA, ESNII, etc.)

Water channel reactor fuels and fuel channels: Design, performance, research and development. Proceedings of a technical committee meeting

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-01-01

The International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT) recommended holding a Technical Committee Meeting on Water Channel Reactor Fuel including into this category fuels and pressure tubes/fuel channels for Atucha-I and II, BWR, CANDU, FUGEN and RBMK reactors. The IWGFPT considered that even if the characteristics of Atucha, CANDUs, BWRs, FUGEN and RBMKs differ considerably, there are also common features. These features include materials aspects, as well as core, fuel assembly and fuel rod design, and some safety issues. There is also some similarity in fuel power history and operating conditions (Atucha-I and II, FUGEN and RBMK). Experts from 11 countries participated at the meeting and presented papers on technology, performance, safety and design, and materials aspects of fuels and pressure tubes/fuel channels for the above types of water channel reactors. Refs, figs, tabs.

Water channel reactor fuels and fuel channels: Design, performance, research and development. Proceedings of a technical committee meeting

International Nuclear Information System (INIS)

1998-01-01

The International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT) recommended holding a Technical Committee Meeting on Water Channel Reactor Fuel including into this category fuels and pressure tubes/fuel channels for Atucha-I and II, BWR, CANDU, FUGEN and RBMK reactors. The IWGFPT considered that even if the characteristics of Atucha, CANDUs, BWRs, FUGEN and RBMKs differ considerably, there are also common features. These features include materials aspects, as well as core, fuel assembly and fuel rod design, and some safety issues. There is also some similarity in fuel power history and operating conditions (Atucha-I and II, FUGEN and RBMK). Experts from 11 countries participated at the meeting and presented papers on technology, performance, safety and design, and materials aspects of fuels and pressure tubes/fuel channels for the above types of water channel reactors

Commercial products and services of research reactors. Proceedings of a technical meeting held in Vienna 28 June - 2 July 2010

International Nuclear Information System (INIS)

2013-07-01

Although the number of operational research reactors is steadily decreasing, more than half of those that remain are greatly underutilized and, in most cases, underfunded. To continue to play a key role in the development of peaceful uses of nuclear technology, the remaining research reactors will need to provide useful products and services to private, national and regional customers, in some cases with adequate revenue generation for reliable, safe and secure facility management and operation. In the light of declining governmental financial support and the need for improved physical security and conversion to low enriched uranium (LEU) fuel, many research reactors have been challenged to generate income to offset increasing operational and maintenance costs. The renewed interest in nuclear power (and therefore in nuclear education and training), the global expansion of diagnostic and therapeutic nuclear medicine, and the extensive use of semiconductors in electronics and in other areas have created new opportunities for research reactors, prominent among them, markets for products and services in regions and countries without such facilities. It is clear that such initiatives towards greater self-reliance will need to address such aspects as market surveys, marketing and business plans, and cost of delivery services. It will also be important to better inform present and future potential end users of research reactor services of the capabilities and products that can be provided. This publication is a compilation of material from an IAEA technical meeting on 'Commercial Products and Services of Research Reactors', held in Vienna, Austria, from 28 June to 2 July 2010. The overall objective of the meeting was to exchange information on good practices and to provide concrete examples, in technical presentations and brainstorming discussions, to promote and facilitate the development of commercial applications of research reactors. The meeting also aimed to enhance

Annual report of Department of Research Reactor and Tandem Accelerator, JFY2014. Operation, Utilization and Technical Development of JRR-3, JRR-4, NSRR, Tandem Accelerator and RI Production Facility

International Nuclear Information System (INIS)

Osa, Akihiko; Imahashi, Masaki; Hirane, Nobuhiko; Motome, Yuiko; Tayama, Hidekazu; Tamura, Itaru; Harada, Yuko; Sakata, Mami; Kadokura, Masakazu; Takita, Chiharu

2017-02-01

The Department of Research Reactor and Tandem Accelerator is in charge of the operation, utilization and technical development of JRR-3 (Japan Research Reactor No.3), JRR-4 (Japan Research Reactor No.4), NSRR (Nuclear Safety Research Reactor), Tandem Accelerator and RI Production Facility. This annual report describes the activities of our department in fiscal year of 2014. We carried out the operation and maintenance, utilization, upgrading of utilization techniques, safety administration, and international cooperation. Also contained are lists of publications, meetings, granted permissions on laws and regulations concerning atomic energy, outcomes in service and technical developments and so on. (author)

Annual report of Department of Research Reactor and Tandem Accelerator, JFY2013. Operation, Utilization and Technical Development of JRR-3, JRR-4, NSRR, Tandem Accelerator and RI Production Facility

International Nuclear Information System (INIS)

Kashima, Yoichi; Murayama, Yoji; Nakamura, Kiyoshi; Uno, Yuki; Hirane, Nobuhiko; Ohuchi, Hitoshi; Ishizaki, Nobuhiro; Matsumura, Taichi; Nagahori, Kazuhisa; Harada, Yuko; Kadokura, Masakazu; Machi, Sumire; Takita, Chiharu

2015-02-01

The Department of Research Reactor and Tandem Accelerator is in charge of the operation, utilization and technical development of JRR-3(Japan Research Reactor No.3), JRR-4(Japan Research Reactor No.4), NSRR(Nuclear Safety Research Reactor), Tandem Accelerator and RI Production Facility. This annual report describes the activities of our department in fiscal year of 2013. We carried out the operation and maintenance, utilization, upgrading of utilization techniques, safety administration and international cooperation. Also contained are lists of publications, meetings, granted permissions on laws and regulations concerning atomic energy, outcomes in service and technical developments and so on. (author)

Results of research and development work in 1982 of the Institut fuer Technische Physik

International Nuclear Information System (INIS)

1983-02-01

The works in the Institute for Technical Physics (ITP) concentrated in 1982 upon the application of high current superconductivity in the fields of energy engineering and fusion. Moreover, the magnet technology for use in other fields of research and technology is advanced. Fundamental investigations on cryogenics and superconductors as well as studies go along with the technical research program. The magnetic fields of stationary and quasi-stationary fusion reactors with magnetic confinement have to be produced by superconductivity. The aim of the magnet developments in the KfK is therefore to make available the technology of superconductivity for large-scale experiments in plasma physics and the later fusion reactors, which is performed in co-operation with industry as part of the European fusion program. (orig./GG) [de

Education and research at the VR-1 Vrabec training reactor facility

International Nuclear Information System (INIS)

Matejka, K.

1993-01-01

The results of 12 years' efforts devoted to the construction of the VR-1 ''Vrabec'' training reactor at the Faculty of Nuclear Science and Physical Engineering, Czech Technical University in Prague and to establishing the training reactor department, as well as the contribution of the training reactor facility to the teaching and scientific activities of the Faculty are presented in a comprehensive manner. The thesis is divided into 2 parts: (i) preconditions, reactor construction and commissioning, and constituting the reactor department, and (ii) basic and comprehensive information concerning the current utilization of the reactor for the benefit of students from various university level institutions. The prospects of scientific activities of the department are also outlined. Attention is paid to selected nuclear safety aspects of the reactor during operation and teaching of students, as well as to its innovated digital control system whose implementation is planned. The results achieved are compared with the initial goals and with similar experience abroad. (P.A.)

Nuclear RB research reactor. Thirty years of anniversary; Istrazhivacki nuklearni reaktor RB. Povodom 30 godina rada

Energy Technology Data Exchange (ETDEWEB)

Pesic, M; Stefanovic, D [Institut za Nuklearne Nauke Boris Kidric, Belgrade (Yugoslavia)

1988-07-01

Nuclear research reactor RB in the Nuclear Engineering Laboratory - NET at the 'Boris Kidric' Institute of Nuclear Sciences in Vinca is the first reactor system built in Yugoslavia in 1958. This year is the thirtieth anniversary of the RB reactor operation, which has survived a series of modifications trying to follow a contemporary nuclear research directions. This report describes its basic technical characteristics and experimental possibilities. Especially, the modifications in the last 25 years are underlined, the experiences gained, and new plans for the future are presented. (author)

Gas-cooled reactor commercialization study. Interim report

International Nuclear Information System (INIS)

1977-01-01

This report of the gas-cooled reactor commercialization study completes the technical and cost evaluation portions of this study contract. A final report in December will update the status of the incentive analyses and the issues of commercialization. This study was designed to bring together potential industry participants (utilities and suppliers) to evaluate the commercial potential of the HTGR-SC and to build channels of communication among the participating organizations at the same time that technical, economic and institutional issues were being evaluated. RAMCO, Inc., in suggesting and using this study approach, believes its application extends to any commercialization problem involving multi-party involvement in high capital, intensive, high risk energy technologies

UCLA program in reactor studies: The ARIES tokamak reactor study

International Nuclear Information System (INIS)

1991-01-01

The ARIES research program is a multi-institutional effort to develop several visions of tokamak reactors with enhanced economic, safety, and environmental features. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Four ARIES visions are currently planned for the ARIES program. The ARIES-1 design is a DT-burning reactor based on ''modest'' extrapolations from the present tokamak physics database and relies on either existing technology or technology for which trends are already in place, often in programs outside fusion. ARIES-2 and ARIES-4 are DT-burning reactors which will employ potential advances in physics. The ARIES-2 and ARIES-4 designs employ the same plasma core but have two distinct fusion power core designs; ARIES-2 utilize the lithium as the coolant and breeder and vanadium alloys as the structural material while ARIES-4 utilizes helium is the coolant, solid tritium breeders, and SiC composite as the structural material. Lastly, the ARIES-3 is a conceptual D- 3 He reactor. During the period Dec. 1, 1990 to Nov. 31, 1991, most of the ARIES activity has been directed toward completing the technical work for the ARIES-3 design and documenting the results and findings. We have also completed the documentation for the ARIES-1 design and presented the results in various meetings and conferences. During the last quarter, we have initiated the scoping phase for ARIES-2 and ARIES-4 designs

Potential for low fracture toughness and lamellar tearing on PWR steam generator and reactor coolant pump supports. Resolution of generic technical activity A-12

International Nuclear Information System (INIS)

Snaider, R.P.; Hodge, J.M.; Levin, H.A.; Zudans, J.J.

1979-10-01

This report summarizes work performed by the Nuclear Regulatory Commission staff and its contractor, Sandia Laboratories, in the resolution of Generic Technical Activity A-12, ''Potential for Low Fracture Toughness and Lamellar Tearing in PWR Steam Generator and Reactor Coolant Pump Supports.'' The report describes the technical issues, the technical studies performed by Sandia describes the technical issues, the technical studies performed by Sandia Laboratories, the NRC staff's technical positions based on these studies, and the staff's plan for implementing its technical positions. It also provides recommendations for further work. The complete technical input from Sandia Laboratories is appended to the report

Technical update on pressure suppression type containments in use in U.S. light water reactor nuclear power plants

International Nuclear Information System (INIS)

1978-07-01

In 1972, Dr. S. H. Hanauer (Technical Advisor to the NRC's Executive Director for Operations) wrote a memorandum that raised several questions on the viability of pressure suppression containment concepts. The concerns raised by Dr. Hanauer have recently become the subject of considerable discussion by several members of the U.S. Congress and public. The report provides a response to these expressed concerns and a status summary for various technical matters that relate to the safety of pressure suppression type containments for light water cooled reactor plants

Overview of research potential of Institute for Nuclear Research

International Nuclear Information System (INIS)

Ciocanescu, Marin

2007-01-01

The main organizations involved in nuclear power production in Romania, under supervision of Presidency, Prime Minister and Parliament are: CNCAN (National Commission for Nuclear Activities Control), Nuclear Agency, Ministry of Economy and Commerce, ANDRAD (Waste Management Agency), SNN (Nuclearelectrica National Society), RAAN (Romanian Authority for Nuclear Activities), ICN (Institute for Nuclear Research - Pitesti), SITON (Center of Design and Engineering for Nuclear Projects- Bucharest); ROMAG-PROD (Heavy Water Plant), CNE-PROD (Cernavoda Nuclear Power Plant - Production Unit), CNE-INVEST (Cernavoda Nuclear Power Plant -Investments Unit), FCN (Nuclear Fuel Factory). The Institute for Nuclear Research, Pitesti INR, Institute for Nuclear Research, Pitesti is endowed with a TRIGA Reactor, Hot Cells, Materials Laboratories, Nuclear Fuel, Nuclear Safety Laboratories, Nuclear Fuel, Nuclear Safety. Waste management. Other research centers and laboratories implied in nuclear activities are: ICIT, National Institute for cryogenics and isotope technologies at Rm Valcea Valcea. with R and D activity devoted to heavy water technologies, IFIN, Institute for nuclear physics and engineering, Bucharest, as well as the educational institutions involved in atomic energy applications and University research, Politechnical University Bucharest, University of Bucharest, University of Pitesti, etc. The INR activity outlined, i.e. the nuclear power research as a scientific and technical support for the Romanian nuclear power programme, mainly dedicated to the existing NPP in the country (CANDU). Focused with priority are: - Nuclear Safety (behavior of plant materials, components, installations during accident conditions and integrity investigations); - Radioactive Waste Management Radioactive; - Radioprotection; Product and services supply for NPP. INR Staff numbers 320 R and D qualified and experienced staff, 240 personnel in devices and prototype workshops and site support

Refinement of nuclear safety education reinforcing technical succession

International Nuclear Information System (INIS)

Yokobori, Seiichi

2008-01-01

In April 2008, Musashi Institute of Technology established another faculty, the Faculty of Nuclear Safety Engineering, to educate students for nuclear engineering to meet the demands of personnel for nuclear business. At this new faculty, students mainly obtain professional knowledge and skills related to nuclear safety issues. This article described refinement of nuclear safety education by reinforcing technical succession topics, such as Rankine cycle, fission, two-phase flow, defense in depth in safety. LOCA/ECCS, seismic effects, reactor maintenance. (T. Tanaka)

Technical meeting on 'Primary coolant pipe rupture event in liquid metal cooled fast reactors'. Working material

International Nuclear Information System (INIS)

2003-01-01

In Liquid Metal cooled Fast Reactors (LMFR) or in accelerator driven sub-critical systems (ADS) with LMFR like sub-critical cores, the primary coolant pipes (PCP) connect the primary coolant pumps to the grid plate. A rupture in one of these pipes could cause significant loss of coolant flow to the core with severe consequences. In loop type reactors, all primary pipelines are provided with double envelopes and inter-space coolant leak monitoring systems that permit leak detection before break. Thus, the PCP rupture event can be placed in the beyond design basis event (BDBE) category. Such an arrangement is difficult to incorporate for pool type reactors, and hence it could be argued that the PCP rupture event needs to be analysed in detail as a design basis event (DBE, category 4 event). The primary coolant pipes are made of ductile austenitic stainless steel material and operate at temperatures of the cold pool and at comparatively low pressures. For such low stressed piping with negligible creep and embrittlement effects, it is of interest to discuss under what design provisions, for pool type reactors, the guillotine rupture of PCP could be placed in the BDBE category. The topical Technical Meeting (TM) on 'Primary Coolant Pipe Rupture Event in Liquid Metal Cooled Reactors' was called to enable the specialists to present the philosophy and analyses applied on this topic in the various Member States for different LMFRs. The scope of the Technical Meeting was to provide a global forum for information exchange on the philosophy applied in the various participating Member States and the analyses performed for different LMFRs with regard to the primary coolant pipe rupture event. More specifically, the objectives of the Technical Meeting were to review the safety philosophy for the PCP rupture event in pool type LMFR, to assess the structural reliability of the PCP and the probability of rupture under different conditions (with/without in-service inspection), to

Characteristics of potential repository wastes: Volume 4, Appendix 4A, Nuclear reactors at educational institutions of the United States; Appendix 4B, Data sheets for nuclear reactors at educational institutions; Appendix 4C, Supplemental data for Fort St. Vrain spent fuel; Appendix 4D, Supplemental data for Peach Bottom 1 spent fuel; Appendix 4E, Supplemental data for Fast Flux Test Facility

International Nuclear Information System (INIS)

1992-07-01

Volume 4 contains the following appendices: nuclear reactors at educational institutions in the United States; data sheets for nuclear reactors at educational institutions in the United States(operational reactors and shut-down reactors); supplemental data for Fort St. Vrain spent fuel; supplemental data for Peach Bottom 1 spent fuel; and supplemental data for Fast Flux Test Facility

History of the research reactor institute of Kyoto University in view of nuclear science information data base (KURRIP)

Energy Technology Data Exchange (ETDEWEB)

Takeuchi, Takayuki; Mizuma, Mitsuo (Kyoto Univ., Kumatori, Osaka (Japan). Research Reactor Inst.); Kimura, Itsuro

1994-02-01

Since the Research Reactor Institute of Kyoto University was established as an inter-university research institute in 1963, a large number of cooperative research projects have been achieved by visiting scientists and its own staff in various research fields, making use of facilities centered around the Kyoto University Reactor, as well as the other experimental facilities. Ten years ago, the construction of the 'KURRIP' data base was initiated to grasp the whole aspect of the research activities at the Institute, in commemoration of its 20th anniversary. At the present time, KURRIP contains the information on 5,910 papers published for 29 years from 1963 to 1991. As this academic year is the 30th anniversary of the Institute, the history of its research activities was reviewed again using this data base. All of the publications were classified by authors's affiliations, kinds of papers, publishers, fields of studies, and research facilities used, and their historical variations are checked and discussed. (author).

Safety philosophy of gas turbine high temperature reactor (GTHTR300)

International Nuclear Information System (INIS)

Shoji Katanishi; Kazuhiko Kunitomi; Shusaku Shiozawa

2002-01-01

Japan Atomic Energy Research Institute (JAERI) has undertaken the study of an original design concept of gas turbine high temperature reactor, the GTHTR300. The general concept of this study is development of a greatly simplified design that leads to substantially reduced technical and cost requirements. Newly proposed design features enable the GTHTR300 to be an efficient and economically competitive reactor in 2010's. Also, the GTHTR300 fully takes advantage of its inherent safety characteristics. The safety philosophy of the GTHTR300 is developed based on the HTTR (High Temperature Engineering Test Reactor) of JAERI which is the first HTGR in Japan. Major features of the newly proposed safety philosophy for the GTHTR300 are described in this article. (authors)

Economic analysis of fast reactor fuel cycle with different modes

International Nuclear Information System (INIS)

Ding Xiaoming

2014-01-01

Because of limitations on the access to technical and economic data and the lack of effective verification, the lack of in-depth study on the economy of fast reactor fuel cycle in China. This paper introduces the analysis and calculation results of the levelized cost of electricity (LCOE) under three different fuel cycle modes including fast reactor fuel cycle carried out by Massachusetts Institute of Technology (MIT). The author used the evaluation method and hypothesis parameters provided by the MIT to carry out the sensitivity analysis for the impact of the overnight cost, the discount rate and changes of uranium price on the LCOE under three fuel cycle modes. Finally, some suggestions are proposed on the study of economy in China's fast reactor fuel cycle. (authors)

Nordic study on reactor waste. Technical part 1 and 2

International Nuclear Information System (INIS)

1981-08-01

An important part of the Nordic studies on system- and safety analysis of the management of low and medium level radioactive waste from nuclear power plants, is the safety analysis of a Reference System. This reference system was established within the study and is described in this Technical Part 1. The reference system covers waste management Schemes that are potential possibilities in either one of the four participating Nordic countries. The reference system is based on: a power reactor system consisting of 6 BWR's of 500 MWe each, operated simultaneously over the same 30 year period, and deep bed granular ion exchange resin wastes from the Reactor Water Clean-Up System (RWCS and powdered ion exchange resin from the Spent Fuel Pool Cleanup System (SFPCS)). Both waste types are supposed to be solidified by mixing with cement and bitumen. Two basic types of containers are considered. Standard 200 liter steel drums and specially made cubicreinforced concrete moulds with a net volume of 1 m 3 . The Nordic study assumes temporary storage of the solidified waste for a maximum of 50 years before the waste is transferred to the disposal site. Transportation of the waste from the storage facilitiy to the disposal site will be by road or sea. Three different disposal facilities are considered: Shallow land burial, near surface concrete bunker, and rock cavern with about 30 m granite cover. (EG)

ATR Technical Specification Upgrade Program

International Nuclear Information System (INIS)

McCracken, R.T.; Durney, J.L.; Freund, G.A.

1990-01-01

The Advanced Test Reactor (ATR) is a 250 MW, uranium-aluminum fueled test reactor which began full power operation in 1969. The initial operation was controlled by an Operating Limits document based on the original Safety Analysis Report. Additional safety bases were later developed to support Technical Specifications which were approved and implemented in 1977. The Technical Specifications which were initially developed with content and format specified in ANSI/ANS--15.1, ''The Development of Technical Specifications for Research Reactors.'' The safety basis documentation and the Technical Specifications have been updated as required to maintain them current with the ATR facility configuration. All revisions have been made with a content, format and style consistent with the original. A major, two-phase program to upgrade the content, format and style is in progress. This paper describes the first phase of this program

Development of an accident consequence assessment code for evaluating site suitability of light- and heavy-water reactors based on the Korean Technical standards

Energy Technology Data Exchange (ETDEWEB)

Hwang, Won Tae; Jeong, Hae Sung; Jeong, Hyo Joon; Kil, A Reum; Kim, Eun Han; Han, Moon Hee [Nuclear Environment Safety Research Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-12-15

Methodologies for a series of radiological consequence assessments show a distinctive difference according to the design principles of the original nuclear suppliers and their technical standards to be imposed. This is due to the uncertainties of the accidental source term, radionuclide behavior in the environment, and subsequent radiological dose. Both types of PWR and PHWR are operated in Korea. However, technical standards for evaluating atmospheric dispersion have been enacted based on the U.S. NRC's positions regardless of the reactor types. For this reason, it might cause a controversy between the licensor and licensee of a nuclear power plant. It was modelled under the framework of the NRC Regulatory Guide 1.145 for light-water reactors, reflecting the features of heavy-water reactors as specified in the Canadian National Standard and the modelling features in MACCS2, such as atmospheric diffusion coefficient, ground deposition, surface roughness, radioactive plume depletion, and exposure from ground deposition. An integrated accident consequence assessment code, ACCESS (Accident Consequence Assessment Code for Evaluating Site Suitability), was developed by taking into account the unique regulatory positions for reactor types under the framework of the current Korean technical standards. Field tracer experiments and hand calculations have been carried out for validation and verification of the models. The modelling approaches of ACCESS and its features are introduced, and its applicative results for a hypothetical accidental scenario are comprehensively discussed. In an applicative study, the predicted results by the light-water reactor assessment model were higher than those by other models in terms of total doses.

A Development of Technical Specification of a Research Reactor with Plate Fuels Cooled by Upward Flow

International Nuclear Information System (INIS)

Park, Sujin; Kim, Jeongeun; Kim, Hyeonil

2016-01-01

The contents of the TS(Technical Specifications) are definitions, safety limits, limiting safety system settings, limiting conditions for operation, surveillance requirements, design features, and administrative controls. TS for Nuclear Power Plants (NPPs) have been developed since many years until now. On the other hands, there are no applicable modernized references of TS for research reactors with many differences from NPPs in purpose and characteristics. Fuel temperature and Departure from Nuclear Boiling Ratio (DNBR) are being used as references from the thermal-hydraulic analysis point of view for determining whether the design of research reactors satisfies acceptance criteria for the nuclear safety or not. Especially for research reactors using plate-type fuels, fuel temperature and critical heat flux, however, are very difficult to measure during the reactor operation. This paper described the outline of main contents of a TS for open-pool research reactor with plate-type fuels using core cooling through passive systems, where acceptance criteria for nuclear safety such as CHF and fuel temperature cannot be directly measured, different from circumstances in NPPs. Thus, three independent variables instead of non-measurable acceptance criteria: fuel temperature and CHF are considered as safety limits, i.e., power, flow, and flow temperature

Annual Report 1991. Institute for systems engineering and informatics

International Nuclear Information System (INIS)

1992-01-01

The report presents the achievements of the Institute for Systems Engineering and Informatics (ISEI) of the Joint Research Centre (JRC) of the Commission of the European Communities (CEC) for 1991. The JRC is a European scientific and technical research centre established by the member states of the CEC. Its four sites in Belgium (Geel), Germany (Karlsruhe), the Netherland (Petten) and Italy (Ispra) house 8 institutes, each with its own focus of expertise. ISEI, based at Ispra, was created in 1990 by the merger of the Institute for Systems Engineering (ISE) and the Centre for Information Technologies and Electronics (CITE). The main areas of activity of the Institute are: - Industrial and Environmental Risk, - Nuclear Safeguards, -Fusion Reactor Systems Integration and Safety, - Solar Energy Systems and Energy Management, - Advanced Computing, - Informatic services

Evolution of the liquid metal reactor: The Integral Fast Reactor (IFR) concept

International Nuclear Information System (INIS)

Till, C.E.; Chang, Y.I.

1989-01-01

The Integral Fast Reactor (IFR) concept has been under development at Argonne National Laboratory since 1984. A key feature of the IFR concept is the metallic fuel. Metallic fuel was the original choice in early liquid metal reactor development. Solid technical accomplishments have been accumulating year after year in all aspects of the IFR development program. But as we make technical progress, the ultimate potential offered by the IFR concept as a next generation advanced reactor becomes clearer and clearer. The IFR concept can meet all three fundamental requirements needed in a next generation reactor. This document discusses these requirements: breeding, safety, and waste management. 5 refs., 4 figs

Current status of operation, utilization and refurbishment of the Dalat nuclear research reactor

International Nuclear Information System (INIS)

Pham Duy Hien.

1993-01-01

The reconstructed nuclear research reactor at Dalat, Vietnam has been put into operation since March 1984. Up to present a cumulative operation time of 13,172 hrs at nominal power (500 kW) has been recorded. Production of radioisotopes for medical uses, element analysis by using activation techniques, as well as fundamental and applied research with filtered neutrons are the main activities of reactor utilizations. The problems facing Dalat Nuclear Research Institute are the ageing of the re-used TRIGA-MARK-II reactor components (especially the corrosion of the reactor tank), as well as the obsolescence of many equipment and components of the reactor control and instrumentation system. Refurbishment works are being in process with the technical and financial supports from the Vietnam government and the IAEA. (author). 7 refs, 2 tabs, 10 figs

Fast Breeder Reactor studies

International Nuclear Information System (INIS)

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

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.

Japan: The institute for the economy of energy recommends a quick re-start of nuclear reactors

International Nuclear Information System (INIS)

Anon.

2014-01-01

The Japanese Institute for the Economy of the Energy (IEEJ) considers that the sooner the nuclear reactors will re-start, the better the Japanese economy and environment will be. The 48 Japanese reactors were stopped after the Fukushima accident and their restart is linked to the implementation of new measures for reinforcing safety. Until now only 2 reactors Sendai 1 and Sendai 2 have been allowed to re-start. The procedure for the safety assessment of the reactors is slower than expected. A study shows that only 7 reactors may be allowed to re-start before march 2015 and a total of 19 units may be operating in march 2016. In this scenario 2% of the electricity will come from nuclear energy in 2014 and 15% in 2015, natural gas imports will still be necessary for the production of electricity and their global cost is estimated to reach 56 billions euros while Japan's rate of energy independence will drop by 4.6%. (A.C.)

Electrical system regulations of the IEA-R1 reactor

International Nuclear Information System (INIS)

Mello, Jose Roberto de; Madi Filho, Tufic

2013-01-01

The IEA-R1 reactor of the Nuclear and Energy Research Institute (IPEN-CNEN/SP), is a research reactor open pool type, designed and built by the U.S. firm Babcock and Wilcox, having, as coolant and moderator, deionized light water and beryllium and graphite, as reflectors. Until about 1988, the reactor safety systems received power from only one source of energy. As an example, it may be cited the control desk that was powered only by the vital electrical system 220V, which, in case the electricity fails, is powered by the generator group: no-break 220V. In the years 1989 and 1990, a reform of the electrical system upgrading to increase the reactor power and, also, to meet the technical standards of the ABNT (Associacao Brasileira de Normas Tecnicas) was carried out. This work has the objective of showing the relationship between the electric power system and the IEA-R1 reactor security. Also, it demonstrates that, should some electrical power interruption occur, during the reactor operation, this occurrence would not start an accident event. (author)

75 FR 62892 - Massachusetts Institute of Technology Research Reactor Environmental Assessment and Finding of No...

Science.gov (United States)

2010-10-13

... NUCLEAR REGULATORY COMMISSION [Docket No. 50-020; NRC-2010-0313] Massachusetts Institute of Technology Research Reactor Environmental Assessment and Finding of No Significant Impact Correction In notice document 2010-24809 beginning on page 61220 in the issue of Monday, October 4, 2010, make the...

Gas-Cooled Thermal Reactor Program. Semiannual technical progress report, October 1, 1982-March 3, 1983

International Nuclear Information System (INIS)

1983-06-01

This report provides descriptions and results of the technical effort during the first half of FY 83 on the Gas-Cooled Thermal Reactor Program. The work on Integration and Management (WBS 01) includes the preparation of the Advanced Systems Concept Evaluation Plan and the Advanced Systems Technology Development Plan in addition to the program management activities. The Market Definition (WBS 03) efforts considered the application of the Modular Reactor System with reforming (MRS-R) to the production of methanol and ammonia and the refining of petroleum. Within the Plant Technology (WBS 13) task there were activities to develop anlytical methods for investigation of Coolant Transport Behavior and to define methods and criteria for High Temperature Structural Engineering design. In addition to the work on the advanced HTGR for process heat users, new activities were initiated in support of the HTGR-SC/C Lead plant Protect (WBS 30 and 31). The Plant Simulation task (WBS 31) was initiated to develop a computer code for simulation of plant operation and for plant transient systems analysis. The efforts on the advanced HTGR systems was performed under the Modular Systems task (WBS 41) to study the potential for multiple small reactors to provide lower costs, improved safety, and higher availability than the large monolithic core reactors

Comparison of best estimate methods for judging design margins of advanced water-cooled reactors. Proceedings of a IAEA technical committee meeting. Working material

International Nuclear Information System (INIS)

1994-01-01

The objectives of the Technical Committee Meeting on Significance of design and Operational Margins for advanced Water Cooled Reactor Systems were: to provide an international forum for presentation and discussion of recent results on best estimate methods for judging design margins of mentioned reactors; to identify and describe the technical features of best estimate methods for predicting margins and to provide input for a status report on a comparison of best estimate methods for assessing margins in different countries and organisations. Participants from thirteen countries presented fifteen papers describing their methods, state of art and experiences. Each of those is presented here by a separate abstract

Comparison of best estimate methods for judging design margins of advanced water-cooled reactors. Proceedings of a IAEA technical committee meeting. Working material

Energy Technology Data Exchange (ETDEWEB)

NONE

1994-12-31

The objectives of the Technical Committee Meeting on Significance of design and Operational Margins for advanced Water Cooled Reactor Systems were: to provide an international forum for presentation and discussion of recent results on best estimate methods for judging design margins of mentioned reactors; to identify and describe the technical features of best estimate methods for predicting margins and to provide input for a status report on a comparison of best estimate methods for assessing margins in different countries and organisations. Participants from thirteen countries presented fifteen papers describing their methods, state of art and experiences. Each of those is presented here by a separate abstract Refs, figs, tabs

Progress of the decommissioning process of Musashi Institute of Technology reactor (4)

International Nuclear Information System (INIS)

Uchiyama, Takafumi; Tanzawa, Tomio; Mitsuhashi, Ishi; Morishima, Kayoko; Matsumoto, Tetsuo

2012-01-01

The research reactor of Tokyo City University Atomic Energy Research Laboratory (Musashi Institute of Technology reactor) is zirconium-moderated water-cooled solid homogeneous type (TRIGA-II type), and its maximum heat output is 100 kW. It got into the first critical state in January 1963, and since then, it has mainly contributed to education and training for upgrading nuclear engineers, radioactivation analysis and reactor physics, and medical researches, as the joint usage research facilities across Japan. Then, after a long-term suspension, the university submitted the file in 2004 to the Ministry of Education, Culture, Sports, Science and Technology on the dismantling for the purpose of facility abolishment. Through the procedure of submitting a decommissioning plan, it was approved. Furthermore, in order to perform the function stop of the disposal facilities of liquid waste, application for change authorization for the decommissioning plan was submitted and approved. Regarding the progress of the decommissioning plan, the dismantling and removal of waste facilities for liquid waste and solid waste was carried out in FY2011 without any trouble. This paper explains this progress and future work plans. (A.O.)

Annual technical report of the prototype fast breeder reactor Monju. 2013

International Nuclear Information System (INIS)

2014-08-01

The prototype fast breeder reactor Monju has accumulated technical achievements in order to establish the fast breeder reactor cycle technology in Japan using the operation and maintenance experience, etc. This annual report summarizes the principal achievements and the data related to the plant management in Monju in fiscal 2013. From the aspect of the management and design evaluation, the following items are summarized: 1) Current status of coping with the order from NRA to alter the safety regulations. 2) Implementation status of reformation of Monju. 3) Current status of the additional geological surveys of crush zones at the Monju site. 4) Development of core seismic assessment method for FBR. Then, from the aspect of the operation and maintenance technology, the following items are summarized: 1) Response to the administrative order to the defect of maintenance management (Part 2). 2) Performance confirmation of the failed fuel detection and location system. 3) Deviation from the limiting conditions for operation in the emergency diesel generator periodic test and so on. Furthermore, from the aspect of the plant management, this report summarizes the data related to the main topics, the history of plant condition, the sodium and water purity management, the radioactive waste management, the equipment inspection and so on. (author)

Institut fuer Heisse Chemie: Results of research and development work in 1984

International Nuclear Information System (INIS)

1985-02-01

The planned facilities for radioactive waste management in the Federal Republic of Germany are the line of orientation along which R and D work proceeds in the Hot Chemistry Institute (IHCh) of KfK, and the Institute thus participates in two major projects, Treatment and Reprocessing of Radioactive Waste, and the Fast Breeder Reactor. The report gives an account of the work accomplished for testing and improving laboratory and technical-scale facilities for use in the PUREX process and for dissolver gas treatment and briefly explains basic chemical and analytical studies. (RB) [de

Some Examples of Accident Analyses for RB Reactor

International Nuclear Information System (INIS)

Pesic, M.

2002-01-01

The RB reactor is heavy water critical assembly operated in the Vinca Institute of Nuclear Sciences, Belgrade, Yugoslavia, since April 1959. The first Safety Analysis Report of the RB critical assembly was prepared in 1961/62. But, the first accidental analysis was done in late 1958 in aim the examine power transient and total equivalent doses received by the staff during the reactivity accident occurred on October 15, 1958. Since 1960, the RB reactor is modified few times. Beside initial natural uranium metal fuel rods, new fuel (TVR-S types) from 2% enriched metal uranium and 80% enriched UO 2 were available since 1962 and 1976, respectively. Also, modifications in control and safety systems of the reactor were done occasionally. Special reactor cores were created using all three types of fuel elements, among them, the coupled fast-thermal ones. Nuclear Safety Committee of the Vinca Institute, an independent regulatory body approved for usage all these modifications of the RB reactor. For those decisions of the Committee, the Preliminary Safety Analysis Reports were prepared that, beside proposed technical modifications and new regulation rules had included analyses of various possible accidents. Special attention is given and new methodology was proposed for thoroughly analyses of design based accidents related to coupled fast-thermal cores, that include reactor central zones filled by fuel elements without moderator. In these accidents, during assumed flooding of the fast zone by moderator, a very high reactivity could be inserted in the system with very high reactivity rate. It was necessary to provide that the safety system of the reactor had fast response to that accident and had enough high (negative) reactivity to shut down the reactor timely. In this paper, a brief overview of some accidents, methodology and computation tools used for the accident analyses at RB reactor are given. (author)

Institutional effectiveness of REDD+ MRV: Countries progress in implementing technical guidelines and good governance requirements

NARCIS (Netherlands)

Ochieng, R.M.; Visseren-Hamakers, Ingrid; Arts, B.; Brockhaus, M.; Herold, M.

2016-01-01

The UNFCCC requires REDD+ countries wishing to receive results-based payments to measure, report and verify (MRV) REDD+ impacts; and outlines technical guidelines and good governance requirements for MRV. This article examines institutional effectiveness of REDD+ MRV by assessing countries’ progress

IAEA Technical Meeting on Status of IAEA Fast Reactor Knowledge Preservation Initiative. Presentations

International Nuclear Information System (INIS)

2013-01-01

The objectives of the technical meeting were to: • exchange information between the Member States/International Organizations on national and international initiatives addressing knowledge preservation and data retrieval/collection in the field of fast neutron systems; • present and discuss the Member States’/International Organizations’ policies and conditions for releasing to the IAEA both publicly available and confidential information on fast neutron systems; • collect data on fast neutron systems provided by participating Member States/International Organizations and encourage participants to contribute in data collection; • provide recommendations for further IAEA initiatives in the field of fast reactor knowledge preservation

Early detection of coolant boiling in research reactors with MTR-type fuel

International Nuclear Information System (INIS)

Kozma, R.; Turkcan, E.; Verhoef, J.P.

1992-10-01

A reactor core monitoring system having the function of early detection of boiling in the coolant channels of research reactors with MTR-type fuel is introduced. The system is based on the on-line analysis of signals of various ex-core and in-core neutron detectors. Early detection of coolant boiling cannot be accomplished by the evaluation of the DC components of these detectors in a number of practically important cases of boiling anomaly. It is shown that the noise component of the available neutron detector signals can be used for the detection of boiling in these cases. Experiments have been carried out at a boiling setup in the research reactor HOR of the Interfaculty Reactor Institute, Technical University of Delft, The Netherlands. (author). 8 refs., 11 figs

Evaluation of the Radiography Program at Caldwell Community College and Technical Institute--Summer, 1982.

Science.gov (United States)

Pipes, V. David

As part of a periodic evaluation of the occupational programs at Caldwell Community College and Technical Institute (CCC&TI), a study of the radiography program was conducted to collect information to facilitate planning, aid in program improvement, and meet accountability demands. The specific objectives of the program evaluation were to…

IAEA Technical Meeting on Status of IAEA Fast Reactor Knowledge Preservation Initiative. Working Material

International Nuclear Information System (INIS)

2013-01-01

In response to needs expressed by Member States and within a broader IAEA-wide effort in nuclear knowledge preservation, the IAEA has been carrying out a dedicated initiative on Fast Reactor Data Knowledge Preservation (FRKP). The main objectives of the FRKP initiative are to: • Halt the on-going loss of information related to Fast Reactors (FR); • Collect, retrieve, preserve and make accessible already existing data and information on FR. These objectives require the implementation of activities supporting digital document archival, exchange, search and retrieval and facilitating, by developing and using suitable standards and IT tools, the knowledge preservation over the next decades. To this purpose the IAEA has developed the Fast Reactor Knowledge Organization System (FRKOS), a web-based application employing IAEA methodology and approach for categorization of FR knowledge domain, which allows creating a comprehensive and well-structured international inventory of fast reactor data and information provided by different Member States. The resulting Web Portal is established and maintained by the IAEA. The IAEA knowledge preservation initiatives and tools in the field of fast neutron systems - which were presented and very well received during the recent IAEA Fast Reactor and Related Fuel Cycles Conference (FR13) - are supposed to be of interest for national nuclear authorities, regulators, scientific and research organizations, commercial companies and all other stakeholders involved in fast reactor activities at national or international level. The objectives of the technical meeting were to: • Exchange information between the member states/international organizations on national and international initiatives addressing knowledge preservation and data retrieval/collection in the field of fast neutron systems; • Present and discuss the member states’/international organizations’ policies and conditions for releasing to the IAEA both publicly

The Integral Fast Reactor

International Nuclear Information System (INIS)

Till, C.E.; Chang, Y.I.; Lineberry, M.J.

1990-01-01

Argonne National Laboratory, since 1984, has been developing the Integral Fast Reactor (IFR). This paper will describe the way in which this new reactor concept came about; the technical, public acceptance, and environmental issues that are addressed by the IFR; the technical progress that has been made; and our expectations for this program in the near term. 5 refs., 3 figs

Technical meeting to 'Review of national programmes on fast reactors and accelerator driven systems (ADS)'. Working material

International Nuclear Information System (INIS)

2002-01-01

The 35th Annual Meeting of the Technical Working Group on Fast Reactors TWG-FR, previously International Working Group on Fast Reactors (IWG-FR, created in 1967), was hosted by the Forschungszentrum Karlsruhe (FZK) and was attended by TWG-FR members and advisers from the following Member States: Brazil, China, France, Germany, India, Japan, the Republic of Kazakhstan, the Republic of Korea, the Russian Federation, and the United States of America. The objectives of the meeting were: to exchange information on the national programmes on Fast Reactors (FR) and Accelerator Driven Systems (ADS); to review the progress since the 34th TWG-FR Annual Meeting, including the status of the actions; to consider meeting arrangements for 2002 and 2003; to review the Agency's co-ordinated research activities in the field of FRs and ADS, as well as co-ordination of the TWG-FR's activities with other organizations

Technical meeting to 'Review of national programmes on fast reactors and accelerator driven systems (ADS)'. Working material

Energy Technology Data Exchange (ETDEWEB)

NONE

2002-07-01

The 35th Annual Meeting of the Technical Working Group on Fast Reactors TWG-FR, previously International Working Group on Fast Reactors (IWG-FR, created in 1967), was hosted by the Forschungszentrum Karlsruhe (FZK) and was attended by TWG-FR members and advisers from the following Member States: Brazil, China, France, Germany, India, Japan, the Republic of Kazakhstan, the Republic of Korea, the Russian Federation, and the United States of America. The objectives of the meeting were: to exchange information on the national programmes on Fast Reactors (FR) and Accelerator Driven Systems (ADS); to review the progress since the 34th TWG-FR Annual Meeting, including the status of the actions; to consider meeting arrangements for 2002 and 2003; to review the Agency's co-ordinated research activities in the field of FRs and ADS, as well as co-ordination of the TWG-FR's activities with other organizations.

The research reactors their contribution to the reactors physics; Les reacteurs de recherche leur apport sur la physique des reacteurs

Energy Technology Data Exchange (ETDEWEB)

Barral, J C [Electricite de France (EDF), 75 - Paris (France); Zaetta, A [CEA/Cadarache, Direction des Reacteurs Nucleaires, DRN, 13 - Saint-Paul-lez-Durance (France); Johner, J [CEA/Cadarache, Dept. de Recherches sur la Fusion Controlee (DRFC), 13 - Saint Paul lez Durance (France); Mathoniere, G [CEA/Saclay, DEN, 91 - Gif sur Yvette (France); and others

2000-07-01

The 19 october 2000, the french society of nuclear energy organized a day on the research reactors. This associated report of the technical session, reactors physics, is presented in two parts. The first part deals with the annual meeting and groups general papers on the pressurized water reactors, the fast neutrons reactors and the fusion reactors industry. The second part presents more technical papers about the research programs, critical models, irradiation reactors (OSIRIS and Jules Horowitz) and computing tools. (A.L.B.)

Fast-neutron dosimetry in the seed-irradiation facility, ASTRA reactor, Seibersdorf

International Nuclear Information System (INIS)

Ahnstroem, G.; Burtscher, A.; Casta, J.

1967-01-01

An important part of the co-ordinated programme on the neutron irradiation of seeds has been the construction of a fast-neutron irradiation facility for swimming-pool reactors. This facility was installed around 70 cm from the core in the ASTRA reactor swimming-pool at the end of December, 1966. Also, for this programme a pair of constant potential ionization chambers have been constructed at the Institute of Biochemistry, Stockholm University. These chambes are of the type described in the technical annex and are the same size as the seed-irradiation vials to be used in the seed-irradiation container (diam. =15 mm, length = 60 mm). Some preliminary dosimetry experiments were undertaken to test the irradiation facility and the ionization chambers, and to investigate the usefulness of the dosimetry instructions in the Technical Annex. The results of these experiments are discussed in this paper. 3 refs, 6 figs, 7 tabs

Systems Sustainability: Implementation of Enhanced Maintenance Programs at the Kurchatov Institute, the All-Russian Research Institute of Experimental physics (VNIIEF) and the All-Russian Scientific Institute for Technical Physics (VNIITF)

International Nuclear Information System (INIS)

Coppinger, M.; Pikula, M.; Randolph, J.D.; Windham, M.

1999-01-01

Implementation of quality maintenance programs is essential to enhancing sustainable continuous operations of United States funded Materials Protection, Control and Accountability (MPC and A) equipment/systems upgrades at various Russian nuclear facilities. An effective maintenance program is expected to provide assurances to both parties for achieving maximum continuous systems operations with minimum down time. To be effective, the program developed must focus on minimum down time for any part of a system. Minimum down time is realized through the implementation of a quality maintenance program that includes preventative maintenance, necessary diagnostic tools, properly trained technical staff, and an in-house inventory of required spare parts for repairing the impacted component of the system. A centralized maintenance management program is logistically essential for the success of this effort because of the large volume of MPC and A equipment/systems installed at those sites. This paper will discuss current programs and conditions at the Russian Research Center-Kurchatov Institute, the All-Russian Scientific Institute for Technical Physics and the All-Russian Research Institute of Experimental Physics and will address those steps necessary to implement an upgraded program at those sites

The International Thermonuclear Experimental Reactor configuration evolution

International Nuclear Information System (INIS)

Lousteau, D.C.; Nelson, B.E.; Lee, V.D.; Thomson, S.L.; Miller, J.M.; Lindquist, W.B.

1989-01-01

The International Thermonuclear Experimental Reactor (ITER) conceptual design activities consist of two phases: a definition phase, completed in September 1988, and a design phase, now in progress. The definition phase was successful in identifying a consistent set of technical characteristics and the broad definition of the required reactor configuration and hardware. Scheduled for completion in November 1990, the design phase is producing a more detailed definition of the required components, a first cost estimate, and a description of site requirements. A major activity in the ITER design phase is the period of joint work conducted at the Max Planck Institute for Plasma Physics, Garching, Federal Republic of Germany, from June through October 1989. An official report of the findings and conclusions of this activity will be submitted to and published by the International Atomic Energy Agency (IAEA). This paper highlights the evolution of the reactor mechanical configuration since the conclusion of the definition phase. 8 figs., 2 tabs

Radiation field studies at the training and research reactor AKR of the Dresden Technical University

International Nuclear Information System (INIS)

Leuschner, A.; Reiss, U.; Pretzsch, G.

1983-01-01

Results of radiation field studies in the experimental channels of the training and research reactor of the Technical University of Dresden are presented. The flux densities of thermal, intermediate and fast neutrons were determined by means of activation detectors., Gamma dose rates have been measured by thermoluminescent dosimeters. The measured results show symmetry with respect to the vertical axis of the reactor and allow to draw conclusions with regard to the efficiency of the individual layers of the shield. They are an essential basis of performing irradiation experiments in the experimental channels. The results of measurements were compared with those of shielding and design calculations. Taking into account the measuring errors and the approximations used in the computational models, no unexpected deviations have been observed. Hence, the measured and calculated results can be assessed to be in good agreement. (author)

Review of accident analyses of RB experimental reactor

International Nuclear Information System (INIS)

Pesic, M.

2003-01-01

The RB reactor is a uranium fuel heavy water moderated critical assembly that has been put and kept in operation by the VINCA Institute of Nuclear Sciences, Belgrade, Serbia and Montenegro, since April 1958. The first complete Safety Analysis Report of the RB reactor was prepared in 1961/62; yet, the first accident analysis had been made in late 1958 with the aim to examine a power transition and the total equivalent doses received by the staff during the reactivity accident that occurred on October 15, 1958. Since 1960, the RB reactor has been modified a few times. Beside the initial natural uranium metal fuel rods, new types of fuel (TVR-S types of Russian origin) consisting of 2% enriched uranium metal and 80% enriched U0 2 , dispersed in aluminum matrix, have been available since 1962 and 1976, respectively. Modifications of the control and safety systems of the reactor were made occasionally. Special reactor cores were designed and constructed using all three types of fuel elements, as well as the coupled fast-thermal ones. The Nuclear Safety Committee of the VINCA Institute, an independent regulator)' body, approved for usage all these modifications of the RB reactor on the basis of the Preliminary Safety' Analysis Reports, which, beside proposed technical modifications and new regulation rules, included safety analyses of various possible accidents. A special attention was given (and a new safety methodology was proposed) to thorough analyses of the design-based accidents related to the coupled fast-thermal cores that included central zones of the reactor filled by the fuel elements without any moderator. In this paper, an overview of some accidents, methodologies and computation tools used for the accident analyses of the RB reactor is given. (author)

Review of accident analyses of RB experimental reactor

Directory of Open Access Journals (Sweden)

Pešić Milan P.

2003-01-01

Full Text Available The RB reactor is a uranium fuel heavy water moderated critical assembly that has been put and kept in operation by the VTNCA Institute of Nuclear Sciences, Belgrade, Serbia and Montenegro, since April 1958. The first complete Safety Analysis Report of the RB reactor was prepared in 1961/62 yet, the first accident analysis had been made in late 1958 with the aim to examine a power transition and the total equivalent doses received by the staff during the reactivity accident that occurred on October 15, 1958. Since 1960, the RB reactor has been modified a few times. Beside the initial natural uranium metal fuel rods, new types of fuel (TVR-S types of Russian origin consisting of 2% enriched uranium metal and 80% enriched UO2 dispersed in aluminum matrix, have been available since 1962 and 1976 respectively. Modifications of the control and safety systems of the reactor were made occasionally. Special reactor cores were designed and constructed using all three types of fuel elements as well as the coupled fast-thermal ones. The Nuclear Safety Committee of the VINĆA Institute, an independent regulatory body, approved for usage all these modifications of the RB reactor on the basis of the Preliminary Safety Analysis Reports, which, beside proposed technical modifications and new regulation rules, included safety analyses of various possible accidents. A special attention was given (and a new safety methodology was proposed to thorough analyses of the design-based accidents related to the coupled fast-thermal cores that included central zones of the reactor filled by the fuel elements without any moderator. In this paper, an overview of some accidents, methodologies and computation tools used for the accident analyses of the RB reactor is given.

Status of innovative small and medium sized reactor designs 2005. Reactors with conventional refuelling schemes

International Nuclear Information System (INIS)

2006-03-01

research and development (R and D) programmes in Member States, and non-technical factors and arrangements that could facilitate successful development and deployment of innovative reactors within the small-to-medium power range. The annexes, intended mostly for designers and technical managers, provide detailed design descriptions of innovative SMRs under development worldwide and are patterned along a newly developed common format, which makes it possible to identify the design philosophy, objectives and approaches, as well as technical features and non-technical factors and arrangements with a potential to provide solutions in the specific areas of concern associated with future nuclear energy systems. Detailed design descriptions in this report were prepared firsthand by the designers and reviewed, updated and approved for publication by the respective vendors, research and design organizations and academic institutions in Member States. The scope of this report is limited to reactors with conventional, proven in operation refuelling schemes and does not include small reactors that could operate without reloading and shuffling of fuel for a long period, from 10 years and more, with no fresh or spent fuel being stored at the site during reactor operation. Because of a large number of inputs, such reactors would require a separate dedicated publication

Progress in design, research and development and testing of safety systems for advanced water cooled reactors. Proceedings of a technical committee meeting

International Nuclear Information System (INIS)

1996-04-01

The meeting covered the following topics: Developments in design of safety-related heat removal components and systems for advanced water cooled reactors; status of test programmes on heat removal components and systems of new designs; range of validity and extrapolation of test results for the qualification of design/licensing computer models and codes for advanced water cooled reactors; future needs and trends in testing of safety systems for advanced water cooled reactors. Tests of heat removal safety systems have been conducted by various groups supporting the design, testing and certification of advanced water cooled reactors. The Technical Committee concluded that the reported test results generally confirm the predicted performance features of the advanced designs. Refs, figs, tabs

Progress in design, research and development and testing of safety systems for advanced water cooled reactors. Proceedings of a technical committee meeting

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-04-01

The meeting covered the following topics: Developments in design of safety-related heat removal components and systems for advanced water cooled reactors; status of test programmes on heat removal components and systems of new designs; range of validity and extrapolation of test results for the qualification of design/licensing computer models and codes for advanced water cooled reactors; future needs and trends in testing of safety systems for advanced water cooled reactors. Tests of heat removal safety systems have been conducted by various groups supporting the design, testing and certification of advanced water cooled reactors. The Technical Committee concluded that the reported test results generally confirm the predicted performance features of the advanced designs. Refs, figs, tabs.

Total Quality Management (TQM) in Self-Financed Technical Institutions: A Quality Function Deployment (QFD) and Force Field Analysis Approach

Science.gov (United States)

Thakkar, Jitesh; Deshmukh, S. G.; Shastree, Anil

2006-01-01

Purpose: To explore the potential for adoption of TQM in self-financed technical institutions in the light of new demands and challenges posed by customers/students and society. Design/methodology/approach: The paper presents use of quality function deployment (QFD) which prioritizes technical requirements and correlates them with various…

Mechanical design and first experimental results of an upgraded technical PERMCAT reactor for tritium recovery in the fuel cycle of a fusion machine

Energy Technology Data Exchange (ETDEWEB)

Welte, S., E-mail: stefan.welte@kit.edu [Karlsruhe Institute of Technology (KIT), Forschungszentrum Karlsruhe, Institute for Technical Physics, Tritium Laboratory Karlsruhe, Hermann v. Helmholtz Platz 1, 76344 Eggenstein Leopoldshafen (Germany); Demange, D.; Wagner, R. [Karlsruhe Institute of Technology (KIT), Forschungszentrum Karlsruhe, Institute for Technical Physics, Tritium Laboratory Karlsruhe, Hermann v. Helmholtz Platz 1, 76344 Eggenstein Leopoldshafen (Germany)

2010-12-15

The PERMCAT process developed for the final clean-up stage of the Tokamak Exhaust Processing systems of the ITER tritium plant combines a catalytic reactor and a Pd/Ag permeator in a single component. A first generation technical PERMCAT has been successfully operated as part of the CAPER experiment at the Tritium Laboratory Karlsruhe for several years. Various alternative PERMCAT mechanical designs were proposed and studied on small-scale prototypes. An upgraded technical PERMCAT reactor was designed, manufactured and commissioned with deuterium. A parallel arrangement of finger-type membranes inserted in a single catalyst bed design was chosen to simplify the geometry and the manufacturing while improving the robustness of the reactor. The component has been designed and manufactured to be fully tritium compatible and also fully compatible with both process and electrical connections of the previous PERMCAT to be replaced. The new PERMCAT mechanical design is more compact and easy to manufacture. This PERMCAT reactor was submitted to functional tests and experiments based on isotopic exchanges between H{sub 2}O and D{sub 2} to measure the processing performances. The first experimental results show decontamination factors versus flow rates better than all previously measured.

Mechanical design and first experimental results of an upgraded technical PERMCAT reactor for tritium recovery in the fuel cycle of a fusion machine

International Nuclear Information System (INIS)

Welte, S.; Demange, D.; Wagner, R.

2010-01-01

The PERMCAT process developed for the final clean-up stage of the Tokamak Exhaust Processing systems of the ITER tritium plant combines a catalytic reactor and a Pd/Ag permeator in a single component. A first generation technical PERMCAT has been successfully operated as part of the CAPER experiment at the Tritium Laboratory Karlsruhe for several years. Various alternative PERMCAT mechanical designs were proposed and studied on small-scale prototypes. An upgraded technical PERMCAT reactor was designed, manufactured and commissioned with deuterium. A parallel arrangement of finger-type membranes inserted in a single catalyst bed design was chosen to simplify the geometry and the manufacturing while improving the robustness of the reactor. The component has been designed and manufactured to be fully tritium compatible and also fully compatible with both process and electrical connections of the previous PERMCAT to be replaced. The new PERMCAT mechanical design is more compact and easy to manufacture. This PERMCAT reactor was submitted to functional tests and experiments based on isotopic exchanges between H 2 O and D 2 to measure the processing performances. The first experimental results show decontamination factors versus flow rates better than all previously measured.

Preliminary decommissioning plan of the reactor IPEN-MB01

International Nuclear Information System (INIS)

Vivas, Ary de Souza

2014-01-01

Around the world, many nuclear plants were built and need to be turned off at a certain time because they are close to their recommended time of use is approximately 50 years. So the IAEA (International Atomic Energy Agency), seeks to guide and recommend a set of guidelines for the conduct of activities of nuclear facilities, with special attention to countries that do not have a framework regulatory Legal that sustain the activities of decommissioning. Brazil, so far, does not have a specific standard to guide the steps of the guidelines regarding decommissioning research reactors. However, in March 2011 a study committee was formed with the main task facing the issues of decommissioning of nuclear installations in Brazil, culminating in Resolution 133 of November 8, 2012, a standard project that treat about the Decommissioning of nucleoelectric plants. O Instituto de Pesquisas Energeticas e Nucleares (IPEN) has two research reactors one being the reactor IPEN/MB-01. The purpose of this master dissertation is to develop a preliminary plan for decommissioning this research reactor, considering the technical documentation of the facility (RAS-Safety Analysis Report), the existing standards of CNEN (National Nuclear Energy Commission), as well as IAEA recommendations. In terms of procedures for decommissioning research reactors, this work was based on what is most modern in experiences, strategies and lessons learned performed and documented in IAEA publications covering techniques and technologies for decommissioning. Considering these technical knowledge and due to the peculiarities of the facility, was selected to immediate dismantling strategy, which corresponds to the start of decommissioning activities once the installation is switched off, dividing it into work sectors. As a resource for monitoring and project management of reactor decommissioning and maintenance of records, we developed a database using Microsoft Access 2007, which contain all the items and

Annual technical report of the prototype fast breeder reactor Monju. 2012

International Nuclear Information System (INIS)

2013-09-01

The prototype fast breeder reactor Monju has accumulated technical achievements in order to establish the fast breeder reactor cycle technology in Japan using the operation and maintenance experience, etc. This annual report summarizes the primary achievements and the data related to the plant management in Monju during fiscal 2012. From the aspect of the design evaluation, the following items are summarized: 1) Comprehensive safety assessments of Monju taking into account the accident at Fukushima Daiichi Nuclear Power Station of Tokyo Electric Power Company, 2) Evaluation of nuclear characteristics based on the data of core confirmation test, 3) Evaluation of hydrogen flux from steam generator tubes, 4) Construction of the advanced safeguards system, 5) Development of a plant dynamics analytical model for the Monju ex-vessel fuel storage system. Then, from the aspect of the maintenance technology, the following items are summarized: 1) Response to the administrative order to the defect of maintenance management, 2) Recovery of in vessel transfer machine dropping accident, 3) Work management by introduction of packaged isolation, 4) Evaluation of result of vibration control of RID sampling blower for secondary sodium loop. Furthermore, from the aspect of the plant management, this report summarizes the data related to the main topics, the history of plant condition, the sodium and water purity management, the radioactive waste management, the equipment inspection and so on. (author)

The research reactor TRIGA Mainz

International Nuclear Information System (INIS)

Hampel, G.; Eberhardt, K.; Trautmann, N.

2006-01-01

The TRIGA Mark II reactor at the Institut fuer Kernchemie became first critical on August 3 rd , 1965. It can be operated in the steady state mode with a maximum power of 100 kWth and in the pulse mode with a peak power of 250 MWth. A survey of the research programmes performed at the TRIGA Mainz is given covering applications in basic research as well as applied science in nuclear chemistry and nuclear physics. Furthermore, the reactor is used for neutron activation analysis and for education and training of scientists, teachers, students and technical personal. Important projects for the future of the TRIGA Mainz are the UCN (ultra cold neutrons) experiment, fast chemical separation, medical applications and the use of the NAA as well as the use of the reactor facility for the training of students in the fields of nuclear chemistry, nuclear physics and radiation protection. Taking into account the past and future operation schedule and the typically low burn-up of TRIGA fuel elements (∝4 g U-235/a), the reactor can be operated for at least the next decade taking into account the fresh fuel elements on stock and without changing spent fuels. (orig.)

Reactor incident status 1981 annual report

International Nuclear Information System (INIS)

Kiser, S.H.

1982-01-01

Reactor Incident followup action is summarized through periodic status reports. This annual report summarizes action taken or anticipated for Reactor Incidents through December 1981. Incidents for which action has been completed, have been deleted from the report. Quarterly addende will update the report by tabulating incidents for each three month period through the coming year. The report consists of a part for the P, K, and C Reactors. Each reactor part is divided into three sections: Further Technical Analysis or Followup Needed; Funding and/or Implementation Needed; and No Further Technical Analysis Anticipated

Next Generation Nuclear Plant Project Technology Development Roadmaps: The Technical Path Forward for 750–800°C Reactor Outlet Temperature

Energy Technology Data Exchange (ETDEWEB)

John Collins

2009-08-01

This document presents the NGNP Critical PASSCs and defines their technical maturation path through Technology Development Roadmaps (TDRMs) and their associated Technology Readiness Levels (TRLs). As the critical PASSCs advance through increasing levels of technical maturity, project risk is reduced and the likelihood of within-budget and on-schedule completion is enhanced. The current supplier-generated TRLs and TDRMs for a 750–800°C reactor outlet temperature (ROT) specific to each supplier are collected in Appendix A.

Armouring facility? Nuclear-weapon and reactor reseach at the Kaiser-Wilhelm Institute for Physics

International Nuclear Information System (INIS)

Hachtmann, R.; Walker, M.

2005-01-01

The Kaiser Wilhelm Institute for Physics is best known as the place where Werner Heisenberg worked on nuclear weapons for Hitler. Although this is essentially true, there is more to the story. At the start of World War II this institute was taken over by the German Army Ordnance to be the central, but not exclusive site for a research project into the economic and military applications of nuclear fission. The Army physicist Kurt Diebner was installed in the institute as its commissarial director. Heisenberg was affiliated with the institute as an advisor at first, and became the director in 1942. Heisenberg and his colleagues, including in particular Karl-Heinz Hoecker, Carl Friedrich von Weizsaecker, and Karl Wirtz, worked on nuclear reactors and isotope separation with the clear knowledge that these were two different paths to atomic bombs [Atombomben]. However, they were clearly ambivalent about what they were doing. New documents recently returned from Russian archives shed new light on this work and the scientists' motivations. (orig.)

Development of a preliminary decommissioning plan of the reactor IPEN/MB-01

International Nuclear Information System (INIS)

Vivas, Ary de Souza; Carneiro, Alvaro Luiz Guimaraes

2013-01-01

Around the world, many nuclear plants were built and need to be turned off at a certain time because they are close to their recommended time of use is approximately 50 years. So the IAEA (International Atomic Energy Agency), seeks to guide and recommend, through publications, guidelines for the conduct of activities both for decommissioning nuclear power plants and for research reactors, with special attention to countries that do not have a framework regulatory Legal that sustain the activities of decommissioning. Brazil, so far, does not have a specific standard to guide the steps of the guidelines regarding decommissioning research reactors, having only a standard applied to decommissioning power plants which was published in November 2012. The Nuclear and Energy Research Institute (IPEN) has two research reactors one being the reactor IPEN/MB-01. The aim of this work is to develop a preliminary plan for decommissioning of nuclear reactor research, considering the technical documentation of the system (RAS-Safety Analysis Report), the existing rules of CNEN (National Nuclear Energy Commission), as well as regulatory instructions and recommendations of the IAEA. The preliminary decommissioning plan consists of the presentation of actions and steps required as well as the strategies to be adopted for the shutdown of the facility under the technical and administrative, seeking the safety, health workers and the general public, minimizing environmental impacts. (author)

Preliminary concept of a zero power nuclear reactor core

International Nuclear Information System (INIS)

Mai, Luiz Antonio; Siqueira, Paulo de Tarso D.

2011-01-01

The purpose of this work is to define a zero power core to study the neutronic behavior of a modern research reactor as the future RMB (Brazilian Nuclear Multipurpose reactor). The platform used was the IPEN/MB-01 nuclear reactor, installed at the Nuclear and Energy Research Institute (IPEN-CNEN/SP). Equilibrium among minimal changes in the current reactor facilities and an arrangement that will be as representative as possible of a future core were taken into account. The active parts of the elements (fuel and control/safety) were determined to be exactly equal the elements of a future reactor. After several technical discussions, a basic configuration for the zero power core was defined. This reactor will validate the neutronic calculations and will allow the execution of countless future experiments aiming a real core. Of all possible alternative configurations for the zero power core representative of a future reactor - named ZPC-MRR (Zero Power Core - Modern Research Reactor), it was concluded, through technical and practical arguments, that the core will have an array of 4 x 5 positions, with 19 fuel elements, identical in its active part to a standard MTR (Material Test Reactor), 4 control/safety elements having a unique flat surface and a central position of irradiation. The specifications of the fuel elements (FEs) are the same as defined to standard MTR in its active part, but the inferior nozzles are differentiated because ZPC-MRR will be a set without heat generation. A study of reactivity was performed using MCNP code, and it was estimated that it will have around 2700 pcm reactivity excess in its 19 FEs configuration (alike the present IPEN/MB-01 reactivity). The effective change in the IPEN/MB-01 reactor will be made only in the control rods drive mechanism. It will be necessary to modify the center of this mechanism. Major modifications in the facility will not be necessary. (author)

Preliminary concept of a zero power nuclear reactor core

Energy Technology Data Exchange (ETDEWEB)

Mai, Luiz Antonio; Siqueira, Paulo de Tarso D., E-mail: lamai@ipen.b, E-mail: ptsiquei@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2011-07-01

The purpose of this work is to define a zero power core to study the neutronic behavior of a modern research reactor as the future RMB (Brazilian Nuclear Multipurpose reactor). The platform used was the IPEN/MB-01 nuclear reactor, installed at the Nuclear and Energy Research Institute (IPEN-CNEN/SP). Equilibrium among minimal changes in the current reactor facilities and an arrangement that will be as representative as possible of a future core were taken into account. The active parts of the elements (fuel and control/safety) were determined to be exactly equal the elements of a future reactor. After several technical discussions, a basic configuration for the zero power core was defined. This reactor will validate the neutronic calculations and will allow the execution of countless future experiments aiming a real core. Of all possible alternative configurations for the zero power core representative of a future reactor - named ZPC-MRR (Zero Power Core - Modern Research Reactor), it was concluded, through technical and practical arguments, that the core will have an array of 4 x 5 positions, with 19 fuel elements, identical in its active part to a standard MTR (Material Test Reactor), 4 control/safety elements having a unique flat surface and a central position of irradiation. The specifications of the fuel elements (FEs) are the same as defined to standard MTR in its active part, but the inferior nozzles are differentiated because ZPC-MRR will be a set without heat generation. A study of reactivity was performed using MCNP code, and it was estimated that it will have around 2700 pcm reactivity excess in its 19 FEs configuration (alike the present IPEN/MB-01 reactivity). The effective change in the IPEN/MB-01 reactor will be made only in the control rods drive mechanism. It will be necessary to modify the center of this mechanism. Major modifications in the facility will not be necessary. (author)

Technical features of advanced boiling water reactor (ABWR)

International Nuclear Information System (INIS)

Horiuchi, Tetsuo; Takashima, Yoshie; Yokomi, Michiro

1986-01-01

As the final stage of the development of ABWRs of 1300 MWe output class carried out since 1984, the design for optimizing the plants has been performed, but it was completed at the end of 1985. Hereafter, the ABWRs will advance to the development of the actual project toward the permission and approval and the construction. By the optimization, the simplification, compacting and the heightening of thermal efficiency of the ABWRs were further promoted. The above promotion was attained by a cylindrical concrete containment vessel constructed in one body with the structure of a reactor building, the optimization of the redundancy of facilities, the optimization of equipment size, the combination of a two-stage reheating, 52 in blade turbine and the recovery of heater drain and so on. The plant characteristics such as the capacity factor, operability and the radiation exposure dose of workers were examined in detail in the process of optimization. As the results, it was shown that the ABWRs have the excellent economical efficiency and the performance characteristics of high level. The technical features of the ABWRs are large capacity and high efficiency plants, the improved core adopting internal pumps and new control rod driving mechanism, rational waste treatment facilities and so on. (Kako, I.)

THE MAN WITHOUT A COUNTRY: LOWELL YEREX, HIS AIRLINE, AND U.S. POLICY CONCERNING INTERNATIONAL COMMERCIAL AVIATION, 1939-44

Directory of Open Access Journals (Sweden)

Erik Benson

2000-01-01

Full Text Available Even as World War Two hung in the balance, the United States looked to the postwar period as a time of opportunity for international airlines. However, U.S. officials disagreed over the best course of action for the future. Would they support one airline or competing carriers? Those favoring competition found a champion in Lowell Yerex, a New Zealand entrepreneur with a thriving Central American airline. However, his nominal British citizenship aroused concerns in an Anglophobic U.S. government, and thus it never accepted him as an option. Still, Yerex played a critical role in the development of U.S. policy and the maturation of modern commercial aviation.

A next-generation reactor concept: The Integral Fast Reactor (IFR)

Energy Technology Data Exchange (ETDEWEB)

Chang, Y.I.

1992-01-01

The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

A next-generation reactor concept: The Integral Fast Reactor (IFR)

Energy Technology Data Exchange (ETDEWEB)

Chang, Y.I.

1992-07-01

The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

A next-generation reactor concept: The Integral Fast Reactor (IFR)

International Nuclear Information System (INIS)

Chang, Y.I.

1992-01-01

The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing

Status of advanced technology and design for water cooled reactors: Heavy water reactors

International Nuclear Information System (INIS)

1989-07-01

In 1987 the IAEA established the International Working Group on Advanced Technologies for Water-Cooled Reactors (IWGATWR). Within the framework of the IWGATWR the IAEA Technical Report on Status of Advanced Technology and Design for Water Cooled Reactors, Part I: Light Water Reactors and Part II: Heavy Water Reactors, has been undertaken to document the major current activities and trends of technological improvement and development for future water reactors. Part I of the report dealing with Light Water Reactors (LWRs) was published in 1988 (IAEA-TECDOC-479). Part II of the report covers Heavy Water Reactors (HWRs) and has now been prepared. This report is based largely upon submissions from Member States. It has been supplemented by material from the presentations at the IAEA Technical Committee and Workshop on Progress in Heavy Water Reactor Design and Technology held in Montreal, Canada, December 6-9, 1988. It is hoped that this part of the report, containing the status of advanced heavy water reactor technology up to 1988 and ongoing development programmes will aid in disseminating information to Member States and in stimulating international cooperation. Refs, figs and tabs

Report on the Survey of the Design Review of New Reactor Applications. Volume 4: Reactor Coolant and Associated Systems

International Nuclear Information System (INIS)

Downey, Steven; Monninger, John; Nevalainen, Janne; Joyer, Philippe; Koley, Jaharlal; Kawamura, Tomonori; Chung, Yeon-Ki; Haluska, Ladislav; Persic, Andreja; Reierson, Craig; Monninger, John; Choi, Young-Joon; )

2017-01-01

At the tenth meeting of the Committee on Nuclear Regulatory Activities (CNRA) Working Group on the Regulation of New Reactors (WGRNR) in March 2013, the Working Group agreed to present the responses to the Second Phase, or Design Phase, of the licensing process survey as a multi-volume text. As such, each report will focus on one of the eleven general technical categories covered in the survey. The general technical categories were selected to conform to the topics covered in the International Atomic Energy Agency (IAEA) Safety Guide GS-G-4.1. This report provides a discussion of the survey responses related to the Reactor Coolant and Associated Systems category. The Reactor Coolant and Associated Systems category includes the following technical topics: overpressure protection, reactor coolant pressure boundary, reactor vessel, and design of the reactor coolant system. For each technical topic, the member countries described the information provided by the applicant, the scope and level of detail of the technical review, the technical basis for granting regulatory authorisation, the skill sets required and the level of effort needed to perform the review. Based on a comparison of the information provided by the member countries in response to the survey, the following observations were made: - Although the description of the information provided by the applicant differs in scope and level of detail among the member countries that provided responses, there are similarities in the information that is required. - All of the technical topics covered in the survey are reviewed in some manner by all of the regulatory authorities that provided responses. - It is common to consider operating experience and lessons learnt from the current fleet during the review process. - The most commonly and consistently identified technical expertise needed to perform design reviews related to this category are mechanical engineering and materials engineering. The complete survey

A review of research activities at the Research Reactor Institute of Kyoto University in view of research publication information

International Nuclear Information System (INIS)

Takeuchi, Takayuki; Mizuma, Mitsuo; Kimura, Itsuro.

1995-01-01

A database of research publication was constructed for the purpose of grasping all of the research activities at the Research Reactor Institute, Kyoto University. The database named KURRIP collects all of the research publications of the Institute by not only its own staff but also visiting scientists. The publications are in the form of original papers, review papers, papers in proceedings, short notes and letters, synopses over 3 pages presented orally at scientific meeting, books and doctoral theses. At present, the KURRIP database contains the information on 6,210 items which have been published for 30 years since the Institute was established as an interuniversity research institute for joint use of a research reactor and other related large facilities in 1963. By utilizing the KURRIP database, the analyses have been done: (1) affiliation of the authors, (2) kind of publications, (3) classification of publishers, (4) research fields, and (5) experimental facilities. The KURRIP database is now stored in the Data Processing Center of Kyoto University and can be utilized through a computer center at one of the main national universities in Japan. (author)

TRIGA Research Reactor Conversion to LEU and Modernization of Safety Related Systems

Energy Technology Data Exchange (ETDEWEB)

Sanda, R. M. [Institute for Nuclear Research Piteşti (SCN-Piteşti), Piteşti (Romania)

2014-08-15

The USA and IAEA proposed an international programme to reduce the enrichment of uranium in research reactors by converting nuclear fuel containing HEU into fuel containing 20% enriched uranium. The Government of Romania joined the programme and actively supported political, scientific, technical and economic actions that led to the conversion of the active area of the 14 MW TRIGA reactor at the Institute for Nuclear Research in Piteşti in May 2006. This confirmed the continuity of the Romanian Government’s non-proliferation policy and their active support of international cooperation. Conversion of the Piteşti research reactor was made possible by completion of milestones in the Research Agreement for Reactor Conversion, a contract signed with the US Department of Energy and Argonne National Laboratory. This agreement provided scientific and technical support and the possibility of delivery of all HEU TRIGA fuel to the United States. Additionally, about 65% of the fresh LEU fuel needed to start the conversion was delivered in the period 1992–1994. Furthermore, conversion was promoted through IAEA Technical Cooperation project ROM/4/024 project funded primarily by the United States that supported technical and scientific efforts and the delivery of the remaining required LEU nuclear fuel to complete the conversion. Nuclear fuel to complete the conversion was made by the French company CERCA with a tripartite contract among the IAEA, CERCA and Romania. The contract was funded by the US Department of Energy with a voluntary contribution by the Romanian Government. The contract stipulated manufacturing and delivery of LEU fuel by CERCA with compliance measures for quality, delivery schedule and safety requirements set by IAEA standards and Romanian legislation. The project was supported by the ongoing technical cooperation, safeguards, legal and procurement assistance of the IAEA, in particular its Department of Nuclear Safety. For Romanian research, the

Safety evaluation report related to the renewal of the operating license for the Worcester Polytechnic Institute open-pool training reactor, Docket No. 50-134

International Nuclear Information System (INIS)

1982-12-01

This Safety Evaluation Report for the application filed by the Worcester Polytechnic Institute (WPI) for a renewal of Operating License R-61 to continue to operate the WPI 10-kW open-pool training reactor has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is owned and operated by the Worcester Polytechnic Institute and is located on the WPI campus in Worcester, Worcester County, Massachusetts. The staff concludes that the reactor facility can continue to be operated by WPI without endangering the health and safety of the public

Progress report on research and development activities of the Institut fuer Neutronenphysik und Reaktortechnik in 1980

International Nuclear Information System (INIS)

1981-02-01

The Institut fuer Neutronenphysik und Reaktortechnik carries out research in nuclear enginering especially in the context of the development of sodium-cooled fast breeder reactors. The institute does theoretical and experimental work in equal shares. Most of the work is done in cooperation with other institutes and industrial groups. In the framework of the Projekt Schneller Brueter , the institute investigates the layout and safety problems, from a reactor physics point of view of large fast breeder cores. Problems concerning the effects of LWR accidents on population and environment are investigated in the Projekt Nukleare Sicherheit . To the Projekt Wiederaufarbeitung , the institute is contributing work on fuel cycle physics and the development of a suitable instrumentation. In the field of nuclear safeguards, the implementation of monitoring systems is being investigated. Research outside projects is being done on technical and physical problems of fusion by inertial confinement and application of nuclear power on the transportation sector with hydrogen as an energy source. The main results obtained in 1980 in the various projects are presented. (orig./RW) [de

Nuclear calculation for employing medium enrichment in reactors of Japan Atomic Energy Research Institute

International Nuclear Information System (INIS)

Miyasaka, Yasuhiko

1979-01-01

The fuel used for the research reactors of Japan Atomic Energy Research Institute (JAERI) is presently highly enriched uranium of 93%. However, the U.S. government (the supplier of fuel) is claiming to utilize low or medium enriched uranium from the viewpoint of resistivity to nuclear proliferation, and the availability of highly enriched uranium is becoming hard owing to the required procedure. This report is described on the results of nuclear calculation which is the basis of fuel design in the countermeasures to the reduction of enrichment. The basic conception in the reduction of enrichment is three-fold: to lower the latent potential of nuclear proliferation as far as possible, to hold the present reactor performance as far as possible, and to limit the reduction in the range which is not accompanied by the modification of reactor core construction and cooling system. This time, the increase of the density and thickness of fuel plates and the effect of enrichment change to 45% on reactivity and neutron flux were investigated. The fuel of UAl sub(x) - Al system was assumed, which was produced by powder metallurgical method. The results of investigations on JRR-2 and JMTR reactors revealed that 45% enriched fuel does not affect the performances much. However, deterioration of the performances is not neglegible if further reduction is needed. In future, the influence of the burn-up effect of fuel on the life of reactor cores must be investigated. (Wakatsuki, Y.)

A technical solution to a difficult problem - An update

International Nuclear Information System (INIS)

Asselstine, J.K.

1985-01-01

A goal of the Reduced Enrichment Research and Test Reactor Program (RERTR) established by the U.S. Department of Energy is to develop the technical means needed to utilize LEU instead of HEU fuels in research and test reactors and to do so without operational penalties and without affecting safety characteristics. This goal has been pursued now for seven years in cooperation with many of you who are participants in this conference. The program has made significant progress in the development, demonstration and application of new LEU research reactor fuels. According to recent projections, the percentage of HEU fuel in international commerce which is technically replaceable with LEU fuels has been growing steadily as research continues. There appears to be no technical barrier to eventual replacement of HEU fuel with LEU fuel in research and test reactors. While there are many politically complex and technically difficult issues involved in nonproliferation, I conclude that the use of highly enriched weapons grade uranium as a fuel in civilian research reactors stands out as one problem which has a straight forward technical solution. The technical solution is the effort you in the RERTR program are embarked on. Once that solution is in hand, the operators of research reactors are urged to take political and legal initiatives to make the conversion happen. By effecting this conversion they will help minimize the trade in highly enriched uranium and make the world a little safer

A technical solution to a difficult problem - An update

Energy Technology Data Exchange (ETDEWEB)

Asselstine, J K

1985-07-01

A goal of the Reduced Enrichment Research and Test Reactor Program (RERTR) established by the U.S. Department of Energy is to develop the technical means needed to utilize LEU instead of HEU fuels in research and test reactors and to do so without operational penalties and without affecting safety characteristics. This goal has been pursued now for seven years in cooperation with many of you who are participants in this conference. The program has made significant progress in the development, demonstration and application of new LEU research reactor fuels. According to recent projections, the percentage of HEU fuel in international commerce which is technically replaceable with LEU fuels has been growing steadily as research continues. There appears to be no technical barrier to eventual replacement of HEU fuel with LEU fuel in research and test reactors. While there are many politically complex and technically difficult issues involved in nonproliferation, I conclude that the use of highly enriched weapons grade uranium as a fuel in civilian research reactors stands out as one problem which has a straight forward technical solution. The technical solution is the effort you in the RERTR program are embarked on. Once that solution is in hand, the operators of research reactors are urged to take political and legal initiatives to make the conversion happen. By effecting this conversion they will help minimize the trade in highly enriched uranium and make the world a little safer.

Decommissioning of fast reactors after sodium draining

International Nuclear Information System (INIS)

2009-11-01

Acknowledging the importance of passing on knowledge and experience, as well mentoring the next generation of scientists and engineers, and in response to expressed needs by Member States, the IAEA has undertaken concrete steps towards the implementation of a fast reactor data retrieval and knowledge preservation initiative. Decommissioning of fast reactors and other sodium bearing facilities is a domain in which considerable experience has been accumulated. Within the framework and drawing on the wide expertise of the Technical Working Group on Fast Reactors (TWG-FR), the IAEA has initiated activities aiming at preserving the feedback (lessons learned) from this experience and condensing those to technical recommendations on fast reactor design features that would ease their decommissioning. Following a recommendation by the TWG-FR, the IAEA had convened a topical Technical Meeting (TM) on 'Operational and Decommissioning Experience with Fast Reactors', hosted by CEA, Centre d'Etudes de Cadarache, France, from 11 to 15 March 2002 (IAEA-TECDOC- 1405). The participants in that TM exchanged detailed technical information on fast reactor operation and decommissioning experience with various sodium cooled fast reactors, and, in particular, reviewed the status of the various decommissioning programmes. The TM concluded that the decommissioning of fast reactors to reach safe enclosure presented no major difficulties, and that this had been accomplished mainly through judicious adaptation of processes and procedures implemented during the reactor operation phase, and the development of safe sodium waste treatment processes. However, the TM also concluded that, on the path to achieving total dismantling, challenges remain with regard to the decommissioning of components after sodium draining, and suggested that a follow-on TM be convened, that would provide a forum for in-depth scientific and technical exchange on this topic. This publication constitutes the Proceedings of

Chernobyl reactor accident

International Nuclear Information System (INIS)

Malinauskas, A.P.; Buchanan, J.R.; Lorenz, R.A.; Yamashita, T.

1986-01-01

On April 26, 1986, an explosion occurred at the newest of four operating nuclear reactors at the Chernobyl site in the USSR. The accident initiated an international technical exchange of almost unprecedented magnitude; this exchange was climaxed with a meeting at the International Atomic Energy Agency in Vienna during the week of August 25, 1986. The meeting was attended by more than 540 official representatives from 51 countries and 20 international organizations. Information gleaned from that technical exchange is presented in this report. A description of the Chernobyl reactor, which differs significantly from commercial US reactors, is presented, the accident scenario advanced by the Russian delegation is discussed, and observations that have been made concerning fission product release are described

Nuclear reactors; graphical symbols

International Nuclear Information System (INIS)

1987-11-01

This standard contains graphical symbols that reveal the type of nuclear reactor and is used to design graphical and technical presentations. Distinguishing features for nuclear reactors are laid down in graphical symbols. (orig.) [de

AREVA Technical Days (ATD) session 3: operations of the reactors and services division, technical and economic aspects; AREVA Technical Days (ATD) session 3: les activites du pole reacteurs et services, aspects techniques et economiques

Energy Technology Data Exchange (ETDEWEB)

NONE

2003-07-01

These technical days organized by the Areva Group aims to explain the group activities in a technological and economic point of view, to provide an outlook of worldwide energy trends and challenges and to present each of their businesses in a synthetic manner. This third session deals with the reactors technologies basics, the EPR and SWR 1000 issues and outlook, the nuclear systems of the future, the business opportunities and business models. (A.L.B.)

Current status of feasibility studies on commercialized fuel cycle system for Fast Breeder Reactor

International Nuclear Information System (INIS)

Ojima, Hisao; Nagaoki, Yoshihiro

2000-01-01

A 'Feasibility Studies on Commercialized Fast Breeder Reactor Cycle System' is underway at the Japan Nuclear Cycle Development Institute (JNC). The study will select the promising concepts with their R and D tasks in order to commercialize the fast breeder reactor (FBR) cycle system. The feasibility studies (F/S) have to present surveyed and screened various relevant technologies, and defined the design requirement of the commercialized fuel cycle system for FBR. The promising technical options are being evaluated and conceptual designs are being examined. At the end of JFY2000, several candidate concepts of the commercialized FBR cycle system will be proposed. (author)

Advanced fuel designs for existing and future generations of reactors: driving factors from technical and economic points of view

International Nuclear Information System (INIS)

Hesketh, Kevin

2003-01-01

This paper reviews the current state of advanced fuel research and development and considers advanced fuel development work in the context of the technical and economic drivers. The scope encompasses evolutionary development for existing light water reactors (LWRs), radical developments for LWRs, most of which are focused on more efficient plutonium consumption and on longer term developments in relation to thermal and fast reactor fuels. The review concludes that there is a gap between near-term research and development to support utilities and the long-term work that focuses on goals such as improved plutonium utilisation, waste reduction, improved proliferation resistance and strategic independence

Guidebook to nuclear reactors

International Nuclear Information System (INIS)

Nero, A.V. Jr.

1976-05-01

A general introduction to reactor physics and theory is followed by descriptions of commercial nuclear reactor types. Future directions for nuclear power are also discussed. The technical level of the material is suitable for laymen

Research Reactor Application for Materials under High Neutron Fluence. Proceedings of an IAEA Technical Meeting (TM-34779)

International Nuclear Information System (INIS)

2011-05-01

Research reactors (RRs) have played, and continue to play, a key role in the development of the peaceful uses of nuclear energy and technology. The role of the IAEA is to assist Member States in the effective utilization of these technologies in various domains of research such as fundamental and applied science, industry, human health care and environmental studies, as well as nuclear energy applications. In particular, material testing reactors (MTRs), serve as unique tools in scientific and technological development and they have quite a wide variety of applications. Today, a large range of different RR designs exist when compared with power reactors and they also have different operating modes, producing high neutron fluxes, which may be steady or pulsed. Recently, an urgent need has arisen for the development of new advanced materials, for example in the nuclear industry, where RRs offer capacities for irradiation programmes. Besides the scientific and research activities and commercial applications, RRs are also used extensively for educational training activities for scientists and engineers. This report is a compilation of outputs of an IAEA Technical Meeting (TM-34779) held on Research Reactor Application for Materials under High Neutron Fluence. The overall objective of the meeting was to review typical applications of small and medium size RRs, such as material characterization and testing, neutron physics and beam research, neutron radiography and imaging as well as isotope production and other types of non-nuclear applications. Several issues were discussed during the meeting, in particular: (1) recent development of irradiation facilities, specific irradiation programmes and their implementation; (2) effective and optimal RR operation regimes for irradiation purposes; (3) sharing of best practices and existing technical knowledge; and (4) fostering of advanced or innovative technologies, e.g. information exchange and effective collaboration. This

Control rod guide tube wear in operating reactors; operating experience report. Technical report December 1977-December 1979

International Nuclear Information System (INIS)

Riggs, R.

1980-04-01

Evidence of control rod guide tube wear has been observed in operating pressurized water reactors. The cause of this wear is identified as flow-induced vibration of the control rods. This report describes the measures being taken by both the industry and the NRC to deal with this matter. The staff also presents its technical positions and requirements to support continued operation of the plants as of December 1979 pending completion of this generic effort

Japan Atomic Energy Research Institute, Reactor Engineering Division annual report

International Nuclear Information System (INIS)

1981-09-01

Research activities in the Division of Reactor Engineering in fiscal 1980 are described. The work of the Division is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and fusion reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and fusion reactor technology, and activities of the Committee on Reactor Physics. (author)

International cooperation in accident analysis of RBMK reactors

International Nuclear Information System (INIS)

Kaliatka, A.; Isag

2005-01-01

Chouha Michel (Institute for Radiological Protection and Nuclear Safety), D'Auria Francesco (Institute of Pisa), Kaliatka Algirdas (Lithuanian Energy Institute), Uspuras Eugenijus (Lithuanian Energy Institute). The safety of nuclear power plants is a primary concern of the European Union (EU) and its Member States. In the early 1990s, the European Union decided to take a prominent role in international efforts to help the New Independent States (NIS) and countries of central Europe to ensure the safety of their nuclear reactors. The Commission's approach to nuclear safety in central and Eastern Europe and the NIS is based on two main objectives, which are fully in line with the policy of the international community as decided by the G7 in 1992: 1) In the short term, to improve operational safety; to make near term technical improvements to plants based on safety assessments and to enhance regulatory regimes; 2) In the longer term, to examine the scope for replacing less safe plants by the development of alternative energy sources and more efficient use of energy and to examine the potential for upgrading plants of more recent design. In this paper the safety concerns, related to RBMK type reactors (Russian acronym for 'Channelized Large Power Reactor) are discussed. These reactors were not exported and were built exclusively in the territory of the former Soviet Union. There are presently plants at Saint Petersburg (Sosnovy Bor), Kursk, Chernobyl and Smolensk. A total of 17 such reactors have been built and 12 are currently in operation. Two international projects: TACIS project 'Development of a code system for severe accident analysis in RBMK reactors' and PHARE projects 'Support to VATESI for Important Tasks Relevant to the Licensing Activities of Ignalina Nuclear Power Plant' are presented. The aim of the TACIS project is to help the Russian Authorities to build such capabilities, for their RBMK nuclear power plants (NPPs). The drawing of the Tacis nuclear

An assessment of space reactor technology needs and recommendations for development

International Nuclear Information System (INIS)

Marshall, A.C.; Wiley, R.L.

1996-01-01

In order to provide a strategy for space reactor technology development, the Defense Nuclear Agency (DNA) has authorized a brief review of potential national needs that may be addressed by space reactor systems. A systematic approach was used to explore needs at several levels that are increasingly specific. sm-bullet Level 0 emdash General Trends and Issues sm-bullet Level 1 emdash Generic Space Capabilities to Address Trends sm-bullet Level 2 emdash Requirements to Support Capabilities sm-bullet Level 3 emdash System Types Capable of Meeting Requirements sm-bullet Level 4 emdash Generic Reactor System Types sm-bullet Level 5 emdash Specific Baseline Systems Using these findings, a strategy was developed to support important space reactor technologies within a limited budget. A preliminary evaluation identified key technical issues and provide a prioritized set of candidate research projects. The evaluation of issues and the recommended research projects are presented in a companion paper. copyright 1996 American Institute of Physics

77 FR 68155 - The Armed Forces Radiobiology Research Institute TRIGA Reactor: Facility Operating License No. R-84

Science.gov (United States)

2012-11-15

... Research Institute TRIGA Reactor: Facility Operating License No. R-84 AGENCY: Nuclear Regulatory Commission... considering an application for the renewal of Facility Operating License No. R-84 (Application), which... the renewal of Facility Operating License No. R-84, which currently authorizes the licensee to operate...

The research reactor TRIGA Mainz. A neutron source for versatile applications in research and education

International Nuclear Information System (INIS)

Eberhardt, K.; Kronenberg, A.

2000-01-01

Currently, four research reactors with a thermal power ranging from 0.1 to 23 MW th are in operation in Germany and one new reactor (20 MW th ) is under construction. The TRIGA Mark II reactor at the Institut fuer Kernchemie became first critical on August 3, 1965. It can be operated in the steady state mode with a maximum power of 100 kW th and in the pulse mode with a peak power of 250 MW th . A survey of the research programmes carried out at the TRIGA Mainz is given covering a wide range of applications in basic and applied science in nuclear chemistry, nuclear- and particle physics. Furthermore, the reactor is used for neutron activation analysis and for education and training of students and technical personal. (orig.) [de

Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR) are compared

International Nuclear Information System (INIS)

Greneche, D.

2014-01-01

This article compares the 2 types of light water reactors that are used to produce electricity: the Pressurized Water Reactor (PWR) and the Boiling Water Reactor (BWR). Historically the BWR concept was developed after the PWR concept. Today 80% of light water reactors operating in the world are of PWR-type. This comparison is comprehensive and detailed. First the main technical features are reviewed and compared: reactor architecture, core and fuel design, reactivity control, reactor vessel, cooling systems and reactor containment. Secondly, various aspects concerning reactor operations like reactor control, fuel management, maintenance, inspections, radiation protection, waste generation and reactor reliability are presented and compared for both reactors. As for the issue of safety, it is highlighted that the accidental situations are too different for the 2 reactors to be compared. The main features of reactor safety are explained for both reactors

Balancing the technical, administrative, and institutional forces in defense waste management

International Nuclear Information System (INIS)

Hindman, T.B.

1988-01-01

Defense radioactive waste results from the Department of Energy's (DOE) national defense and nuclear weapons production activities. In 1983, the President submitted to Congress the Defense Waste Management Plan (DWMP) for defense high-level and transuranic wastes. The Plan proposed a workable approach for the final disposition of these wastes. The Department is still following the path laid out in this Plan. The proper management of this waste requires that technical, administrative, and institutional forces which are often neither well understood nor well documented be properly balanced. This paper clarifies the role these three forces play in the Defense waste management programs and provides examples of their impacts on specific programs

Forschungszentrum Rossendorf, Institute of Safety Research. Annual report 1997

Energy Technology Data Exchange (ETDEWEB)

Weiss, F P; Rindelhardt, U [eds.

1998-10-01

The research work of the institute aims at the assessment and increase of the safety of technical facilities. Subject of the investigation are equally nuclear plants and installations of process industries. To analyse thermo-fluiddynamic phenomena of accident scenarios physical models and computer codes are developed as well for multi-phase and multi-component flows as for the time and space dependent power release (neutron kinetics in light water reactors, reaction kinetics of exothermic chemical reactions). Emphasis is put on the description of spatial flows and the transient evolution of flow patterns. (orig.)

Development of physical conceptions of fast reactors

International Nuclear Information System (INIS)

Khomyakov, Yu.S.; Matveev, V.I.; Moiseev, A.V.

2013-01-01

• Russian experience in developing fast reactors has proved clearly scientific justification of conceptual physical principles and their technical feasibility. • However, the potential of fast reactors caused by their physical features has not been fully realized. • In order to assure the real possibility of transition to the nuclear power with fast reactors by about 2030 it is necessary to consistently update fast reactor designs for solving the following key problems: - increasing of self-protection level of reactor core; - improvement of technical and economical characteristics; - solution of the problems related to the fuel supply of nuclear power and assimilation of closed nuclear fuel cycle; - disposal of long lived radioactive waste and transmutation of minor actinides. • Russian program (2010-2020) on the development of basic concepts of the new generation reactors implies successive solution of the above problems. • New technical decisions will be demonstrated by development and assimilation of the new reactors: - BN-800 – development of the fuel cycle infrastructure and mastering of the new types of fuel; - BN-1200 reactor – demonstration economical efficiency of fast reactor and new level of safety; - BREST development and demonstration new heavy liquid metal coolant technology and alternative design concept

Franco-German cooperation for the physical protection of the EPR reactor

International Nuclear Information System (INIS)

Jalouneix, J.; Hagemann, A.

2001-01-01

This article presents the proceeding that has been followed in the EPR (European pressurized water reactor) project concerning physical protection against malevolent actions and robbery of nuclear materials. Before the different options of the nuclear island were definitely set, a task group had been constituted to examine if these options could hamper the setting of physical protection measures that are required by the legislation of the 2 countries. Another group composed of experts from IPSN/GRS (Institut de Protection et de Surete Nucleaire / Gesellschaft fur Anlagen und Reaktorsicherheit) had the task to define common requirements concerning the physical protection of reactors in Germany and in France. In this framework the EPR project team has prepared a technical document reviewing the different dispositions that have been retained to assure the physical protection of the reactor. (A.C.)

Japan Atomic Energy Research Institute, Reactor Engineering Division annual report

International Nuclear Information System (INIS)

1979-09-01

Research activities in the Division of Reactor Engineering in fiscal 1978 are described. Works of the Division are development of multi-purpose Very High Temperature Gas Cooled Reactor, fusion reactor engineering, and development of Liquid Metal Fast Breeder Reactor for Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology, and Committees on Reactor Physics and in Decommissioning of Nuclear Facilities. (author)

Assessment of the Capability of Molten Salt Reactors as a Next Generation High Temperature Reactors

International Nuclear Information System (INIS)

Elsheikh, B.M.

2017-01-01

Molten Salt Reactor according to Aircraft Reactor Experiment (ARE) and the Molten Salt Reactor Experiment (MSRE) programs, was designed to be the first full-scale, commercial nuclear power plant utilizing molten salt liquid fuels that can be used for producing electricity, and producing fissile fuels (breeding)burning actinides. The high temperature in the primary cycle enables the realization of efficient thermal conversion cycles with net thermal efficiencies reach in some of the designs of nuclear reactors greater than 45%. Molten salts and liquid salt because of their low vapor pressure are excellent candidates for meeting most of the requirements of these high temperature reactors. There is renewed interest in MSRs because of changing goals and new technologies in the use of high-temperature reactors. Molten Salt Reactors for high temperature create substantial technical challenges to have high effectiveness intermediate heat transfer loop components. This paper will discuss and investigate the capability and compatibility of molten salt reactors, toward next generation high temperature energy system and its technical challenges

Pool type liquid metal fast breeder reactors

International Nuclear Information System (INIS)

Guthrie, B.M.

1978-08-01

Various technical aspects of the liquid metal fast breeder reactor (LMFBR), specifically pool type LMFBR's, are summarized. The information presented, for the most part, draws upon existing data. Special sections are devoted to design, technical feasibility (normal operating conditions), and safety (accident conditions). A survey of world fast reactors is presented in tabular form, as are two sets of reference reactor parameters based on available data from present and conceptual LMFBR's. (auth)

Combined use of the RPI [Rensselaer Polytechnic Institute] reactor for training and critical experiments

International Nuclear Information System (INIS)

Harris, D.R.; Rohr, R.R.; Rodriguez-Vera, F.

1990-01-01

The Rensselaer Polytechnic Institute (RPI) reactor critical facility (RCF) has provided educational and research opportunities for RPI and other students for >25 yr. The RCF was built by the American Locomotive Company (ALCO) in the 1950s as a critical facility in support of the army package power reactor program, and, when ALCO went out of business in 1964, the RCF was acquired by RPI. Since that time, RPI has operated the RCF primarily in a teaching mode in the nuclear engineering department, although reactor research, activation analyses, and reactivity assays have been carried out as well. Until recently, the RCF was fueled by plates containing highly enriched uranium as a cermet in stainless steel. This highly enriched uranium (HEU) fuel was replaced recently by 4.81 wt% enriched UO 2 high-density pellets clad in stainless steel rods. The use of these SPERT (F1) fuel rods in the RCF provided a cost-effective method for conversion of the core from HEU to low-enriched uranium and for enhancement of the RCF training and research program. The RCF is the only facility in the United States that provides reactor training on a core containing fuel that is similar to that used in power industry light water reactors (LWRs). Moreover, the RCF is the only facility in the United States currently available for supplying critical experimental data in support of the LWR power industry. Thus, the RCF is in a unique position to carry out important training and research services consistent with RPI's nuclear engineering objectives

Progress report on research and development in 1991, Institute of Reactor Development, KfK

International Nuclear Information System (INIS)

1992-03-01

Progress report on research and development in 1991 Institute of Reactor Development. The papers on nuclear fusion concentrate on the design and material selection for highly stressed components as well as on safety matters. Experiments with the thermomechanical behaviour of different material samples continued, with selected materials being put to a load of up to 10 000 cycles. Carbon fiber reinforced composite materials proved to be very stable as regards their form, and unproblematic from a thermomechanical viewpoint, even at high cycle numbers. The papers on handling techniques refer to specific requirements of nuclear fusion with applications at JET and NET, to the development of system solutions to be used in the classical industrial area, and to standardization accompanying the developments. The system for physical simulation of working scenes was refined and extended by models for the prototype of a testing device to be handled in the torus of a fusion machine. Control of the articulated boom has been further improved. Under the nuclear safety research project, studies have been made of the dynamic behaviour of fast reactors under incident conditions, of the possible propagation of local cooling incidents in the reactor core as well as of core monitoring. The further development of physical models and computer programs on the dynamic behaviour of fast sodium-cooled reactors has been supported by experimental results. (orig./DG) [de

Design measures for prevention and mitigation of severe accidents at advanced water cooled reactors. Proceedings of a technical committee meeting

International Nuclear Information System (INIS)

1998-06-01

Over 8500 reactor-years of operating experience have been accumulated with the current nuclear energy systems. New generations of nuclear power plants are being developed, building upon this background of experience. During the last decade, requirements for equipment specifically intended to minimize releases of radioactive material to the environment in the event of a core melt accident have been introduced, and designs for new plants include measures for preventing and mitigating a range of severe accident scenarios. The IAEA Technical Committee Meeting on Impact of Severe Accidents on Plant Design and Layout of Advanced Water Cooled Reactors was jointly organized by the Department of Nuclear Energy and the Department of Nuclear Safety to review measures which are being incorporated into advanced water cooled reactor designs for preventing and mitigating severe accidents, the status of experimental and analytical investigations of severe accident phenomena and challenges which support design decisions and accident management procedures, and to understand the impact of explicitly addressing severe accidents on the cost of nuclear power plants. This publication is intended to provide an objective source of information on this topic. It includes 14 papers presented at the Technical Committee meeting held in Vienna between 21-25 October 1996. It also includes a Summary and Findings of the Working Groups. The papers were grouped in three sections. A separate abstract was prepared for each paper

Safety studies concerning nuclear power reactors

International Nuclear Information System (INIS)

Bailly, Jean; Pelce, Jacques

1980-01-01

The safety of nuclear installations poses different technical problems, whether concerning pressurized water reactors or fast reactors. But investigating methods are closely related and concern, on the one hand, the behavior of shields placed between fuel and outside and, on the other, analysis of accidents. The article is therefore in two parts based on the same plan. Concerning light water reactors, the programme of studies undertaken in France accounts for the research carried out in countries where collaboration agreements exist. Concerning fast reactors, France has the initiative of their studies owing to her technical advance, which explains the great importance of the programmes under way [fr

Retrospect over past 25 years at Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology

International Nuclear Information System (INIS)

Aoki, Shigebumi

1983-01-01

Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology, was established on April 1, 1956, with the aims of the investigation on the peaceful use of nuclear energy and of the education of scientists and engineers in this field. This report reviews the history of the Laboratory during 25 years and traces the process of growth concerning research divisions, buildings, large-scale experimental facilities and the education in the graduate course for nuclear engineering. In addition, considering what the Laboratory has to be and what the future plan will be, it is mentioned that the research interest should be extended to the field of nuclear fusion reactor, especially the blanket engineering, as a long-term future project of the Research Laboratory. (author)

Materials for advanced water cooled reactors

International Nuclear Information System (INIS)

1992-09-01

The current IAEA programme in advanced nuclear power technology promotes technical information exchange between Member States with major development programmes. The International Working Group on Advanced Technologies for Water Cooled Reactors recommended to organize a Technical Committee Meeting for the purpose of providing an international forum for technical specialists to review and discuss aspects regarding development trends in material application for advanced water cooled reactors. The experience gained from the operation of current water cooled reactors, and results from related research and development programmes, should be the basis for future improvements of material properties and applications. This meeting enabled specialists to exchange knowledge about structural materials application in the nuclear island for the next generation of nuclear power plants. Refs, figs, tabs

Computerized reactor monitor and control for research reactors

International Nuclear Information System (INIS)

Buerger, L.; Vegh, E.

1981-09-01

The computerized process control system developed in the Central Research Institute for Physics, Budapest, Hungary, is described together with its special applications at research reactors. The nuclear power of the Hungarian research reactor is controlled by this computerized system, too, while in Lybia many interesting reactor-hpysical calculations are built into the computerized monitor system. (author)

The development of reactor vessel internal heavy forging for 1000 MW pressurized-water reactor nuclear power plant

International Nuclear Information System (INIS)

Zhang Zhifeng; Chen Yongbo; Ding Xiuping; Zhang Lingfang

2012-01-01

This Paper introduced the development of Reactor Vessel Internal (RVI) heavy forgings for 1000 MW Pressurized Water Reactor (PWR) nuclear power plant, analyzed the manufacture difficulties and technical countermeasures. The testing result of the product indicated that the performance of RVI heavy forgings manufactured by Shanghai Heavy Machinery Plant Ld. (SHMP) is outstanding and entirely satisfy the technical requirements for RVI product. (authors)

20 CFR 411.167 - What is an educational institution or a technical, trade or vocational school?

Science.gov (United States)

2010-04-01

... Education. (e) Approval by a State agency or subdivision of the State includes approval of a school, college... 20 Employees' Benefits 2 2010-04-01 2010-04-01 false What is an educational institution or a technical, trade or vocational school? 411.167 Section 411.167 Employees' Benefits SOCIAL SECURITY...

[Fortieth Annual] Meeting of the Technical Working Group on Fast Reactors (TWG-FR). Working Material

International Nuclear Information System (INIS)

2008-01-01

The objectives of the meeting were to: - Exchange information on the national programmes on Fast Reactors (FR) and Accelerator Driven Systems (ADS); - Review the progress since the 39th TWG-FR Annual Meeting, including the status of the actions; - Consider meeting arrangements for 2007, 2008 and 2009; - Review the Agency’s ongoing information exchange and co-ordinated research activities in the technical fields relevant to the TWG-FR (FRs and ADS), as well as coordination of the TWG-FR’s activities with other organizations; - Discuss future joint activities in view of the Agency’s Programme and Budget Cycle 2008–2009 (and beyond)

NRC review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document - Evolutionary plant designs, Chapter 1, Project No. 669

International Nuclear Information System (INIS)

1992-08-01

The staff of the US Nuclear Regulatory Commission has prepared Volume 2 (Parts 1 and 2) of a safety evaluation report (SER), ''NRC Review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document -- Evolutionary Plant Designs,'' to document the results of its review of the Electric Power Research Institute's ''Advanced Light Water Reactor Utility Requirements Document.'' This SER gives the results of the staff's review of Volume II of the Requirements Document for evolutionary plant designs, which consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant (approximately 1300 megawatts-electric)

Experimental tests and qualification of analytical methods to address thermohydraulic phenomena in advanced water cooled reactors. Proceedings of a technical committee meeting

International Nuclear Information System (INIS)

2000-05-01

Worldwide there is considerable experience in nuclear power technology, especially in water cooled reactor technology. Of the operating plants, in September 1998, 346 were light water reactors (LWRs) totalling 306 GW(e) and 29 were heavy water reactors (HWRs) totalling 15 GW(e). The accumulated experience and lessons learned from these plants are being incorporated into new advanced reactor designs. Utility requirements documents have been formulated to guide these design activities by incorporating this experience, and results from research and development programmes, with the aim of reducing costs and licensing uncertainties by establishing the technical bases for the new designs. Common goals for advanced designs are high availability, user-friendly features, competitive economics and compliance with internationally recognized safety objectives. Large water cooled reactors with power outputs of 1300 MW(e) and above, which possess inherent safety characteristics (e.g. negative Doppler moderator temperature coefficients, and negative moderator void coefficient) and incorporate proven, active engineered systems to accomplish safety functions are being developed. Other designs with power outputs from, for example, 220 MW(e) up to about 1300 MW(e) which also possess inherent safety characteristics and which place more emphasis on utilization of passive safety systems are being developed. Passive systems are based on natural forces and phenomena such as natural convection and gravity, making safety functions less dependent on active systems and components like pumps and diesel generators. In some cases, further experimental tests for the thermohydraulic conditions of interest in advanced designs can provide improved understanding of the phenomena. Further, analytical methods to predict reactor thermohydraulic behaviour can be qualified for use by comparison with the experimental results. These activities should ultimately result in more economical designs. The

Space reactor fuels performance and development issues

International Nuclear Information System (INIS)

Wewerka, E.M.

1984-01-01

Three compact reactor concepts are now under consideration by the US Space Nuclear Power Program (the SP-100 Program) as candidates for the first 100-kWe-class space reactor. Each of these reactor designs puts unique constraints and requirements on the fuels system, and raises issues of fuel systems feasibility and performance. This paper presents a brief overview of the fuel requirements for the proposed space reactor designs, a delineation of the technical feasibility issues that each raises, and a description of the fuel systems development and testing program that has been established to address key technical issues

PWR reactor pressure vessel internals license renewal industry report; revision 1. Final report

International Nuclear Information System (INIS)

Schwirian, R.; Robison, G.

1994-07-01

The U.S. nuclear power industry, through coordination by the Nuclear Management and Resources Council (NUMARC), and sponsorship by the U.S. Department of Energy (DOE) and the Electric Power Research Institute (EPRI), has evaluated age-related degradation effects for a number of major plant systems, structures and components, in the license renewal technical Industry Reports (IRs). License renewal applicants may choose to reference these IRs in support of their plant-specific license renewal applications, as an equivalent to the integrated plant assessment provisions of the license renewal rule (10 CFR Part 54). Pressurized water reactor (PWR) reactor pressure vessel (RPV) internals designed by all three U.S. PWR nuclear steam supply system vendors have been evaluated relative to the effects of age-related degradation mechanisms; the capability of current design limits; inservice examination, testing, repair, refurbishment, and other programs to manage these effects; and the assurance that these internals can continue to perform their intended safety functions in the license renewal term. This industry report (IR), one of a series of ten, provides a generic technical basis for evaluation of PWR reactor pressure vessel internals for license renewal

IEA-R1 research reactor: operational life extension and considerations regarding future decommissioning

International Nuclear Information System (INIS)

Frajndlich, Roberto

2009-01-01

The IEA-R1 reactor is a pool type research reactor moderated and cooled by light water and uses graphite and beryllium reflectors. The reactor is located at the Instituto de Pesquisas Energeticas e Nucleares (IPEN-CNEN/SP), in the city of Sao Paulo, Brazil. It is the oldest research reactor in the southern hemisphere and one of the oldest of this kind in the world. The first criticality of the reactor was obtained on September 16, 1957. Given the fact that Brazil does not have yet a definitive radioactive waste repository and a national policy establishing rules for the spent fuel storage, the institutions which operate the research reactors for more than 50 years in the country have searched internal solutions for continued operation. This paper describes the spent fuel assemblies and radioactive waste management process for the IEA-R1 reactor and the refurbishment and modernization program adopted to extend its lifetime. Some considerations about the future decommissioning of the reactor are also discussed which, in my opinion, might help the operating organization to make decisions about financial, legal and technical aspects of the decommissioning procedures in a time frame of 10-15 years(author)

Gas Reactor International Cooperative Program. Interim report. Construction and operating experience of selected European Gas-Cooled Reactors

International Nuclear Information System (INIS)

1978-09-01

The construction and operating experience of selected European Gas-Cooled Reactors is summarized along with technical descriptions of the plants. Included in the report are the AVR Experimental Pebble Bed Reactor, the Dragon Reactor, AGR Reactors, and the Thorium High Temperature Reactor (THTR). The study demonstrates that the European experience has been favorable and forms a good foundation for the development of Advanced High Temperature Reactors

Gas-cooled fast breeder reactor

International Nuclear Information System (INIS)

Yoshida, Hiroyuki

1982-07-01

Almost all the R D works of gas-cooled fast breeder reactor in the world were terminated at the end of the year 1980. In order to show that the R D termination was not due to technical difficulties of the reactor itself, the present paper describes the reactor plant concept, reactor performances, safety, economics and fuel cycle characteristics of the reactor, and also describes the reactor technologies developed so far, technological problems remained to be solved and planned development schedules of the reactor. (author)

RA Research reactor, Annual report 1988

International Nuclear Information System (INIS)

Sotic, O.

1988-12-01

Annual report concerning the project 'RA research nuclear reactor' for 1989, financed by the Serbian ministry of science is divided into two parts. First part is concerned with RA reactor operation and maintenance, which is the task of the Division for reactor engineering of the Institute for multidisciplinary studies and RA reactor engineering. Second part deals with radiation protection activities at the RA reactor which is the responsibility of the Institute for radiation protection. Scientific council of the Institute for multidisciplinary studies and RA reactor engineering has stated that this report describes adequately the activity and tasks fulfilled at the RA reactor in 1989. The scope and the quality of the work done were considered successful both concerning the maintenance and reconstruction, as well as radiation protection activities [sr

Uncertainty Quantification in the Reliability and Risk Assessment of Generation IV Reactors: Final Scientific/Technical Report

Energy Technology Data Exchange (ETDEWEB)

Vierow, Karen; Aldemir, Tunc

2009-09-10

The project entitled, “Uncertainty Quantification in the Reliability and Risk Assessment of Generation IV Reactors”, was conducted as a DOE NERI project collaboration between Texas A&M University and The Ohio State University between March 2006 and June 2009. The overall goal of the proposed project was to develop practical approaches and tools by which dynamic reliability and risk assessment techniques can be used to augment the uncertainty quantification process in probabilistic risk assessment (PRA) methods and PRA applications for Generation IV reactors. This report is the Final Scientific/Technical Report summarizing the project.

Annual report on operation, utilization and technical development of research reactors and hot laboratory, from April 1, 1987 to March 31, 1988

International Nuclear Information System (INIS)

1988-10-01

Activities of the Department of Research Reactor Operation in fiscal year 1987 are described. The department is responsible for operation and maintenance of JRR-2, JRR-4, Research Reactor Development Division which performed upgraded JRR-3 and other R D, and Hot Laboratory. In the above connection various other work has also been performed, such as technical management of fuel and coolant, radiation control, irradiation technique, etc. In Hot Laboratory, we have performed post-irradiation examinations of fuels and materials, and also development of examination procedures, too. (author)

Annual report on operation, utilization and technical development of research reactors and hot laboratory, from April 1, 1985 to March 31, 1986

International Nuclear Information System (INIS)

1986-10-01

Activities of the Department of Research Reactor Operation in fiscal year 1985 are described. The department is responsible for operation and maintenance of JRR-2, JRR-4, Research Reactor Development Division which performed upgraded JRR-3 and other R and D, and Hot Laboratory. In the above connection various other work has also been performed, such as technical management of fuel and coolant, radiation control, irradiation technique, etc. In Hot Laboratory, we have performed post-irradiation examinations of fuels and materials, and also development of examination procedures, too. (author)

Flood Protection in Venice under Conditions of Sea-Level Rise: An Analysis of Institutional and Technical Measures

NARCIS (Netherlands)

Munaretto, S.; Vellinga, P.; Tobi, H.

2012-01-01

It is widely acknowledged that in times of climate change loss of coastal resources and risk for human life can be minimized by implementing adaptation strategies. Such strategies need to encompass a balanced mix of non-structural (institutional) and structural (technical) measures based on sound

Research reactors and materials testing

International Nuclear Information System (INIS)

Vidal, H.

1986-01-01

Research reactors can be classified in three main groups according to the moderator which is used. Their technical characteristics are given and the three most recent research and materials testing reactors are described: OSIRIS, ORPHEE and the high-flux reactor of Grenoble. The utilization of research reactors is reviewed in four fields of activity: training, fundamental or applied research and production (eg. radioisotopes) [fr

Gas-Cooled Thermal Reactor Program. Annual technical progress report for the period ending September 30, 1981

International Nuclear Information System (INIS)

1982-03-01

This report provides descriptions and results of the technical effort during FY81 on the Gas-Cooled Thermal Reactor Program. The FY81 work was organized according to the Work Breakdown Structure (WBS) for the National HTGR Program, and fell within five of the WBS tasks. The work on Market Definition and Development (WBS 03) was associated with estimating product costs for HTGR systems and their alternatives, projecting markets and market penetrations for these systems, and providing costs and market input to application analyses and component design. The Plant Technology (WBS 13) effort was mainly in the development of the systems dynamic computer code, STAR, for the transient analysis of HTGR's in reformer applications. The analysis of pebble bed reactors (PBR) was performed under Technology Transfer (WBS 15). The effort on components and systems within the nuclear heat source for reforming plants was performed under High Temperature Nuclear Heat Source (WBS 42)

Progress report on R + D activities of the Institute fuer Reaktorenentwicklung in 1980

International Nuclear Information System (INIS)

1981-02-01

The Institute fuer Reaktorentwicklung is mostly concerned with technical problems in the development and safety assurance of nuclear facilities. The tasks, some of which are carried out in cooperation with other internal and external institutions, are of a theoretical and experimental nature, most of them in the field of reactor safety. The two main fields are the Projekt Schneller Brueter (PSB) and Projekt Nukleare Sicherheit (PNS). The contributions for PSB, i.e. layout and safety of sodium-cooled fast breeder reactors, refer to investigations of the thermohydraulic and neutron-kinetical response of the core to accidents of very low probability as well as investigations of the effects of local core coling disturbances, with special consideration to possible damage propagation mechanisms and the possibility of detecting these processes. In the field of LWR safety, dynamic loads on core internals have been investigated during the HDR blowdown experiments as well as dynamic loads on containment structures and fuel rod behaviour during LOCA. (orig.) [de

Report on the Survey of the Design Review of New Reactor Applications. Volume 3: Reactor

International Nuclear Information System (INIS)

Downey, Steven; Monninger, John; Nevalainen, Janne; Lorin, Aurelie; ); Webster, Philip; Joyer, Philippe; Kawamura, Tomonori; Lankin, Mikhail; Kubanyi, Jozef; Haluska, Ladislav; Persic, Andreja; Reierson, Craig; Kang, Kyungmin; Kim, Walter

2016-01-01

At the tenth meeting of the CNRA Working Group on the Regulation of New Reactors (WGRNR) in March 2013, the Working Group agreed to present the responses to the Second Phase, or Design Phase, of the Licensing Process Survey as a multi-volume text. As such, each report will focus on one of the eleven general technical categories covered in the survey. The general technical categories were selected to conform to the topics covered in the International Atomic Energy Agency (IAEA) Safety Guide GS-G-4.1. This document, which is the third report on the results of the Design Phase Survey, focuses on the Reactor. The Reactor category includes the following technical topics: fuel system design, reactor internals and core support, nuclear design and core nuclear performance, thermal and hydraulic design, reactor materials, and functional design of reactivity control system. For each technical topic, the member countries described the information provided by the applicant, the scope and level of detail of the technical review, the technical basis for granting regulatory authorisation, the skill sets required and the level of effort needed to perform the review. Based on a comparison of the information provided by the member countries in response to the survey, the following observations were made: - Although the description of the information provided by the applicant differs in scope and level of detail among the member countries that provided responses, there are similarities in the information that is required. - All of the technical topics covered in the survey are reviewed in some manner by all of the regulatory authorities that provided responses. - Design review strategies most commonly used to confirm that the regulatory requirements have been met include document review and independent verification of calculations, computer codes, or models used to describe the design and performance of the core and the fuel. - It is common to consider operating experience and

Twenty years of chemistry associated with the needs and utilization of nuclear reactors at the 'Boris Kidric' Institute of nuclear sciences, Vinca, Yugoslavia

International Nuclear Information System (INIS)

1969-01-01

This publication covers nine review papers on the following topics related to the needs and utilization of nuclear reactors in the Boris Kidric Institute of nuclear sciences during previous twenty years: radiochemistry, hot atom chemistry, isotope production, spent nuclear fuel reprocessing, chemistry of transuranium elements; liquid radioactive waste processing, purification of reactor coolant water by inorganic ion exchangers, research related to deuterium concentration processes, and chemical dosimetry at the RA reactor [sr

The safety of Ontario's nuclear power reactor. A scientific and technical review. Report to the Minister

International Nuclear Information System (INIS)

Hare, F.K.

1988-01-01

In December 1986 a study of the safety of the design, operating procedures and emergency plans associated with Ontario Hydro's nuclear generating plants was commissioned by the government of the province of Ontario. After receiving briefs from many interested groups and individuals, visiting the power plants, and consulting with nuclear industry and regulatory representatives in Canada and other countries, the commissioner presented this report to the Minister of Energy for Ontario. His major conclusion is that Ontario Hydro reactors are being operated safely and at high standards of technical performance. No significant adverse impact has been detected in either the work force or the public. The risk of accidents serious enough to affect the public adversely can never be zero, but is very remote. Major recommendations are that: Ontario Hydro re-examine its operational organization closely and commission a study of factors affecting human performance; and, that priority be given to finding a solution to pressure tube performance problems and to improving in-reactor monitoring. Sixteen other recommendations are presented relating to research and development, information exchange with other organizations, reactor performance, training, severe accident analysis, the provincial nuclear emergency plan, epidemiological studies, the Atomic Energy Control Board, public hearings, and women in the nuclear industry

Guidebook on destructive examination of water reactor fuel

International Nuclear Information System (INIS)

1997-01-01

As a result of common efforts of fuel vendors, utilities and research institutes the average burnup pf design batch fuels was increased for both PWRs and BWRs and the fuel failure rate has been reduced. The previously published Guidebook on Non-Destructive Examination of Water Reactor Fuel recommended that more detailed destructive techniques are required for complete understanding of fuel performance. On the basis of contributions of the 14 participants in the ED-WARF-II CRP and proceedings of IAEA Technical Committee on Recent Developments in Post-irradiation Examination Techniques for Water Reactor Fuel this guidebook was compiled. It gives a complete survey of destructive techniques available to date worldwide. The following examination techniques are described in detailed including major principles of equipment design: microstructural studies; elemental analysis; isotopic analysis; measurement of physical properties; measurement of mechanical properties. Besides the examination techniques, methods for refabrication of experimental rods from high burnup power reactor rods as well as methods for verification of non-destructive techniques by using destructive techniques is included

Scientific and Technical Report, 1997

International Nuclear Information System (INIS)

Metivier, Henri; Mur, Emmanuelle

1998-01-01

This report highlights the most significant scientific and technical achievements of the Institute of Nuclear Protection and Safety in 1997. In all 23 reports are grouped in the following 7 sections: 1.Reactor safety; 2.Safety of nuclear facility; 3.Safeguards for radioactive materials and transports; 4.Public protection; 5.Environment protection; 6.Radioactive waste safety; 7.Crisis management. Special attention is paid to the pressurized water and fast neutron reactors and, accordingly, the program PHEBUS PF and the first results obtained with it concerning nuclear safety are reported. The safety of nuclear fuel cycle and spent fuel storage is addressed in a couple of reports, as well as the problem of inspection, measurement and containers for nuclear materials. Concerning the public and environment protection, studies are reported on medical surveillance of people working with CEA Group, and clinical dosimetric follow-up of radiological events as well as the strategy of contaminated site rehabilitation. In the field of radioactive waste management, the IPSN project of Tournemire tunnel and the project of deep disposal are presented. Finally, the issue of major nuclear accidents is approached through the exercise Becquerel, the SESAME system of predicting the radioactive products release in PWR reactors and ASTRAL, the evaluation system of aftermath of an accidental environmental contamination by radioactive release. Each section of this IPSN report is opened by a review paper

Licensed reactor nuclear safety criteria applicable to DOE reactors

International Nuclear Information System (INIS)

1993-11-01

This document is a compilation and source list of nuclear safety criteria that the Nuclear Regulatory Commission (NRC) applies to licensed reactors; it can be used by DOE and DOE contractors to identify NRC criteria to be evaluated for application to the DOE reactors under their cognizance. The criteria listed are those that are applied to the areas of nuclear safety addressed in the safety analysis report of a licensed reactor. They are derived from federal regulations, USNRC regulatory guides, Standard Review Plan (SRP) branch technical positions and appendices, and industry codes and standards

Fast reactors worldwide

International Nuclear Information System (INIS)

Hall, R.S.; Vignon, D.

1985-01-01

The paper concerns the evolution of fast reactors over the past 30 years, and their present status. Fast reactor development in different countries is described, and the present position, with emphasis on cost reduction and collaboration, is examined. The French development of the fast breeder type reactor is reviewed, and includes: the acquisition of technical skills, the search for competitive costs and the spx2 project, and more advanced designs. Future prospects are also discussed. (U.K.)

Heat Transfer Behaviour and Thermohydraulics Code Testing for Supercritical Water Cooled Reactors (SCWRs)

International Nuclear Information System (INIS)

2014-08-01

The supercritical water cooled reactor (SCWR) is an innovative water cooled reactor concept which uses water pressurized above its thermodynamic critical pressure as the reactor coolant. This concept offers high thermal efficiencies and a simplified reactor system, and is hence expected to help to improve economic competitiveness. Various kinds of SCWR concepts have been developed, with varying combinations of reactor type (pressure vessel or pressure tube) and core spectrum (thermal, fast or mixed). There is great interest in both developing and developed countries in the research and development (R&D) and conceptual design of SCWRs. Considering the high interest shown in a number of Member States, the IAEA established in 2008 the Coordinated Research Project (CRP) on Heat Transfer Behaviour and Thermo-hydraulics Code Testing for SCWRs. The aim was to foster international collaboration in the R&D of SCWRs in support of Member States’ efforts and under the auspices of the IAEA Nuclear Energy Department’s Technical Working Groups on Advanced Technologies for Light Water Reactors (TWG-LWR) and Heavy Water Reactors (TWG-HWR). The two key objectives of the CRP were to establish accurate databases on the thermohydraulics of supercritical pressure fluids and to test analysis methods for SCWR thermohydraulic behaviour to identify code development needs. In total, 16 institutes from nine Member States and two international organizations were involved in the CRP. The thermohydraulics phenomena investigated in the CRP included heat transfer and pressure loss characteristics of supercritical pressure fluids, development of new heat transfer prediction methods, critical flow during depressurization from supercritical conditions, flow stability and natural circulation in supercritical pressure systems. Two code testing benchmark exercises were performed for steady state heat transfer and flow stability in a heated channel. The CRP was completed with the planned outputs in

Heat Transfer Behaviour and Thermohydraulics Code Testing for Supercritical Water Cooled Reactors (SCWRs)

Energy Technology Data Exchange (ETDEWEB)

NONE

2014-08-15

The supercritical water cooled reactor (SCWR) is an innovative water cooled reactor concept which uses water pressurized above its thermodynamic critical pressure as the reactor coolant. This concept offers high thermal efficiencies and a simplified reactor system, and is hence expected to help to improve economic competitiveness. Various kinds of SCWR concepts have been developed, with varying combinations of reactor type (pressure vessel or pressure tube) and core spectrum (thermal, fast or mixed). There is great interest in both developing and developed countries in the research and development (R&D) and conceptual design of SCWRs. Considering the high interest shown in a number of Member States, the IAEA established in 2008 the Coordinated Research Project (CRP) on Heat Transfer Behaviour and Thermo-hydraulics Code Testing for SCWRs. The aim was to foster international collaboration in the R&D of SCWRs in support of Member States’ efforts and under the auspices of the IAEA Nuclear Energy Department’s Technical Working Groups on Advanced Technologies for Light Water Reactors (TWG-LWR) and Heavy Water Reactors (TWG-HWR). The two key objectives of the CRP were to establish accurate databases on the thermohydraulics of supercritical pressure fluids and to test analysis methods for SCWR thermohydraulic behaviour to identify code development needs. In total, 16 institutes from nine Member States and two international organizations were involved in the CRP. The thermohydraulics phenomena investigated in the CRP included heat transfer and pressure loss characteristics of supercritical pressure fluids, development of new heat transfer prediction methods, critical flow during depressurization from supercritical conditions, flow stability and natural circulation in supercritical pressure systems. Two code testing benchmark exercises were performed for steady state heat transfer and flow stability in a heated channel. The CRP was completed with the planned outputs in

Regulations for RA reactor operation

International Nuclear Information System (INIS)

1980-09-01

Regulations for RA reactor operation are written in accordance with the legal regulations defined by the Law about radiation protection and related legal acts, as well as technical standards according to the IAEA recommendations. The contents of this book include: fundamental data about the reactor; legal regulations for reactor operation; organizational scheme for reactor operation; general and detailed instructions for operation, behaviour in the reactor building, performing experiments; operating rules for operation under steady state and accidental conditions [sr

Standard technical specifications: Babcock and Wilcox Plants. Revision 1

International Nuclear Information System (INIS)

1995-04-01

This report documents the results of the combined effort of the NRC and the industry to produce improved Standard Technical Specifications (STS), Revision 1 for Babcock ampersand Wilcox Plants. The changes reflected in Revision 1 resulted from the experience gained from license amendment applications to convert to these improved STS or to adopt partial improvements to existing technical specifications. This NUREG is the result of extensive public technical meetings and discussions between the Nuclear Regulatory Commission (NRC) staff and various nuclear power plant licensees, Nuclear Steam Supply System (NSSS) Owners Groups, NSSS vendors, and the Nuclear Energy Institute (NEI). The improved STS were developed based on the criteria in the Final Commission Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, dated July 22, 1993. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume 1 contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. Volume 3 contains the Bases for Sections 3.4--3.9 of the improved STS

Productivity Measurement Of Technical Institute

Directory of Open Access Journals (Sweden)

Miss.Pooja Yadav

2015-05-01

Full Text Available Abstract The increasing concern over globalization between other things has made the achievement of lifetime skills vital for all classes of people. Achievement of such skills can be developed from technical education programs in the advance of any nation. Skill gaining provides knowledge and inculcates the approaches that are essential for arrival and progress into an occupation. When skills are developed in any occupation it will provide and improve the standard of living with the cover beside poverty thereby behind national development. This paper observes sampling distribution and moving average concept that how technical education can generate the required skills the significance of skill acquisition its roles benefits and sustainability in national development. It also mentions among others that government should look into and improve productivity that can encourage skill achievement for the youths. The solutions are divided the sampling box into few group before stacking up workstation and use table and some other wood product. Cost reduction for each solution was intended and the solution with highest cost decrease is chosen to increase the productivity of the college.

The Role of Stakeholders in the Decommissioning of Salaspils Research Reactor

International Nuclear Information System (INIS)

Abramenkovs, A.

2009-01-01

The paper describes the role of different stakeholders in the decommissioning of the Salaspils Research Reactor. Decommissioning was a large challenge for the Latvia, since the country in this moment had no decommissioning experience and necessary technologies for the implementation of the defined goals by the Government. In this case for facilitation of the decommissioning of Salaspils Research Reactor (SRR), the significant role plays the local and international stakeholders. The paper deals with information on the basic stages of decommissioning of SRR and the role of the wide spectrum of stakeholders in preparation, upgrade and implementation of the decommissioning plan. The role of governmental institutions in the decommissioning of Salaspils research reactor is discussed. It was shown, that local municipalities are very important stakeholders, which significantly influence the decommissioning of SRR. The Salaspils municipalities positive impact on the decommissioning processes are discussed. Basic problems with the Baldone municipality in context of radioactive wastes management are indicated. The role of international stakeholders in decommissioning of Salaspils research reactor is discussed. It was shown, that the support from International Atomic Energy Agency significantly promotes the decommissioning of SRR. The main issues were expert support for solution of different technical problems in radioactive wastes management, area monitoring, and verification of decommissioning plans, training of staff and technical expertise during whole process of decommissioning. It was shown, that technical and economical support from DOE, USA provides the possibility to solve the fuel problem during decommissioning of SRR, as well as, to increase the physical safety of SRR and repository 'Radons'. It was shown, that a proper coordination of all activities and using the services from stakeholders can significantly reduce the total project expenses. The cooperation between

Preparing the construction of a school reactor

International Nuclear Information System (INIS)

Matejka, K.

1977-01-01

The possibilities are discussed of teaching and training nuclear reactor operation and control, teaching experimental reactor physics and investigating reactor lattice parameters using a training reactor to be installed at the Faculty of Nuclear Science and Physical Engineering in Prague. Requirements are indicated for the reactor's technical design and the Faculty's possibilities to contribute to its construction. (J.B.)

Comparison between TRU burning reactors and commercial fast reactor

International Nuclear Information System (INIS)

Fujimura, Koji; Sanda, Toshio; Ogawa, Takashi

2001-03-01

Research and development for stabilizing or shortening the radioactive wastes including in spent nuclear fuel are widely conducted in view point of reducing the environmental impact. Especially it is effective way to irradiate and transmute long-lived TRU by fast reactors. Two types of loading way were previously proposed. The former is loading relatively small amount of TRU in all commercial fast reactors and the latter is loading large amount of TRU in a few TRU burning reactors. This study has been intended to contribute to the feasibility studies on commercialized fast reactor cycle system. The transmutation and nuclear characteristics of TRU burning reactors were evaluated and compared with those of conventional transmutation system using commercial type fast reactor based upon the investigation of technical information about TRU burning reactors. Major results are summarized as follows. (1) Investigation of technical information about TRU burning reactors. Based on published reports and papers, technical information about TRU burning reactor concepts transmutation system using convectional commercial type fast reactors were investigated. Transmutation and nuclear characteristics or R and D issue were investigated based on these results. Homogeneously loading of about 5 wt% MAs on core fuels in the conventional commercial type fast reactor may not cause significant impact on the nuclear core characteristics. Transmutation of MAs being produced in about five fast reactors generating the same output is feasible. The helium cooled MA burning fast reactor core concept propose by JAERI attains criticality using particle type nitride fuels which contain more than 60 wt% MA. This reactor could transmute MAs being produced in more than ten 1000 MWe-LWRs. Ultra-long life core concepts attaining more than 30 years operation without refueling by utilizing MA's nuclear characteristics as burnable absorber and fertile nuclides were proposed. Those were pointed out that

NRC review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document - Evolutionary plant designs, Chapters 2--13, Project No. 669

International Nuclear Information System (INIS)

1992-08-01

The staff of the US Nuclear Regulatory Commission has prepared Volume 2 (Parts 1 and 2) of a safety evaluation report (SER), ''NRC Review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document -- Evolutionary Plant Designs,'' to document the results of its review of the Electric Power Research Institute's ''Advanced Light Water Reactor Utility Requirements Document.'' This SER gives the results of the staff's review of Volume II of the Requirements Document for evolutionary plant designs, which consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant (approximately 1300 megawatts-electric)

Review of light water reactor safety through the Three Mile Island accident

International Nuclear Information System (INIS)

Phung, D.L.

1984-05-01

This review of light water reactor safety through the Three Mile Island accident has the purpose of establishing the baseline over which safety achievement post-TMI is assessed, and the need for new reactor designs and business direction is judged. Five major areas of reactor safety pre-TMI are examined: (1) safety philosophy and institutions, (2) reactor design criteria, (3) operational problems, (4) the Rasmussen reactor safety study, and (5) the TMI accident and repercussions. Although nuclear power has made spectacular achievements over the period pre-TMI and although TMI is technically a minor accident, this review concludes that there were basic flaws in the technology and in the manner safety philosophy was conceived and carried out. These flaws included (1) a reactor design that has high core power density, low heat capacity, and low system tolerance to upsets, (2) reactor deployment that had been expedited without extensive operational experience, (3) rules and regulations that had to play catch-up with commercial reactor development, (4) an industry that was fragmented, short-sighted, and tended to rely on the Nuclear Regulatory Commission for safety guidance, (5) information that was not effectively shared, and (6) attention that was inadequate to the human aspects of reactor operation and to public reaction to the specter of a reactor accident, major or minor

Fast breeder reactor research

International Nuclear Information System (INIS)

1975-01-01

Full text: The meeting was attended by 15 participants from seven countries and two international organizations. The Eighth Annual Meeting of the International Working Group on Fast Reactors (IWGFR) was attended by representatives from France, Fed. Rep. Germany, Italy, Japan, United Kingdom, Union of Soviet Socialist Republics and the United States of America - countries that have made significant progress in developing the technology and physics of sodium cooled fast reactors and have extensive national programmes in this field - as well as by representatives of the Commission of the European Communities and the IAEA. The design of fast-reactor power plants is a more difficult task than developing facilities with thermal reactors. Different reactor kinetics and dynamics, a hard neutron spectrum, larger integral doses of fuel and structural material irradiation, higher core temperatures, the use of an essentially novel coolant, and, as a result of all these factors, the additional reliability and safety requirements that are imposed on the planning and operation of sodium cooled fast reactors - all these factors pose problems that can be solved comprehensively only by countries with a high level of scientific and technical development. The exchange of experience between these countries and their combined efforts in solving the fundamental problems that arise in planning, constructing and operating fast reactors are promoting technical progress and reducing the relative expenditure required for various studies on developing and introducing commercial fast reactors. For this reason, the meeting concentrated on reviewing and discussing national fast reactor programmes. The situation with regard to planning, constructing and operating fast experimental and demonstration reactors in the countries concerned, the experience accumulated in operating them, the difficulties arising during operation and ways of over-coming them, the search for optimal designs for the power

HTGR type reactors in West Germany. Realizations and prospects

International Nuclear Information System (INIS)

Dauenert, U.

1978-01-01

The framework within which the research studies on high temperature reactors have been pursued in West Germany since 1960 is recalled. The principles guiding the present policy of the country in this domain are given: choice of a single technical conception that be applied both to reactors generating electricity and reactors producing high temperature heat for industrial processes such as coal gasification; to group the technical and industrial potentials of West Germany in this domain; financial and technical participation of electricity producers in the expected realizations; international cooperation. In this technique, West Germany is at present among the most advanced nations with the realization of a prototype 300 MWe reactor, financed by the electricity producers and a contribution of government [fr

Research and materials irradiation reactors

International Nuclear Information System (INIS)

Ballagny, A.; Guigon, B.

2004-01-01

Devoted to the fundamental and applied research on materials irradiation, research reactors are nuclear installations where high neutrons flux are maintained. After a general presentation of the research reactors in the world and more specifically in France, this document presents the heavy water cooled reactors and the water cooled reactors. The third part explains the technical characteristics, thermal power, neutron flux, operating and details the Osiris, the RHF (high flux reactor), the Orphee and the Jules Horowitz reactors. The last part deals with the possible utilizations. (A.L.B.)

Civilian Power Program. Part 1, Summary, Current status of reactor concepts

Energy Technology Data Exchange (ETDEWEB)

Author, Not Given

1959-09-01

This study group covered the following: delineation of the specific objectives of the overall US AEC civilian power reactor program, technical objectives of each reactor concept, preparation of a chronological development program for each reactor concept, evaluation of the economic potential of each reactor type, a program to encourage the the development, and yardsticks for measuring the development. Results were used for policy review by AEC, program direction, authorization and appropriation requests, etc. This evaluation encompassed civilian power reactors rated at 25 MW(e) or larger and related experimental facilities and R&D. This Part I summarizes the significant results of the comprehensive effort to determine the current technical and economic status for each reactor concept; it is based on the 8 individual technical status reports (Part III).