Nuclear and radiological safety 1980-1994. International Atomic Energy Agency Publications

International Nuclear Information System (INIS)

1995-06-01

This catalogue lists all sales publications of the International Atomic Energy Agency dealing with Nuclear and Radiological Safety issued during the period 1980-1994. The following aspects are covered: Uranium mining and milling, Fuel fabrication and storage, Nuclear power plants, Research reactors, Radiation sources and accelerators, Transport of radioactive materials, Waste repositories, Radiation protection, Accident response, Radioactive waste management, Safety analysis, Quality management, Legal and governmental aspects

Commissioning of research reactors. Safety guide

International Nuclear Information System (INIS)

2006-01-01

The objective of this Safety Guide is to provide recommendations on meeting the requirements for the commissioning of research reactors on the basis of international best practices. Specifically, it provides recommendations on fulfilling the requirements established in paras 6.44 and 7.42-7.50 of International Atomic Energy Agency, Safety of Research Reactors, IAEA Safety Standards Series No. NS-R-4, IAEA, Vienna (2005) and guidance and specific and consequential recommendations relating to the recommendations presented in paras 615-621 of International Atomic Energy Agency, Safety in the Utilization and Modification of Research Reactors, Safety Series No. 35-G2, IAEA, Vienna (1994) and paras 228-229 of International Atomic Energy Agency, Safety Assessment of Research Reactors and Preparation of the Safety Analysis Report, Safety Series No. 35-G1, IAEA, Vienna (1994). This Safety Guide is intended for use by all organizations involved in commissioning for a research reactor, including the operating organization, the regulatory body and other organizations involved in the research reactor project

International bulletin on atomic and molecular data for fusion. No. 46

International Nuclear Information System (INIS)

Botero, J.

1993-06-01

The bulletin is published by the International Atomic Energy Agency to provide atomic and molecular data relevant to fusion research and technology. In Part I the indexed papers are listed separately for (i) structure and spectra (energy levels, wavelengths; transition probabilities, oscillator strengths; interatomic potentials); (ii) atomic and molecular collisions (photon collisions; electron collisions; heavy-particle collisions; homonuclear sequences; isoelectronic sequences), and (iii) surface interactions (sputtering; chemical reactions; trapping and detrapping; surface damage; blistering, flaking; secondary electron emission). Part II contains the bibliographic data for the above listed topics and for high energy laser- and beam-matter interaction; interaction of atomic particles with fields. The atomic and molecular data needs in fusion research, as identified during the IAEA Consultants' Meeting on 'Atomic and Molecular Database for Hydrogen Recycling and Helium Exhaust from Fusion Reactors', June 1992, Vienna, are listed, covering (i) atomic and molecular collision processes, (ii) particle-surface interaction processes, and (iii) the status of data bases on atomic and molecular data and plasma-surface interactions. News on the ALADDIN (A labelled Atomic Data INterface) system is provided. Finally, a list of evaluated atomic and molecular data bases is provided

Proceedings of the international symposium on materials testing reactors

International Nuclear Information System (INIS)

Ishihara, Masahiro; Kawamura, Hiroshi

2009-01-01

This report is the Proceedings of the International Symposium on Materials Testing Reactors hosted by Japan Atomic Energy Agency (JAEA). The symposium was held on July 16 to 17, 2008, at the Oarai Research and Development Center of JAEA. This symposium was also held for the 40th anniversary ceremony of Japan Materials Testing Reactor (JMTR) from achieving its first criticality. The objective of the symposium is to exchange the information on current status, future plan and so on among each testing reactors for the purpose of mutual understanding. There were 138 participants from Argentina, Belgium, France, Indonesia, Kazakhstan, Korea, the Russian Federation, Sweden, the United State, Vietnam and Japan. The symposium was divided into four technical sessions and three topical sessions. Technical sessions addressed the general topics of 'status and future plan of materials testing reactors', 'material development for research and testing reactors', irradiation technology (including PIE technology)' and 'utilization with materials testing reactors', and 21 presentations were made. Also the topical sessions addressed 'establishment of strategic partnership', 'management on re-operation work at reactor trouble' and 'basic technology for neutron irradiation tests in MTRs', and panel discussion was made. The 21 of the presented papers are indexed individually. (J.P.N.)

On Brazil's participation in the International Project on Innovative Nuclear Reactors and Fuels Cycles (INPRO)

International Nuclear Information System (INIS)

Goncalves Filho, Orlando Joao Agostinho

2007-01-01

In response to a resolution of its 44th General Conference (GC(44)/RES/21) held in September 2000, the International Atomic Energy Agency launched in May 2001 the International Project on Innovative Nuclear Reactors and Fuels Cycles (INPRO) with the objective of supporting the safe, sustainable, economic and proliferation-resistant use of nuclear technology to meet the global energy needs of the 21st century. Brazil joined the project from its beginnings and in 2005 submitted a proposal for the screening assessment using INPRO methodology of two small-size light-water reactors as potential components of an innovative nuclear reactor system (INS) completed with a conventional open nuclear fuel cycle. The INS reactor components currently being assessed are the International Reactor Innovative and Secure (IRIS) that is being developed by an international consortium made of 21 organizations from 10 countries (Brazil included) led by the Westinghouse Company, and the Fixed Bed Nuclear Reactor (FBNR) that is being developed at the Federal University of Rio Grande do Sul. This paper gives an overview of Brazil's participation in INPRO, highlighting the objective, scope and intermediate results of the assessment study being performed, and the possibilities for participation in one or two collaborative research projects under INPRO Phase 2 Action Plan for 2008-2009. (author)

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

International Nuclear Information System (INIS)

Joo, Young Sang; Lee, Jae Han

2007-02-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Joo, Young Sang; Lee, Jae Han

2007-02-15

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

Generation IV reactors: international projects

International Nuclear Information System (INIS)

Carre, F.; Fiorini, G.L.; Kupitz, J.; Depisch, F.; Hittner, D.

2003-01-01

Generation IV international forum (GIF) was initiated in 2000 by DOE (American department of energy) in order to promote nuclear energy in a long term view (2030). GIF has selected 6 concepts of reactors: 1) VHTR (very high temperature reactor system, 2) GHR (gas-cooled fast reactor system), 3) SFR (sodium-cooled fast reactor system, 4) SCWR (super-critical water-cooled reactor system), 5) LFR (lead-cooled fast reactor system), and 6) MFR (molten-salt reactor system). All these 6 reactor systems have been selected on criteria based on: - a better contribution to sustainable development (through their aptitude to produce hydrogen or other clean fuels, or to have a high energy conversion ratio...) - economic profitability, - safety and reliability, and - proliferation resistance. The 6 concepts of reactors are examined in the first article, the second article presents an overview of the results of the international project on innovative nuclear reactors and fuel cycles (INPRO) within IAEA. The project finished its first phase, called phase-IA. It has produced an outlook into the future role of nuclear energy and defined the need for innovation. The third article is dedicated to 2 international cooperations: MICANET and HTR-TN. The purpose of MICANET is to propose to the European Commission a research and development strategy in order to develop the assets of nuclear energy for the future. Future reactors are expected to be more multiple-purposes, more adaptable, safer than today, all these developments require funded and coordinated research programs. The aim of HTR-TN cooperation is to promote high temperature reactor systems, to develop them in a long term perspective and to define their limits in terms of burn-up and operating temperature. (A.C.)

International Thermonuclear Experimental Reactor

International Nuclear Information System (INIS)

Blevins, J.D.; Stasko, R.R.

1989-09-01

An international design team comprised of members from Canada, Europe, Japan, the Soviet Union, and the United States of America, are designing an experimental fusion test reactor. The engineering and testing objectives of this International Thermonuclear Experimental Reactor (ITER) are to validate the design and to demonstrate controlled ignition, extended burn of a deuterium and tritium plasma, and achieve steady state using technology expected to be available by 1990. The concept maximizes flexibility while allowing for a variety of plasma configurations and operating scenarios. During physics phase operation, the machine produces a 22 MA plasma current. In the technology phase, the machine can be reconfigured with a thicker shield and a breeding blanket to operate with an 18 MA plasma current at a major radius of 5.5 meters. Canada's involvement in the areas of safety, facility design, reactor configuration and maintenance builds on our internationally recognized design and operational expertise in developing tritium processes and CANDU related technologies

Nuclear reactor internals arrangement

International Nuclear Information System (INIS)

Frisch, E.; Andrews, H.N.

1976-01-01

A nuclear reactor internals arrangement is disclosed which facilitates reactor refueling. A reactor vessel and a nuclear core is utilized in conjunction with an upper core support arrangement having means for storing withdrawn control rods therein. The upper core support is mounted to the underside of the reactor vessel closure head so that upon withdrawal of the control rods into the upper core support, the closure head, the upper core support and the control rods are removed as a single unit thereby directly exposing the core for purposes of refueling

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

International Nuclear Information System (INIS)

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

1995-09-01

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

International project GT-MHR - New generation of nuclear reactors

International Nuclear Information System (INIS)

Vasyaev, A.; Kodochigov, N.; Kuzavkov, N.; Kuznetsov, L.

2001-01-01

Gas turbine-modular helium reactor (GT-MHR) is the reactor of new generation, which satisfies the requirements of the progressing large-scale nuclear power engineering. The activities in GT-MHR Project started in 1995. In 1997 the Conceptual Design was developed under four-side Agreement (MINATOM, General Atomics, FRAMATOME, Fuji Electric); it has passed through the internal and international reviews, has been approved and recommended for further development as one of new trends in creation of new generation plants. Starting from 1999, the activities in the development of the Preliminary Design of the plant were deployed under the Agreement between the Government of the United States of America and the Government of the Russian Federation on Scientific and Technical Cooperation in the Management of Plutonium That Has Been Withdrawn From Nuclear Military Programs dated July 24, 1998. The activities are established under the Contract between MINATOM and OKBM Russia, and under the General Agreement between Department of Energy (DOE), USA and OKBM. The GT-MHR Project is included into 'Development Strategy of Russian Nuclear Power in the first Half of the XXI-st Century' providing for 'the participation in an international project on the development and construction of GT-MHR nuclear power plant till year 2010 and 'operation of GT-MHR prototype unit and creation of fuel fabrication facility (within framework of International Project) till year 2030'. (author)

International thermal reactor development

International Nuclear Information System (INIS)

Zebroski, E.L.

1977-01-01

The worldwide development of nuclear power plants is reviewed. Charts are presented which show the commitment to light-water reactor capacity construction with breakdown by region and country. Additional charts show the major nuclear research centers which have substantial scope in light water reactor development and extensive international activities

Method for temporary shielding of reactor vessel internals

International Nuclear Information System (INIS)

Grimm, N.P.; Sejvar, J.

1991-01-01

This patent describes a method for shielding stored internals for reactor vessel annealing. It comprises removing nuclear fuel from the reactor vessel containment building; removing and storing upper and lower core internals under water in a refueling canal storage area; assembling a support structure in the refueling canal between the reactor vessel and the stored internals; introducing vertical shielding tanks individually through a hatch in the containment building and positioning each into the support structure; introducing horizontal shielding tanks individually through a hatch in the containment building and positioning each above the stored internals and vertical tanks; draining water from the refueling canal to the level of a flange of the reactor vessel; placing an annealing apparatus in the reactor vessel; pumping the remaining water from the reactor vessel; and annealing the reactor vessel

Preliminary Design Concept for a Reactor-internal CRDM

International Nuclear Information System (INIS)

Lee, Jae Seon; Kim, Jong Wook; Kim, Tae Wan; Choi, Suhn; Kim, Keung Koo

2013-01-01

A rod ejection accident may cause severer result in SMRs because SMRs have relatively high control rod reactivity worth compared with commercial nuclear reactors. Because this accident would be perfectly excluded by adopting a reactor-internal CRDM (Control Rod Drive Mechanism), many SMRs accept this concept. The first concept was provided by JAERI with the MRX reactor which uses an electric motor with a ball screw driveline. Babcock and Wilcox introduced the concept in an mPower reactor that adopts an electric motor with a roller screw driveline and hydraulic system, and Westinghouse Electric Co. proposes an internal Control Rod Drive in its SMR with an electric motor with a latch mechanism. In addition, several other applications have been reported thus far. The reactor-internal CRDM concept is now widely adopted in many SMR designs, and this concept may also be applied in an evolutionary reactor development. So the preliminary study is conducted based on the SMART CRDM design. A preliminary design concept for a reactor-internal CRDM was proposed and evaluated through an electromagnetic analysis. It was found that there is an optimum design for the motor housing, and the results may contribute to the realization a reactor-internal CRDM for an evolutionary reactor development. More detailed analysis results will be reported later

[International Thermonuclear Experimental Reactor support

International Nuclear Information System (INIS)

Dean, S.O.

1990-01-01

This report summarizes the activities under LLNL Purchase Order B089367, the purpose of which is to ''support the University/Lawrence Livermore National Laboratory Magnetic Fusion Program by evaluating the status of research relative to other national and international programs and assist in long-range plans and development strategies for magnetic fusion in general and for ITER in particular.'' Two specific subtasks are included: ''to review the LLNL Magnet Technology Development Program in the context of the International Thermonuclear Experimental Reactor Design Study'' and to ''assist LLNL to organize and prepare materials for an International Thermonuclear Experimental Reactor Design Study information meeting.''

Results of assembly test of HTTR reactor internals

International Nuclear Information System (INIS)

Maruyama, S.; Saikusa, A.; Shiozawa, S.; Tsuji, N.; Miki, T.

1996-01-01

The assembly test of the HTTR actual reactor internals had been carried out at the works, prior to their installation in the actual reactor pressure vessel(RPV) at the construction site. The assembly test consists of several items such as examining fabricating precision of each component and alignment of piled-up structures, measuring circumferential coolant velocity profile in the passage between the simulated RPV and the reactor internals as well as under the support plates, measuring by-pass flow rate through gaps between the reactor internals, and measuring the binding force of the core restraint mechanism. Results of the test showed good performance of the HTTR reactor internals. Installation of the reactor internals in the actual RPV was started at the construction site of HTTR in April, 1995. In the installation process, main items of the assembly test at the works were repeated to investigate the reproducibility of installation. (author). 5 refs, 11 figs

Progress toward international agreement to improve reactor safety

International Nuclear Information System (INIS)

Lieberman, J.I.; Graham, B.

1993-01-01

Representatives of nearly one-half of the 114 member states of the International Atomic Energy Agency (IAEA), including the United States, have participated in the development of an international nuclear safety conventions proposed multilateral treaty to improve civil nuclear power reactor safety. A preliminary draft of the convention has been developed (referred to as the draft convention for this report), but discussions are continuing, and when the final convention text will be completed and presented to IAEA member states for signature is uncertain. This report responds to the former and current Chairman's request that we provide information on the development of the nuclear safety convention, including a discussion of (1) the draft convention's scope and objectives, (2) how the convention will be implemented and monitored, (3) the views of selected country representatives on what provisions should be included in the draft convention, and (4) the convention's potential benefits and limitations

Reliability test for reactor internals rejuvenation technology

International Nuclear Information System (INIS)

Uchiyama, Junichi

1998-01-01

41 transparencies were presented on the subject of 'Reliability test for reactor internals rejuvenation technology'. The items presented give an introduction on the management of plant life in Japan and introduce the Nuclear Power Engineering Corporation (NUPEC). The question of what reliability tests for rejuvenation of reactor internals are is discussed in some detail and an outline of each test is given. Altogether six methods to rejuvenate reactor internals are presented, two of which have already been applied to actual plants. The presentation was supported by many detailed drawings and images

Introduction to reactor internal materials for pressurized water reactor

Energy Technology Data Exchange (ETDEWEB)

Ryu, Woo Suk; Hong, Joon Hwa; Jee, Se Hwan; Lee, Bong Sang; Kuk, Il Hyun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1994-06-01

This report reviewed the R and D states of reactor internal materials in order to be a reference for researches and engineers who are concerning on localization of the materials in the field or laboratory. General structure of PWR internals and material specification for YGN 3 and 4 were reviewed. States-of-arts on R and D of stainless steel and Alloy X-750 were reviewed, and degradation mechanisms of the components were analyzed. In order to develop the good domestic materials for reactor internal, following studies would be carried out: microstructure, sensitization behavior, fatigue property, irradiation-induced stress corrosion cracking/radiation-induced segregation, radiation embrittlement. (Author) 7 refs., 14 figs., 5 tabs.,.

Introduction to reactor internal materials for pressurized water reactor

International Nuclear Information System (INIS)

Ryu, Woo Suk; Hong, Joon Hwa; Jee, Se Hwan; Lee, Bong Sang; Kuk, Il Hyun

1994-06-01

This report reviewed the R and D states of reactor internal materials in order to be a reference for researches and engineers who are concerning on localization of the materials in the field or laboratory. General structure of PWR internals and material specification for YGN 3 and 4 were reviewed. States-of-arts on R and D of stainless steel and Alloy X-750 were reviewed, and degradation mechanisms of the components were analyzed. In order to develop the good domestic materials for reactor internal, following studies would be carried out: microstructure, sensitization behavior, fatigue property, irradiation-induced stress corrosion cracking/radiation-induced segregation, radiation embrittlement. (Author) 7 refs., 14 figs., 5 tabs.,

Atomic Physics 16: Sixteenth International Conference on Atomic Physics. Proceedings

International Nuclear Information System (INIS)

Baylis, W.E.; Drake, G.W.

1999-01-01

These proceedings represent papers presented at the 16th International Conference on Atomic Physics held in Windsor, Ontario, Canada, in August, 1998. The topics discussed included a wide array of subjects in atomic physics such as atom holography, alignment in atomic collisions, coulomb-interacting particles, muon experiments, x-rays from comets, atomic electron collisions in intense laser fields, spectroscopy of trapped ions, and Bose-Einstein condensates. This conference represents the single most important meeting world wide on fundamental advances in atomic physics. There were 30 papers presented at the conference,out of which 4 have been abstracted for the Energy, Science and Technology database

International benchmark on the natural convection test in Phenix reactor

International Nuclear Information System (INIS)

Tenchine, D.; Pialla, D.; Fanning, T.H.; Thomas, J.W.; Chellapandi, P.; Shvetsov, Y.; Maas, L.; Jeong, H.-Y.; Mikityuk, K.; Chenu, A.; Mochizuki, H.; Monti, S.

2013-01-01

Highlights: ► Phenix main characteristics, instrumentation and natural convection test are described. ► “Blind” calculations and post-test calculations from all the participants to the benchmark are compared to reactor data. ► Lessons learned from the natural convection test and the associated calculations are discussed. -- Abstract: The French Phenix sodium cooled fast reactor (SFR) started operation in 1973 and was stopped in 2009. Before the reactor was definitively shutdown, several final tests were planned and performed, including a natural convection test in the primary circuit. During this natural convection test, the heat rejection provided by the steam generators was disabled, followed several minutes later by reactor scram and coast-down of the primary pumps. The International Atomic Energy Agency (IAEA) launched a Coordinated Research Project (CRP) named “control rod withdrawal and sodium natural circulation tests performed during the Phenix end-of-life experiments”. The overall purpose of the CRP was to improve the Member States’ analytical capabilities in the field of SFR safety. An international benchmark on the natural convection test was organized with “blind” calculations in a first step, then “post-test” calculations and sensitivity studies compared with reactor measurements. Eight organizations from seven Member States took part in the benchmark: ANL (USA), CEA (France), IGCAR (India), IPPE (Russian Federation), IRSN (France), KAERI (Korea), PSI (Switzerland) and University of Fukui (Japan). Each organization performed computations and contributed to the analysis and global recommendations. This paper summarizes the findings of the CRP benchmark exercise associated with the Phenix natural convection test, including blind calculations, post-test calculations and comparisons with measured data. General comments and recommendations are pointed out to improve future simulations of natural convection in SFRs

Fifteenth annual meeting of the International Working Group on Fast Reactors. Summary report

International Nuclear Information System (INIS)

1982-09-01

The Fifteenth Annual Meeting of the IWGFR was held in accordance with the recommendation of the previous Annual Group Meeting, in Obninsk, USSR, Vienna from 30 March to 2 April 1982. The meeting was attended by the Member States of the group: France, the Federal Republic of Germany, Italy, Japan, the United Kingdom, and the USA, as well as by representatives from CEC, IAEA and OECD and observer from the USSR. This document includes: review of the IWGFR Activities for the period since the Eleventh Annual Meeting of the Group; preliminary programme of international conference on breeder reactors as a world energy resource and the breeder fuel cycle; list of meetings on atomic energy which may be of interest to the IWGFR Members; IWGFR criteria for supporting some of the international conferences; list of proposed topics for the IWGFR Specialists' Meetings; list of topics for review articles on LMFBR recommended for publication by the IAEA; list of meetings sponsored by the IWGFR; a list of members of the International Working Group on Fast Reactors

Sixteenth annual meeting of the International Working Group on Fast Reactors. Summary report

Energy Technology Data Exchange (ETDEWEB)

NONE

1983-10-01

The Sixteenth Annual Meeting of the IWGFR was held in accordance with the recommendation of the previous Annual Meeting Group, in Vienna from 12-15 April 1983. The meeting was attended by the Member States of the group: France, the Federal Republic of Germany, Italy, Japan, the United Kingdom, and the USA, as well as by representatives from CEC, IAEA and OECD and observer from the USSR. This document includes: review of the IWGFR Activities for the period since the Eleventh Annual Meeting of the Group; preliminary programme of international conference on breeder reactors as a world energy resource and the breeder fuel cycle; list of meetings on atomic energy which may be of interest to the IWGFR Members; IWGFR criteria for supporting some of the international conferences; list of proposed topics for the IWGFR Specialists' Meetings; list of topics for review articles on LMFBR recommended for publication by the IAEA; list of meetings sponsored by the IWGFR; a list of members of the International Working Group on Fast Reactors.

Fifteenth annual meeting of the International Working Group on Fast Reactors. Summary report

Energy Technology Data Exchange (ETDEWEB)

NONE

1982-09-01

The Fifteenth Annual Meeting of the IWGFR was held in accordance with the recommendation of the previous Annual Group Meeting, in Obninsk, USSR, Vienna from 30 March to 2 April 1982. The meeting was attended by the Member States of the group: France, the Federal Republic of Germany, Italy, Japan, the United Kingdom, and the USA, as well as by representatives from CEC, IAEA and OECD and observer from the USSR. This document includes: review of the IWGFR Activities for the period since the Eleventh Annual Meeting of the Group; preliminary programme of international conference on breeder reactors as a world energy resource and the breeder fuel cycle; list of meetings on atomic energy which may be of interest to the IWGFR Members; IWGFR criteria for supporting some of the international conferences; list of proposed topics for the IWGFR Specialists' Meetings; list of topics for review articles on LMFBR recommended for publication by the IAEA; list of meetings sponsored by the IWGFR; a list of members of the International Working Group on Fast Reactors.

Sixteenth annual meeting of the International Working Group on Fast Reactors. Summary report

International Nuclear Information System (INIS)

1983-10-01

The Sixteenth Annual Meeting of the IWGFR was held in accordance with the recommendation of the previous Annual Meeting Group, in Vienna from 12-15 April 1983. The meeting was attended by the Member States of the group: France, the Federal Republic of Germany, Italy, Japan, the United Kingdom, and the USA, as well as by representatives from CEC, IAEA and OECD and observer from the USSR. This document includes: review of the IWGFR Activities for the period since the Eleventh Annual Meeting of the Group; preliminary programme of international conference on breeder reactors as a world energy resource and the breeder fuel cycle; list of meetings on atomic energy which may be of interest to the IWGFR Members; IWGFR criteria for supporting some of the international conferences; list of proposed topics for the IWGFR Specialists' Meetings; list of topics for review articles on LMFBR recommended for publication by the IAEA; list of meetings sponsored by the IWGFR; a list of members of the International Working Group on Fast Reactors

Assessment of the requirements for placing and maintaining Savannah River Site spent fuel storage basins under International Atomic Energy Agency safeguards

International Nuclear Information System (INIS)

Amacker, O.P. Jr.; Curtis, M.M.; Delegard, C.H.; Hsue, S.T.; Whitesel, R.N.

1997-03-01

The United States is considering the offer of irradiated research reactor spent fuel (RRSF) for international safeguards applied by the International Atomic Energy Agency (IAEA). The offer would be to add one or more spent fuel storage basins to the list of facilities eligible for IAEA safeguards. The fuel to be safeguarded would be stored in basins on the Savannah River Site (SRS). This RRSF potentially can include returns of Material Test Reactor (MTR) VAX fuel from Argentina, Brazil, and Chile (ABC); returns from other foreign research reactors; and fuel from domestic research reactors. Basins on the SRS being considered for this fuel storage are the Receiving Basin for Offsite Fuel (RBOF) and the L-Area Disassembly Basin (L-Basin). A working group of SRS, U.S. Department of Energy International Safeguards Division (NN-44), and National Laboratory personnel with experience in IAEA safeguards was convened to consider the requirements for applying the safeguards to this material. The working group projected the safeguards requirements and described alternatives

Internal exposure profile of occupational workers of a BWR type atomic power station

International Nuclear Information System (INIS)

Hegde, A.G.; Bhat, I.S.

1979-01-01

The internal exposure profile of major radionuclides, for Tarapur Atomic Power Station (India) occupational staff for the last 9 years (1970-1978) of station operation, is presented. This power station has two boiling water reactor units. The occupational staff were monitored for internal exposure with the whole body counter. It has been observed that 60 Co, 134 Cs and 137 Cs are major contaminants. The highest yearly average of internal exposure was less than 1% of maximum permissible body burden recommended by ICRP. Depending on the nature of exposures the power station employees were classified under four different groups, (i) maintenance, (ii) operations, (iii) techanical and (iv) non-technical. This study revealed that maintenance group had highest incidence of internal exposure among these. It is also observed that contribution of 60 Co is maximum in the exposure of this group. (B.G.W.)

International Atomic Energy Agency (IAEA) Activity on Technical Influence of High Burnup UOX and MOX Water Reactor Fuel on Spent Fuel Management

International Nuclear Information System (INIS)

Lovasic, Z.; Einziger, R.

2009-01-01

This paper briefly reviews the results of the International Atomic Energy Agency (IAEA) project investigating the influence of high burnup and mixed-oxide (MOX) fuels, from water power reactors, on spent fuel management. These data will provide information on the impacts, regarding spent fuel management, for those countries operating light-water reactors (LWR)s and heavy-water reactors (HWR)s with zirconium alloy-clad uranium dioxide (UOX) fuels, that are considering the use of higher burnup UOX or the introduction of reprocessing and MOX fuels. The mechanical designs of lower burnup UOX and higher burnup UOX or MOX fuel are very similar, but some of the properties (e.g., higher fuel rod internal pressures; higher decay heat; higher specific activity; and degraded cladding mechanical properties of higher burnup UOX and MOX spent fuels) may potentially significantly affect the behavior of the fuel after irradiation. These properties are reviewed. The effects of these property changes on wet and dry storage, transportation, reprocessing, re-fabrication of fuel, and final disposal were evaluated, based on regulatory, safety, and operational considerations. Political and strategic considerations were not taken into account since relative importance of technical, economic and strategic considerations vary from country to country. There will also be an impact of these fuels on issues like non-proliferation, safeguards, and sustainability, but because of the complexity of factors affecting those issues, they are only briefly discussed. Data gaps were also identified during this investigation. The pros and cons of using high burnup UOX or MOX, for each applicable issue in each stage of the back end of the fuel cycle, were evaluated and are discussed.. Although, in theory, higher burnup fuel and MOX fuels mean a smaller quantity of spent fuel, the potential need for some changes in design of spent fuel storage, transportation, handling, reprocessing, re-fabrication, and

An internally illuminated monolith reactor: Pros and cons relative to a slurry reactor

NARCIS (Netherlands)

Carneiro, Joana T.; Carneiro, J.T.; Berger, Rob; Moulijn, Jacob A.; Mul, Guido

2009-01-01

In the present study, kinetic models for the photo-oxidation of cyclohexane in two different photoreactor systems are discussed: a top illumination reactor (TIR) representative of a slurry reactor, and the so-called internally illuminated monolith reactor (IIMR) representing a reactor containing

Accelerators and nuclear reactors as tools in hot atom chemistry

International Nuclear Information System (INIS)

Lindner, L.

1975-01-01

The characteristics of accelerators and of nuclear reactors - the latter to a lesser extent - are discussed in view of their present and future use in hot atom chemistry research and its applications. (author)

Reactor Structure Materials: Corrosion of Reactor Core Internals

International Nuclear Information System (INIS)

Van Dyck, S.

2000-01-01

The objectives of SCK-CEN's R and D programme on the corrosion of reactor core internals are: (1) to gain mechanistic insight into the Irradition Assisted Stress Corrosion Cracking (IASCC) phenomenon by studying the influence of separate parameters in well controlled experiments; (2) to develop and validate a predictive capability on IASCC by model description and (3) to define and validate countermeasures and monitoring techniques for application in reactors. Progress and achievements in 1999 are described

Reliability tests for reactor internals rejuvenation technology

International Nuclear Information System (INIS)

Fujimaki, Katsumi; Hitoki, Yoichi; Otsubo, Toru; Uchiyama, Junichi

1998-01-01

Structural damage due to aging degradation of LWR reactor internals has been reported in several nuclear plants. NUPEC has started a project to test the reliability of the technology for rejuvenating reactor internals which has been funded by the Ministry of International Trade and Industry (MITI) of Japan since 1995. The project follows the policy of a report that the MITI has formally issued in April 1996 summarizing the countermeasures to be considered for aging nuclear plants and equipment. This paper gives an outline of the test plans and results which are directed at preventive maintenance before damage and repair after damage for reactor internals aging degradation. The test results for the replacement methods of ICM housing and BWR core shroud have shown that the methods were reliable and the structural integrity was appropriate based on the evaluation. (author)

Nuclear data for reactors. Proceedings of the second international conference. Vol. II

International Nuclear Information System (INIS)

1970-01-01

The Second International Conference on Nuclear Data for Reactors, held in Helsinki at the invitation of the Finnish Government, was convened by the International Atomic Energy Agency from 15 to 19 June 1970. The Conference, held as a result of recommendations made by the International Nuclear Data Committee, was attended by 163 participants from 28 countries and four international organizations, and 21 invited and 98 contributed papers were presented. This Conference was the second held by the IAEA on Nuclear Data for Reactors. Almost four years have elapsed since the first was held in Paris in 1966. During these years gratifying progress has been made by reactor, nuclear and evaluation physicists, whose collaboration has been greatly enhanced. As a result, many laboratories have concentrated their efforts on items of particular importance for reactor research and development, and many measurements are now available. The main purpose of this Conference was to provide an opportunity to review results of recent basic neutron-physics investigations against a background need for basic information, especially concerning reactors. The Conference itself, together with the preparatory meetings of IAEA experts in Studsvik on the status of α( 239 Pu) and the ν-bar-values for fissionable nuclei, showed an emphasis on the nuclear data aspects most important for nuclear technology. Most contributors dealt with the measurement and analysis of neutron cross-sections. This extensive new cross-section information can be attributed to several factors, the most important being the development and systematic exploitation of high-intensity neutron sources, such as modern linear accelerators, modern cyclotrons and underground nuclear explosions, improvements in instrumentation and in sample preparation techniques, and other technical improvements. Compared with the first IAEA Conference on Nuclear Data for Reactors this one has many more contributions on neutron data evaluation. Many

Nuclear data for reactors. Proceedings of the second international conference. Vol. I

International Nuclear Information System (INIS)

1970-01-01

The Second International Conference on Nuclear Data for Reactors, held in Helsinki at the invitation of the Finnish Government, was convened by the International Atomic Energy Agency from 15 to 19 June 1970. The Conference, held as a result of recommendations made by the International Nuclear Data Committee, was attended by 163 participants from 28 countries and four international organizations, and 21 invited and 98 contributed papers were presented. This Conference was the second held by the IAEA on Nuclear Data for Reactors. Almost four years have elapsed since the first was held in Paris in 1966. During these years gratifying progress has been made by reactor, nuclear and evaluation physicists, whose collaboration has been greatly enhanced. As a result, many laboratories have concentrated their efforts on items of particular importance for reactor research and development, and many measurements are now available. The main purpose of this Conference was to provide an opportunity to review results of recent basic neutron-physics investigations against a background need for basic information, especially concerning reactors. The Conference itself, together with the preparatory meetings of IAEA experts in Studsvik on the status of α( 239 Pu) and the ν-bar-values for fissionable nuclei, showed an emphasis on the nuclear data aspects most important for nuclear technology. Most contributors dealt with the measurement and analysis of neutron cross-sections. This extensive new cross-section information can be attributed to several factors, the most important being the development and systematic exploitation of high-intensity neutron sources, such as modern linear accelerators, modern cyclotrons and underground nuclear explosions, improvements in instrumentation and in sample preparation techniques, and other technical improvements. Compared with the first IAEA Conference on Nuclear Data for Reactors this one has many more contributions on neutron data evaluation. Many

Reliability tests for reactor internals replacement technology

International Nuclear Information System (INIS)

Fujimaki, K.; Uchiyama, J.; Ohtsubo, T.

2000-01-01

Structural damage due to aging degradation of LWR reactor internals has been reported in several nuclear plants. NUPEC has started a project to test the reliability of the technology for replacing reactor internals, which was directed at preventive maintenance before damage and repair after damage for the aging degradation. The project has been funded by the Ministry of International Trade and Industry (MITI) of Japan since 1995, and it follows the policy of a report that the MITI has formally issued in April 1996 summarizing the countermeasures to be considered for aging nuclear plants and equipment. This paper gives an outline of the whole test plans and the test results for the BWR reactor internals replacement methods; core shroud, ICM housing, and CRD Housing and stub tube. The test results have shown that the methods were reliable and the structural integrity was appropriate based on the evaluation. (author)

Agreement among the Portuguese Republic, the Government of the United States of America and the International Atomic Energy Agency for assistance in securing nuclear fuel for a research reactor

International Nuclear Information System (INIS)

2007-01-01

The text of the Agreement among the Portuguese Republic, the Government of the United States of America and the International Atomic Energy Agency for Assistance in Securing Nuclear Fuel for a Research Reactor is reproduced in this document for the information of all Members of the Agency. The Agency's Board of Governors approved the above mentioned Agreement on 14 June 2006. The Agreement was signed by the authorized representatives of Portugal on 27 June 2006 and the United States on 13 December 2006, and by the Director General of the IAEA on 14 December 2006. Pursuant to the Article XII.1 of the Agreement, the Agreement entered into force on 19 April 2007, the date on which the Agency received written notification from Portugal that its internal requirements for entry into force had been met

The law for the regulations of nuclear source materials, nuclear fuel materials and reactors

International Nuclear Information System (INIS)

1980-01-01

The law intends under the principles of the atomic energy act to regulate the refining, processing and reprocessing businesses of nuclear raw and fuel metarials and the installation and operation of reactors for the peaceful and systematic utilization of such materials and reactors and for securing public safety by preventing disasters, as well as to control internationally regulated things for effecting the international agreements on the research, development and utilization of atomic energy. Basic terms are defined, such as atomic energy; nuclear fuel material; nuclear raw material; nuclear reactor; refining; processing; reprocessing; internationally regulated thing. Any person who is going to engage in refining businesses other than the Power Reactor and Nuclear Fuel Development Corporation shall get the special designation by the Prime Minister and the Minister of International Trade Industry. Any person who is going to engage in processing businesses shall get the particular admission of the Prime Minister. Any person who is going to establish reactors shall get the particular admission of the Prime Minister, The Minister of International Trade and Industry or the Minister of Transportation according to the kinds of specified reactors, respectively. Any person who is going to engage in reprocessing businesses other than the Power Reactor and Nuclear Fuel Development Corporation and the Japan Atomic Energy Research Institute shall get the special designation by the Prime Minister. The employment of nuclear fuel materials and internationally regulated things is defined in detail. (Okada, K.)

Thai research reactor

International Nuclear Information System (INIS)

Aramrattana, M.

1987-01-01

The Office of Atomic Energy for Peace (OAEP) was established in 1962, as a reactor center, by the virtue of the Atomic Energy for Peace Act, under operational policy and authority of the Thai Atomic Energy for Peace Commission (TAEPC); and under administration of Ministry of Science, Technology and Energy. It owns and operates the only Thai Research Reactor (TRR-1/M1). The TRR-1/M1 is a mixed reactor system constituting of the old MTR type swimming pool, irradiation facilities and cooling system; and TRIGA Mark III core and control instrumentation. The general performance of TRR-1/M1 is summarized in Table I. The safe operation of TRR-1/M1 is regulated by Reactor Safety Committee (RSC), established under TAEPC, and Health Physics Group of OAEP. The RCS has responsibility and duty to review of and make recommendations on Reactor Standing Orders, Reactor Operation Procedures, Reactor Core Loading and Requests for Reactor Experiments. In addition,there also exist of Emergency Procedures which is administered by OAEP. The Reactor Operation Procedures constitute of reactor operating procedures, system operating procedures and reactor maintenance procedures. At the level of reactor routine operating procedures, there is a set of Specifications on Safety and Operation Limits and Code of Practice from which reactor shift supervisor and operators must follow in order to assure the safe operation of TRR-1/M1. Table II is the summary of such specifications. The OAEP is now upgrading certain major components of the TRR-1/M1 such as the cooling system, the ventilation system and monitoring equipment to ensure their adequately safe and reliable performance under normal and emergency conditions. Furthermore, the International Atomic Energy Agency has been providing assistance in areas of operation and maintenance and safety analysis. (author)

1989 basic plan for atomic energy development and utilization

International Nuclear Information System (INIS)

1989-01-01

A Basic Plan for Atomic Energy Development and Utilization has been established each year based on the guidelines set up by the Atomic Energy Commission of Japan, with the aim of promoting the development and utilization of atomic energy schematically and efficiently. The Basic Plan shows specific projects to achieve the objectives specified in the Long-Range Plan for Atomic Energy Development and Utilization. The Basic Plan specifies efforts to be made for overall strengthening of safety measures (safety policies, safety research, disaster prevention, etc.), promotion of nuclear power generation, establishment of the nuclear fuel cycle (securing of uranium, technology for uranium enrichment, reprocessing, etc.), development of new types of power reactors (fast breeder reactor, new types of converter reactors, plutonium fuel processing technology), promotion of leading projects (nuclear fusion, utilization of radiations, atomic powered ships, high-temperature engineering tests), promotion of basic technology development (basic research, training of scientists and engineers), voluntary and active international activities (international cooperation), and acquisition of understanding and cooperation of the general public. (N,K.)

Proceedings of the 4th international symposium on material testing reactors

International Nuclear Information System (INIS)

Ishihara, Masahiro; Suzuki, Masahide

2012-03-01

This report is the Proceedings of the fourth International Symposium on Material Testing Reactors hosted by Japan Atomic Energy Agency (JAEA). The first symposium was held on 2008, at the Oarai Research and Development Center of JAEA, the second, 2009, Idaho National Laboratory (INL) of United States and the third 2010, Nuclear Research Institute (NRI) in Czech Republic to exchange information for deep mutual understanding of material testing reactors. The fourth symposium was originally scheduled to be held INVAP in Argentina. However, the aftermath of volcanic explosion at Chili forced the symposium to change place. Total 111 participants attended from Argentina, Belgium, France, Germany, Indonesia, Malasia, Korea, South Africa, Switzerland, the United State and Japan. This symposium addressed the general topics of 'status and future plan of material testing reactors', 'advancement of irradiation technology', 'expansion of industry use(RI)', 'facility, upgrade, aging management', 'new generation MTR', 'advancement of PIE technology', 'development of advanced driver fuel', and 'nuclear human resource development(HRD) for next generation', and 39 presentations were made. Furthermore, three topics, 'Necessity of cooperation for Mo-99 production by (n,gamma) reaction', 'Necessity of standardization of irradiation technology' and 'Conceptual design of next generation materials testing reactor by collaboration', were selected and discussed. (author)

Proceedings of the 4th international symposium on material testing reactors

Energy Technology Data Exchange (ETDEWEB)

Ishihara, Masahiro; Suzuki, Masahide [Japan Atomic Energy Agency, Oarai Research and Development Center, Oarai, Ibaraki (Japan)

2012-03-15

This report is the Proceedings of the fourth International Symposium on Material Testing Reactors hosted by Japan Atomic Energy Agency (JAEA). The first symposium was held on 2008, at the Oarai Research and Development Center of JAEA, the second, 2009, Idaho National Laboratory (INL) of United States and the third 2010, Nuclear Research Institute (NRI) in Czech Republic to exchange information for deep mutual understanding of material testing reactors. The fourth symposium was originally scheduled to be held INVAP in Argentina. However, the aftermath of volcanic explosion at Chili forced the symposium to change place. Total 111 participants attended from Argentina, Belgium, France, Germany, Indonesia, Malasia, Korea, South Africa, Switzerland, the United State and Japan. This symposium addressed the general topics of 'status and future plan of material testing reactors', 'advancement of irradiation technology', 'expansion of industry use(RI)', 'facility, upgrade, aging management', 'new generation MTR', 'advancement of PIE technology', 'development of advanced driver fuel', and 'nuclear human resource development(HRD) for next generation', and 39 presentations were made. Furthermore, three topics, 'Necessity of cooperation for Mo-99 production by (n,gamma) reaction', 'Necessity of standardization of irradiation technology' and 'Conceptual design of next generation materials testing reactor by collaboration', were selected and discussed. (author)

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

International trend on development of an innovative nuclear reactor and its meanings

Energy Technology Data Exchange (ETDEWEB)

Matsui, Kazuaki [Institute of Applied Energy, Tokyo (Japan)

2002-01-01

On outlining on flow of so-called innovative or new type nuclear reactor, at first, an improvement line of large-scale WHR, contains ABWR-2, APWR and its successive APWR+ in Japan, APR in Korea, and EPR in Europe, all of which have super large-scale output of 1.5MKW to use their scale merits in maximum. And, the second type is fast reactor only in Russia and Japan which are under reviewing its actual using plan of its already established development route. Furthermore, nuclear industry in the world is allowable to say a has-been industry, even its R and D system is decrepit, its researchers are much aged, and even utilization and foreign development of nuclear energy as a protecting measure of global warming are pronounced its self-control at the Bonn Conference in last year. However, the Generation 4 International Forum led by U.S.A. since early of 2000 and the Innovative Reactor Development Program (INPRO) through the International Atomic Energy Association (IAEA) due to initiative of Russia are planned to cooperatively promote their programs. In order to obtain any priority on small-scale production considerable technical jump is required or R and D and technical development elements with technical gap is necessary, which must be proved establishment of a target to overcome their scale demerit. (G.K.)

Agreement of 27 February 1992 between the government of the People's Democratic Republic of Algeria and the International Atomic Energy Agency for the application of safeguards in connection with the supply of a research reactor from the People's Republic of China

International Nuclear Information System (INIS)

1992-07-01

The document informs that the Agreement between the Government of the People's Democratic Republic of Algeria and the International Atomic Energy Agency for the Application of Safeguards in connection with the Supply of a Research Reactor from the People's Republic of China, which has been applied provisionally as from 27 February 1992, entered into force on 2 June 1992

Lifetime assessment on PWR reactor vessel internals in Korea

International Nuclear Information System (INIS)

Jung, Sung-Gyu; Jin, Tae-Eun; Jeong, Ill-Seok

2002-01-01

In order to extend the operating time of the Kori Unit 1 reactor internals, a comprehensive review of the potential ageing problems and a safety assessment have been performed. As the plant ages, reactor internal components which are subject to various ageing mechanism should be identified and evaluated based on the systematic technical procedure. In this respect, technical procedure for lifetime evaluation had been developed and applied to reactor internals. This paper describes a overall assessment and ageing management procedure and evaluation results for reactor internals. Also this paper suggests the optimal ageing management programs to maintain the integrity of reactor internals beyond design life based on the evaluation results. A review of all known potential ageing mechanisms was performed for each of the reactor internal subcomponents. From these results, 8 ageing mechanisms such as void swelling, irradiation and thermal embrittlement, fatigue, stress corrosion cracking, IASCC, stress relaxation, and wear for the reactor internal components were expected to be of major concerns during the current or extended plant life. In this study, 8 ageing mechanisms were identified for lifetime evaluation. For these ageing mechanisms, lifetime assessment was performed. As a result of this evaluation, it is expected that core barrel will exceed the IASCC threshold value during 40 operating years, and baffle/former and baffle former bolts will exceed the threshold value for void swelling, irradiation embrittlement, IASCC, stress relaxation during 40 operating years. However, for all other reactor internals subcomponents, thermal embrittlement, fatigue, SCC, and wear were identified as nonsignificant. As a result of lifetime evaluations, 4 ageing mechanisms were established to be plausible for 3 subcomponents. These results are shown. The existing ageing management programs (AMPs) for Kori Unit 1, such as ISI, water chemistry control, rod drop time testing etc., were

Atomization of U3Si2 for research reactor fuel

International Nuclear Information System (INIS)

Kim, C.K.; Kim, K.H.; Lee, C.T.; Kuk, I.H.

1995-01-01

Rotating disk atomization technique is applied to KMRR (Korea Multi-purpose Research Reactor) fuel fabrication. A rotating disk atomizer is designed and manufactured locally and U-4.0 wt. % Si alloy powders are produced. The atomized powders are heat-treated to transform into U 3 Si and the mixture of U 3 Si and Al are extruded to fuel meat. Most of the atomized powders are spherical in shape. The microstructure of the powder is fine due to the rapid solidification. The time required for peritectoid reaction is reduced due to the fine microstructures and the resultant U 3 Si grain size is finer than ever obtained from ingot process. The mechanical properties of the fuel meat are improved: yield strength about 30 %, tensile strength 10% and elongation 250 % increased. (author)

Twelfth annual meeting of the International Working Group on Fast Reactors. Summary report. Part I

Energy Technology Data Exchange (ETDEWEB)

NONE

1979-05-01

The Twelfth Annual Meeting of the IWGFR was held in accordance with the recommendation of the previous AGM,in Vienna from 27 to 30 March 1979. The meeting was attended by the Member States of the group: France, the Federal Republic of Germany, Italy, Japan, the United Kingdom, and the USA, as well as by representatives from CEC, IAEA and OECD and observer from the USSR. This document includes: review of the IWGFR Activities for the period since the Eleventh Annual Meeting of the Group; preliminary programme of international conference on breeder reactors as a world energy resource and the breeder fuel cycle; list of meetings on atomic energy which may be of interest to the IWGFR Members; IWGFR criteria for supporting some of the international conferences; list of proposed topics for the IWGFR Specialists' Meetings; list of topics for review articles on LMFBR recommended for publication by the IAEA; list of meetings sponsored by the IWGFR; a list of members of the International Working Group on Fast Reactors.

Twelfth annual meeting of the International Working Group on Fast Reactors. Summary report. Part I

International Nuclear Information System (INIS)

1979-05-01

The Twelfth Annual Meeting of the IWGFR was held in accordance with the recommendation of the previous AGM,in Vienna from 27 to 30 March 1979. The meeting was attended by the Member States of the group: France, the Federal Republic of Germany, Italy, Japan, the United Kingdom, and the USA, as well as by representatives from CEC, IAEA and OECD and observer from the USSR. This document includes: review of the IWGFR Activities for the period since the Eleventh Annual Meeting of the Group; preliminary programme of international conference on breeder reactors as a world energy resource and the breeder fuel cycle; list of meetings on atomic energy which may be of interest to the IWGFR Members; IWGFR criteria for supporting some of the international conferences; list of proposed topics for the IWGFR Specialists' Meetings; list of topics for review articles on LMFBR recommended for publication by the IAEA; list of meetings sponsored by the IWGFR; a list of members of the International Working Group on Fast Reactors

International Thermonuclear Experimental Reactor: Physics issues, capabilities and physics program plans

International Nuclear Information System (INIS)

Wesley, J.C.

1997-01-01

Present status and understanding of the principal plasma-performance determining physics issues that affect the physics design and operational capabilities of the International Thermonuclear Experimental Reactor (ITER) [ITER EDA Agreement and Protocol 2 (International Atomic Energy Agency, Vienna, 1994)] are presented. Emphasis is placed on the five major physics-basis issues emdash energy confinement, beta limit, density limit, impurity dilution and radiation loss, and the feasibility of obtaining partial-detached divertor operation emdash that directly affect projections of ITER fusion power and burn duration performance. A summary of these projections is presented and the effect of uncertainties in the physics-basis issues is examined. ITER capabilities for experimental flexibility and plasma-performance optimization are also described, and how these capabilities may enter into the ITER physics program plan is discussed. copyright 1997 American Institute of Physics

Numerical analysis of reactor internals under hydrodynamic loads

Energy Technology Data Exchange (ETDEWEB)

Kim, Da Hye; Chang, Yoon Suk [Kyung Hee Univ., Yongin (Korea, Republic of); Jhung, Myung Jo [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2013-10-15

In the present study, six kinds of major equipment of a typical reactor internals were identified by incorporating recent research trend. Based on this, detailed numerical models were developed and used for establishment of optimum analysis methodology subjected to hydrodynamic loads. As a result, stress values of the major equipment were calculated through the acoustic-structure analysis under periodic hydrodynamic load and the turbulence-structure analysis under random hydrodynamic load. The numerical analysis scheme can be used for development of preventive action plan and management procedures of the reactor internals. Reactor internals installed in a pressure vessel have been exposed to harsh environment such as high neutron irradiation and temperature with complex fluid flow. As the increase of operational years of NPPs(Nuclear Power Plants), possibility of functional loss of the reactor internals is increased due to degradation caused by radiation embrittlement, thermal aging, fatigue, corrosion and FIV(Flow-Induced Vibration) etc. In practice, defects were detected at core support structure as well as upper and lower parts of structural assembly in European and United States NPPs. Recently, in a GALL(Generic Aging Lessons Learned) report, US NRC(Nuclear Regulatory Commission) identified reactor internals as a high priority component and addressed relevant management programs. In Korea, similar activities have been conducted for long-term operation beyond design lifetime but most of them were limited to qualitative evaluation based on examination and maintenance programs. Therefore, not only to reduce repair and replacement efforts but also to secure the stability of NPPs, necessity for development of quantitative evaluation technique as well as establishment of preventive action plan and management procedures is on the rise. The FIV represents the structural vibration phenomenon induced by liquid flow and generally occurs at contact surfaces. In the present

International review on safety requirements for the prototype fast breeder reactor “Monju”

International Nuclear Information System (INIS)

2016-01-01

In response to the lessons learned from the serious nuclear accidents at the TEPCO's Fukushima Daiichi Nuclear Power Stations, an advisory committee, which was set up by the Japan Atomic Energy Agency, issued the report “Safety Requirements Expected to the Prototype Fast Breeder Reactor Monju” taking into account the SFR specific safety characteristics in July 2014. The report was reviewed by the leading international experts on SFR safety from five countries and one international organization in order to obtain independent and objective evaluation. The international review comments on each subsection were collected and compiled, and then a summary of results was derived through the discussion at the review meeting and individual feedbacks. As a result the basic concept for prevention of severe accidents and mitigation of their consequences of Monju is appropriate in consideration of SFR specific safety characteristics, and is in accordance with international common understanding. (author)

Aging considerations for PWR [pressurized water reactor] control rod drive mechanisms and reactor internals

International Nuclear Information System (INIS)

Ware, A.G.

1988-01-01

This paper describes age-related degradation mechanisms affecting life extension of pressurized water reactor control rod drive mechanisms and reactor internals. The major sources of age-related degradation for control rod drive mechanisms are thermal transients such as plant heatups and cooldowns, latchings and unlatchings, long-term aging effects on electrical insulation, and the high temperature corrosive environment. Flow induced loads, the high-temperature corrosive environment, radiation exposure, and high tensile stresses in bolts all contribute to aging related degradation of reactor internals. Another problem has been wear and fretting of instrument guide tubes. The paper also discusses age-related failures that have occurred to date in pressurized water reactors

Atoms: for war or peace

Energy Technology Data Exchange (ETDEWEB)

Subrahmanyam, K V

1981-08-01

History of nuclear power generation starting from the experimental split of uranium atom in 1938 to the establishment of the International Atomic Energy Agency is traced. In India, the Atomic Energy Commission was established with the major objective of developing nuclear power to make up India's deficiencies in energy sources. It is noted that from the very beginning the commission's activities were covered under a blanket of secrecy. According to the author, India's atomic energy programme stagnated after Dr. Bhabha's death. The Department of Atomic Energy diverted its attention to the nuclear explosion which was carried out in 1974. This event caused a great setback to the collaboration with Canada and USA in the nuclear power programme. The resulting problems are still not fully solved. The author maintains that the Department of Atomic Energy should have confined its efforts to the reactor development with special reference to the fast breeder reactor so that thorium can be utilised to the maximum advantage.

The international thermonuclear reactor project

International Nuclear Information System (INIS)

James, T.R.

1993-01-01

The International Thermonuclear Experimental Reactor Project is a 6-year collaborative effort involving the U.S., Europe, Japan, and the Russian Federation to produce a detailed engineering design for the next-step fusion device

International cooperation in nuclear safety

International Nuclear Information System (INIS)

Rosen, M.

1991-01-01

The mechanisms of international co-operations, co-ordinated by International Atomic Energy Agency, are presented. These co-operations are related to international safety standards, to the safety of the four hundred existing reactors in operation, to quick help and information in case of emergency, and to the already valid international conventions. The relation between atomic energy and environmental protection is also discussed briefly. (K.A.)

State of the art of toshiba maintenance techniques for reactor internals

International Nuclear Information System (INIS)

Maekawa, Osamu; Hattori, Yasuhiro; Sudo, Akira

2002-01-01

As the number of aged plants increases, maintaining the integrity of the reactor pressure vessel and reactor internals in aged plants has become an essential issue to ensure continued stable operation and achieve higher plant operability. A major issue with regard to reactor internals is stress corrosion cracks (SCCs). Laser-applying techniques have many features suitable for preventive maintenance work on reactor internals. Toshiba has developed various laser-applying preventive maintenance techniques and accumulated considerable field experience utilizing these techniques in various aged plants. Moreover, in view of the importance of confirming the soundness of reactor internals in aged plants, Toshiba has developed and applied sophisticated nondestructive testing techniques for this purpose. (author)

International topical meeting. Research Reactor Fuel Management (RRFM) and meeting of the International Group on Reactor Research (IGORR)

Energy Technology Data Exchange (ETDEWEB)

NONE

2007-07-01

Nuclear research and test reactors have been in operation for over 60 years, over 270 research reactors are currently operating in more than 50 countries. This meeting is dedicated to different aspects of research reactor fuels: new fuels for new reactors, the conversion to low enriched uranium fuels, spent fuel management and computational tools for core simulation. About 80 contributions are reported in this document, they are organized into 7 sessions: 1) international topics and overview on new projects and fuel, 2) new projects and upgrades, 3) fuel development, 4) optimisation and research reactor utilisation, 5) innovative methods in research reactors physics, 6) safety, operation and research reactor conversion, 7) fuel back-end management, and a poster session. Experience from Australian, Romanian, Libyan, Syrian, Vietnamese, South-African and Ghana research reactors are reported among other things. The Russian program for research reactor spent fuel management is described and the status of the American-driven program for the conversion to low enriched uranium fuels is presented. (A.C.)

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

International Nuclear Information System (INIS)

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

2006-02-01

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

ITER [International Thermonuclear Experimental Reactor] reactor building design study

International Nuclear Information System (INIS)

Thomson, S.L.; Blevins, J.D.; Delisle, M.W.

1989-01-01

The International Thermonuclear Experimental Reactor (ITER) is at the midpoint of a two-year conceptual design. The ITER reactor building is a reinforced concrete structure that houses the tokamak and associated equipment and systems and forms a barrier between the tokamak and the external environment. It provides radiation shielding and controls the release of radioactive materials to the environment during both routine operations and accidents. The building protects the tokamak from external events, such as earthquakes or aircraft strikes. The reactor building requirements have been developed from the component designs and the preliminary safety analysis. The equipment requirements, tritium confinement, and biological shielding have been studied. The building design in progress requires continuous iteraction with the component and system designs and with the safety analysis. 8 figs

International bulletin on atomic and molecular data for fusion

International Nuclear Information System (INIS)

Stephens, J.A.; Bannister, M.E.; Fuhr, J.

1999-12-01

The International Bulletin on Atomic and Molecular Data for Fusion is prepared by the Atomic and Molecular Data Unit of the International Atomic Energy Agency. It is distributed free of charge by the IAEA to assist in the development of fusion research and technology. In part 1, the Atomic and Molecular Data Information System (AMDIS) is presented. In Part 2, the indexed papers are listed separately for structure and spectra, atomic and molecular collisions and surface interactions. Part 3 contains all the bibliographic data for both the indexed and non-indexed references. Finally, the Author Index (part 4) refers to the bibliographic references contained in part 3

International cooperation in accident analysis of RBMK reactors

International Nuclear Information System (INIS)

Kaliatka, A.; Isag

2005-01-01

safety programme reflected International Atomic Energy Agency safety recommendations. The Specific Objective of this project is to provide the Russian Authorities with a detailed code system able to reliably evaluate the core behavior (and all the associated safety outcomes) during severe transients and accidents, including extensive fuel melting, individual fuel channel rupture and then multiple channel ruptures. This code system should enable experts to evaluate the safety of core designs, and to give advice how to lessen-and mitigate- the consequences of core severe transients and accidents. The use of this code system shall lead to a safety improvement, which should be reflected in Russian and international safety review reports. The beneficiary in this project is the Russian Authorities 'ROSENERGOATOM' and 'GOSATOMNADZOR', the contractor - University of Pisa. The sub-contractor - Russian research institute, designer of RBMK - MINATOM/NIKIET. The key experts for this project are involved from different countries and institutions: LEI and Ignalina NPP (Lithuania), IAEA, PSU (USA), FZK (Germany), ITER (Italy) and other. The scope of the second project (PHARE projects) is to enhance the experience and capabilities and to establish a powerful infrastructure to Lithuanian State Nuclear Power Safety Inspectorate (VATESI) for following purposes: 1) Development of requirements for the Equipment Qualification Programme at Ignalina NPP, 2) Development of a regulatory guide on implementation of requirement for Ignalina NPP accident analysis, 3) Development of governing procedures for the operation of VATESI Emergency Centre accident analysis group, 4) Development of requirements on beyond design basis accident assessment and management of RBMK-1500 reactors, The beneficiary in this project was the Lithuanian State Nuclear Power Safety Inspectorate (VATESI) and its technical support organizations. The contractor was RISKAUDIT; which formed an EU project team with GRS, IRSN, SIP

Advanced robotic remote handling system for reactor dismantlement

International Nuclear Information System (INIS)

Shinohara, Yoshikuni; Usui, Hozumi; Fujii, Yoshio

1991-01-01

An advanced robotic remote handling system equipped with a multi-functional amphibious manipulator has been developed and used to dismantle a portion of radioactive reactor internals of an experimental boiling water reactor in the program of reactor decommissioning technology development carried out by the Japan Atomic Energy Research Institute. (author)

Intergovernmental organisation activities: European Atomic Energy Community, International Atomic Energy Agency, OECD Nuclear Energy Agency

International Nuclear Information System (INIS)

Anon.

2012-01-01

European Atomic Energy Community: Proposed legislative instruments, Adopted legislative instruments, Non-legislative instruments, Other activities (meetings). International Atomic Energy Agency: IAEA Action Plan on Nuclear Safety. OECD Nuclear Energy Agency: The Russian Federation to join the OECD Nuclear Energy Agency; Participation by the regulatory authorities of India and the United Arab Emirates in the Multinational Design Evaluation Programme (MDEP); NEA International Workshop on Crisis Communication, 9-10 May 2012; International School of Nuclear Law: 2013; Next NEA International Nuclear Law Essentials Course

Reactor Vessel and Reactor Vessel Internals Segmentation at Zion Nuclear Power Station - 13230

Energy Technology Data Exchange (ETDEWEB)

Cooke, Conrad; Spann, Holger [Siempelkamp Nuclear Services: 5229 Sunset Blvd., (Suite M), West Columbia, SC, 29169 (United States)

2013-07-01

Zion Nuclear Power Station (ZNPS) is a dual-unit Pressurized Water Reactor (PWR) nuclear power plant located on the Lake Michigan shoreline, in the city of Zion, Illinois approximately 64 km (40 miles) north of Chicago, Illinois and 67 km (42 miles) south of Milwaukee, Wisconsin. Each PWR is of the Westinghouse design and had a generation capacity of 1040 MW. Exelon Corporation operated both reactors with the first unit starting production of power in 1973 and the second unit coming on line in 1974. The operation of both reactors ceased in 1996/1997. In 2010 the Nuclear Regulatory Commission approved the transfer of Exelon Corporation's license to ZionSolutions, the Long Term Stewardship subsidiary of EnergySolutions responsible for the decommissioning of ZNPS. In October 2010, ZionSolutions awarded Siempelkamp Nuclear Services, Inc. (SNS) the contract to plan, segment, remove, and package both reactor vessels and their respective internals. This presentation discusses the tools employed by SNS to remove and segment the Reactor Vessel Internals (RVI) and Reactor Vessels (RV) and conveys the recent progress. SNS's mechanical segmentation tooling includes the C-HORCE (Circumferential Hydraulically Operated Cutting Equipment), BMT (Bolt Milling Tool), FaST (Former Attachment Severing Tool) and the VRS (Volume Reduction Station). Thermal segmentation of the reactor vessels will be accomplished using an Oxygen- Propane cutting system. The tools for internals segmentation were designed by SNS using their experience from other successful reactor and large component decommissioning and demolition (D and D) projects in the US. All of the designs allow for the mechanical segmentation of the internals remotely in the water-filled reactor cavities. The C-HORCE is designed to saw seven circumferential cuts through the Core Barrel and Thermal Shield walls with individual thicknesses up to 100 mm (4 inches). The BMT is designed to remove the bolts that fasten the Baffle

Outlines of guidelines for the inspection and evaluation of reactor vessel internals

International Nuclear Information System (INIS)

Seki, Hiroaki; Kobayashi, Hiroyuki; Nakano, Morihito; Murai, Soutarou; Nomoto, Toshiharu

2014-01-01

'The guideline committee for the inspection and evaluation of Reactor Vessel Internals' of JANSI (Japan Nuclear Safety Institute) has been developing many guidelines based on principle which the conservative methodology, and covered both individual inspection method of reactor internals and application of repair methods for reactor internals. In this paper, some aspects of the JANSI-VIP-03 (Guidelines for the inspection and evaluation of Reactor Vessel Internals, revised Dec.2013) which is summary document of the committee activity, are introduced. (author)

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.)

The situation of Chinese atomic energy and cooperation

International Nuclear Information System (INIS)

Nagasaki, Takao

2003-01-01

China will have 8.7 million kW atomic energy in 2005. Japan will complete with China in a sale war of international atomic energy and domestic power source. The position, development and situation of Chinese atomic energy and the future nuclear fuel cycle are reported. 5.4 million kW of 7 atomic power plants in China and 45.9 million kW of 53 plants in Japan are running. 3.3 million kW of 4 plants in China and 4 million kW of 4 plants in Japan are building. New type reactor, the fast breeder and high temperature gas-cooled reactor are developing. Radiation exposure to food, radiation therapy, Radio-pharmaceuticals, polymerization and treatment of sewage and smoke are carried out. The situation of atomic energy co-operation between China and Japan and other countries are stated. Japan has to change to advance mutual interests type co-operation with China. Construction of the nuclear community in Asia area and development of the international long big project are proposed. (S.Y.)

International Working Group on Fast Reactors Sixth Annual Meeting. Summary Report

International Nuclear Information System (INIS)

1973-01-01

The Agenda of the Meeting was as follows: 1. Review of IWGFR Activities - 1a. Approval of the minutes of the Fifth IWGFR Meeting. 1b. Report by Scientific Secretary regarding the activities of the Group. 2. Comments on National Programmes on Fast Breeder Reactors. 3. International Coordination of the Schedule for Major Fast Reactor Meetings and other major international meetings having a predominant fast reactor interest. 4. Consideration of Conferences on Fast Reactors. 4a. IAEA Symposium on Fuel and Fuel Elements for Fast Reactors, Brussels, Belgium 2-6 July 1973. 4b. International Symposium on Physics of Fast Reactors, Tokyo, Japan, 16 to 23 October 1973. 4c. International Conference on Fast Reactor Power Stations, London, UK, 11 to 14 March 1974 . 4d. Suggestions of the IWGFR members on other conferences. 5. Consideration of a Schedule for Specialists' Meetings in 1973-74. 6. Other Business - 6a. First-aid in Sodium Burns. 6b. Principles of Good Practice for Safe Operation of Sodium Circuits. 6c. Bibliography on Fast Reactors. 7. The Date and Place of the Seventh Annual Meeting of the IWGFR

Proceedings of the international topical seminar on management of ageing of research reactors, Geesthacht/Hamburg, May 8-12, 1995

International Nuclear Information System (INIS)

Alcala-Ruiz, F.; Krull, W.

1995-01-01

The GKSS Research Centre Geesthacht GmbH and the International Atomic Energy Agency (IAEA) held a joint seminar on 'Management of Ageing of Research Reactors' from May 8-12, 1995, at Geesthacht/Hamburg. More than 100 participants from 36 countries and two international organizations were present. During the seminar 52 papers have been presented discussing the large variety of ageing effects from physical ageing to design ageing and staff ageing and the measures taken to identify and to overcome the ageing effects. (orig.) [de

International breeder reactor development

International Nuclear Information System (INIS)

Traube, K.

1976-01-01

For more than a decade, sodium cooled breeder reactors have now been in the focus of advanced nuclear power development in the major industrialized countries. In the sixties, a total of seven small experimental nuclear power stations were commissioned. Two of these have been shut down in the meantime, the others continue to work satisfactorily, their main purpose being the development of fuel elements. The years 1972-1974 saw the commissioning of the prototype power stations in the 300 MWe power category in France, the United Kingdom and the Soviet Union. Presently, other experimental reactors are under construction in the Federal Republic of Germany, Italy, Japan, the United States, plus another Soviet 600 MWe prototype reactor and the SNR 300 DeBeNeLux prototype at Kalkar. A comparison of the technological features either implemented or planned in the prototype and experimental power plants and of their fuel elements reveals a remarkable similarity in the basic concepts pursued in different countries. The two types of breeder reactors, viz. the loop and the pool types, show a closer resemblance to each other than do pressurized and boilling water reactors. The growing awareness of administrative problems emerging in the approaching phase of the introduction of large breeder power stations in a number of European countries has recently led to a streamlining effort in the structure of industries and to tentative steps towards international cooperation on a broad basis. (orig.) [de

Nuclear reactor internals construction and failed fuel rod detection system

International Nuclear Information System (INIS)

Frisch, E.; Andrews, H.N.

1976-01-01

A system is provided for determining during operation of a nuclear reactor having fluid pressure operated control rod mechanisms the exact location of a fuel assembly with a defective fuel rod. The construction of the reactor internals is simplified in a manner to facilitate the testing for defective fuel rods and the reduce the cost of producing the upper internals of the reactor. 13 claims, 10 drawing figures

Dimensional control and check of field machining parts for reactor internals installation

International Nuclear Information System (INIS)

Zhang Caifang

2010-01-01

Some key issues of dimensional control for reactor internals installation are analyzed, and important technical requirements of crucial quality control elements on the measurement, machining, and checking of reactor internals filed machining parts are discussed. Moreover, provisions on quality control and risk prevention of reactor internals filed machining parts are presented in this paper. (author)

IRPhEP-handbook, International Handbook of Evaluated Reactor Physics Benchmark Experiments

International Nuclear Information System (INIS)

Sartori, Enrico; Blair Briggs, J.

2008-01-01

1 - Description: The purpose of the International Reactor Physics Experiment Evaluation Project (IRPhEP) is to provide an extensively peer-reviewed set of reactor physics-related integral data that can be used by reactor designers and safety analysts to validate the analytical tools used to design next-generation reactors and establish the safety basis for operation of these reactors. This work of the IRPhEP is formally documented in the 'International Handbook of Evaluated Reactor Physics Benchmark Experiments,' a single source of verified and extensively peer-reviewed reactor physics benchmark measurements data. The IRPhE Handbook is available on DVD. You may request a DVD by completing the DVD Request Form available at: http://irphep.inl.gov/handbook/hbrequest.shtml The evaluation process entails the following steps: 1. Identify a comprehensive set of reactor physics experimental measurements data, 2. Evaluate the data and quantify overall uncertainties through various types of sensitivity analysis to the extent possible, verify the data by reviewing original and subsequently revised documentation, and by talking with the experimenters or individuals who are familiar with the experimental facility, 3. Compile the data into a standardized format, 4. Perform calculations of each experiment with standard reactor physics codes where it would add information, 5. Formally document the work into a single source of verified and peer reviewed reactor physics benchmark measurements data. The International Handbook of Evaluated Reactor Physics Benchmark Experiments contains reactor physics benchmark specifications that have been derived from experiments that were performed at various nuclear experimental facilities around the world. The benchmark specifications are intended for use by reactor physics personal to validate calculational techniques. The 2008 Edition of the International Handbook of Evaluated Reactor Physics Experiments contains data from 25 different

German atomic energy law in the international framework

International Nuclear Information System (INIS)

Pelzer, N.

1992-01-01

The regional conference was devoted to the legal problems that ensue from German reunification against the background of the integration of German atomic energy law within international law. The elements of national atomic energy legislation required by international law and recent developments in international nuclear liability law were discussed from different perspectives. The particular problems of the application of the German Atomic Energy Act in the 5 new Laender (the territories of the former GDR) were presented and discussed, namely: The continued validity of old licences issued by the GDR; practical legal problems connected with the construction of nuclear power plants in the 5 new Laender; the legal issues connected with the final repository for radioactive wastes at Morsleben; and the new developments in radiation protection law following from the Unification Treaty and the new ICRP recommendations. All 14 lectures have been abstracted and indexed individually. (orig.) [de

Age-related degradation of boiling water reactor vessel internals

International Nuclear Information System (INIS)

Ware, A.G.; Shah, V.N.

1992-01-01

Researchers at the Idaho National Engineering Laboratory performed an assessment of the aging of the reactor internals in boiling water reactors (BWRs), and identified the unresolved technical issues related to the degradation of these components. The overall life-limiting mechanism is intergranular stress corrosion cracking (IGSCC). Irradiation-assisted stress corrosion cracking, fatigue, and thermal aging embrittlement are other potential degradation mechanisms. Several failures in BWR internals have been caused by a combination of factors such as environment, high residual or preload stresses, and flow-induced vibration. The ASME Code Section XI in-service inspection requirements are insufficient for detecting aging-related degradation at many locations in reactor internals. Many of the potential locations for IGSCC or fatigue are not accessible for inspection. (orig.)

On the International Atomic Energy Agency

Energy Technology Data Exchange (ETDEWEB)

Eklund, S [International Atomic Energy Agency, Vienna (Austria)

1963-07-15

The main concepts motivating the decision to establish an international agency for peaceful uses of atomic energy are presented in the paper. They consists of: 1) co-ordination in the fields of safety field, legal liability and safeguards; 2) ensuring that scientific and technical data are made freely accessible on a worldwide scale and 3) assisting the developing countries in benefiting from this new science and technology and use the atomic energy for economic and social development

The International Atomic Energy Agency's safeguards system

International Nuclear Information System (INIS)

Wagner, W.

2000-01-01

A system of international safeguards has been established to provide assurance that nuclear materials in civilian use are not diverted from their peaceful purpose. The safeguards system is administered by the International Atomic Energy Agency/Department of Safeguards and devolves from treaties and other international agreements. Inspectors from the Agency verify reports from States about nuclear facilities by audits, observation, and measurements. (author)

Segmentation and packaging reactor vessels internals

International Nuclear Information System (INIS)

Boucau, Joseph

2014-01-01

Document available in abstract form only, full text follows: With more than 25 years of experience in the development of reactor vessel internals and reactor vessel segmentation and packaging technology, Westinghouse has accumulated significant know-how in the reactor dismantling market. The primary challenges of a segmentation and packaging project are to separate the highly activated materials from the less-activated materials and package them into appropriate containers for disposal. Since disposal cost is a key factor, it is important to plan and optimize waste segmentation and packaging. The choice of the optimum cutting technology is also important for a successful project implementation and depends on some specific constraints. Detailed 3-D modeling is the basis for tooling design and provides invaluable support in determining the optimum strategy for component cutting and disposal in waste containers, taking account of the radiological and packaging constraints. The usual method is to start at the end of the process, by evaluating handling of the containers, the waste disposal requirements, what type and size of containers are available for the different disposal options, and working backwards to select a cutting method and finally the cut geometry required. The 3-D models can include intelligent data such as weight, center of gravity, curie content, etc, for each segmented piece, which is very useful when comparing various cutting, handling and packaging options. The detailed 3-D analyses and thorough characterization assessment can draw the attention to material potentially subject to clearance, either directly or after certain period of decay, to allow recycling and further disposal cost reduction. Westinghouse has developed a variety of special cutting and handling tools, support fixtures, service bridges, water filtration systems, video-monitoring systems and customized rigging, all of which are required for a successful reactor vessel internals

International Atomic Energy Agency and Malaysia

International Nuclear Information System (INIS)

Abd Rahim Mohd Nor

1985-01-01

A review on IAEA (International Atomic Energy Agency) and its relation with Malaysia is given. This article also discusses the background history of IAEA, its organization and functions in the field of nuclear energy

Design of reactor internals in larger high-temperature reactors with spherical fuel elements

International Nuclear Information System (INIS)

Elter, C.

1981-01-01

In his paper, the author analyzes and summarizes the present state of the art with emphasis on the prototype reactor THTR 300 MWe, because in addition to spherical fuel elements, this type includes other features of future HTR design such as the same flow direction of cooland gas through the core. The paper on hand also elaborates design guidelines for reactor internals applicable with large HTR's of up to 1200 MWe. Proved designs will be altered so as to meet the special requirements of larger cores with spherical elements to be reloaded according to the OTTO principle. This paper is furthermore designed as a starting point for selective and swift development of reactor internals for large HTR's to be refuelled according to the OTTO principle. (orig./GL) [de

International review on safety requirements for the prototype fast breeder reactor “Monju” (Translated document)

International Nuclear Information System (INIS)

2016-02-01

In response to the lessons learned from the serious nuclear accidents at the TEPCO's Fukushima Daiichi Nuclear Power Stations, an advisory committee, which was set up by the Japan Atomic Energy Agency, issued the report “Safety Requirements Expected to the Prototype Fast Breeder Reactor Monju” taking into account the SFR specific safety characteristics in July 2014. The report was reviewed by the leading international experts on SFR safety from five countries and one international organization in order to obtain independent and objective evaluation. The international review comments on each subsection were collected and compiled, and then a summary of results was derived through the discussion at the review meeting and individual feedbacks. As a result the basic concept for prevention of severe accidents and mitigation of their consequences of Monju is appropriate in consideration of SFR specific safety characteristics, and is in accordance with international common understanding. (author)

Reactors physics. Bases of nuclear physics

International Nuclear Information System (INIS)

Diop, Ch.M.

2006-01-01

The aim of nuclear reactor physics is to quantify the relevant macroscopic data for the characterization of the neutronic state of a reactor core and to evaluate the effects of radiations (neutrons and gamma radiations) on organic matter and on inorganic materials. This first article presents the bases of nuclear physics in the context of nuclear reactors: 1 - reactor physics and nuclear physics; 2 - atomic nucleus - basic definitions: nucleus constituents, dimensions and mass of the atomic nucleus, mass defect, binding energy and stability of the nucleus, strong interaction, nuclear momentums of nucleons and nucleus; 3 - nucleus stability and radioactivity: equation of evolution with time - radioactive decay law; alpha decay, stability limit of spontaneous fission, beta decay, electronic capture, gamma emission, internal conversion, radioactivity, two-body problem and notion of radioactive equilibrium. (J.S.)

Feedback from Westinghouse experience on segmentation of reactor vessel internals - 59013

International Nuclear Information System (INIS)

Kreitman, Paul J.; Boucau, Joseph; Segerud, Per; Fallstroem, Stefan

2012-01-01

With more than 25 years of experience in the development of reactor vessel internals segmentation and packaging technology, Westinghouse has accumulated significant know-how in the reactor dismantling market. Building on tooling concepts and cutting methodologies developed decades ago for the successful removal of nuclear fuel from the damaged Three Mile Island Unit 2 reactor (TMI-2), Westinghouse has continuously improved its approach to internals segmentation and packaging by incorporating lessons learned and best practices into each successive project. Westinghouse has developed several concepts to dismantle reactor internals based on safe and reliable techniques, including plasma arc cutting (PAC), abrasive water-jet cutting (AWJC), metal disintegration machining (MDM), or mechanical cutting. Westinghouse has applied its technology to all types of reactors covering Pressurized Water Reactors (PWR's), Boiling Water Reactors (BWR's), Gas Cooled Reactors (GCR's) and sodium reactors. The primary challenges of a segmentation and packaging project are to separate the highly activated materials from the less-activated materials and package them into appropriate containers for disposal. Since space is almost always a limiting factor it is therefore important to plan and optimize the available room in the segmentation areas. The choice of the optimum cutting technology is important for a successful project implementation and depends on some specific constraints like disposal costs, project schedule, available areas or safety. Detailed 3-D modeling is the basis for tooling design and provides invaluable support in determining the optimum strategy for component cutting and disposal in waste containers, taking account of the radiological and packaging constraints. Westinghouse has also developed a variety of special handling tools, support fixtures, service bridges, water filtration systems, video-monitoring systems and customized rigging, all of which are required for a

Studies on reactor physics

International Nuclear Information System (INIS)

1960-01-01

Most of the peaceful applications of atomic energy are inherently dependent on advances in the science and technology of nuclear reactors, and aspects of this development are part of a major programme of the International Atomic Energy Agency. The most useful role that the Agency can play is as a co-ordinating body or central forum where the trends can be reviewed and the results assessed. Some of the basic studies are carried out by members of the Agency's own scientific staff. The Agency also convenes groups of experts from different countries to examine a particular problem in detail and make any necessary recommendations. Some of the important subjects are discussed at international scientific meetings held by the Agency. One of the subjects covered by such studies is the physics of nuclear reactors and a specific topic recently discussed was Codes for Reactor Computations, on which a seminar was held in Vienna in April this year. Another The members of the Panel described the development of heavy water reactors, the equipment and methods of research currently used, and plans for further development in their respective countries meeting of Panel of Experts on Heavy Water Lattices was held in Vienna in August 1959

What is the future for fast reactor technology?

International Nuclear Information System (INIS)

Kraev, Kamen

2017-01-01

NucNet spoke to Vladimir Kriventsev, team leader for fast reactor technology development at the International Atomic Energy Agency (IAEA), about the possibilities and challenges of technology development in the fast reactor sector. Today, the field of fast reactors is vibrant and full of fascinating developments, some which will have an impact in the nearer term and others in the longer term.

What is the future for fast reactor technology?

Energy Technology Data Exchange (ETDEWEB)

Kraev, Kamen [NucNet, Brussels (Belgium). The Independent Global Nuclear News Agency

2017-08-15

NucNet spoke to Vladimir Kriventsev, team leader for fast reactor technology development at the International Atomic Energy Agency (IAEA), about the possibilities and challenges of technology development in the fast reactor sector. Today, the field of fast reactors is vibrant and full of fascinating developments, some which will have an impact in the nearer term and others in the longer term.

Reactors of the world

International Nuclear Information System (INIS)

1971-01-01

Basic data relating to 127 power reactors in 15 countries which are expected to be in operation at the end of this year, with a total installed electrical generating capacity of 35 340.15 MW(e), and a listing of 361 research reactors in 46 countries are given in the 1971 edition of the IAEA handbook, Power and Research Reactors in Member States, which has just been published. This edition, the fourth, was prepared especially for the Fourth International Conference on the Peaceful Uses of Atomic Energy. (author)

Fabrication of internally instrumented reactor fuel rods

International Nuclear Information System (INIS)

Schmutz, J.D.; Meservey, R.H.

1975-01-01

Procedures are outlined for fabricating internally instrumented reactor fuel rods while maintaining the original quality assurance level of the rods. Instrumented fuel rods described contain fuel centerline thermocouples, ultrasonic thermometers, and pressure tubes for internal rod gas pressure measurements. Descriptions of the thermocouples and ultrasonic thermometers are also contained

Decommissioning of the Dragon High Temperature Reactor (HTR) Located at the Former United Kingdom Atomic Energy Authority (UKAEA) Research Site at Winfrith - 13180

Energy Technology Data Exchange (ETDEWEB)

Smith, Anthony A. [Research Sites Restoration Ltd, Winfrith, Dorset (United Kingdom)

2013-07-01

The Dragon Reactor was constructed at the United Kingdom Atomic Energy Research Establishment at Winfrith in Dorset through the late 1950's and into the early 1960's. It was a High Temperature Gas Cooled Reactor (HTR) with helium gas coolant and graphite moderation. It operated as a fuel testing and demonstration reactor at up to 20 MW (Thermal) from 1964 until 1975, when international funding for this project was terminated. The fuel was removed from the core in 1976 and the reactor was put into Safestore. To meet the UK's Nuclear Decommissioning Authority (NDA) objective to 'drive hazard reduction' [1] it is necessary to decommission and remediate all the Research Sites Restoration Ltd (RSRL) facilities. This includes the Dragon Reactor where the activated core, pressure vessel and control rods and the contaminated primary circuit (including a {sup 90}Sr source) still remain. It is essential to remove these hazards at the appropriate time and return the area occupied by the reactor to a safe condition. (author)

Thermal compatibility of U-2wt.%Mo and U-10wt.%Mo fuel prepared by centrifugal atomization for high density research reactor fuels

International Nuclear Information System (INIS)

Kim Ki Hwan; Lee Don Bae; Kim Chang Kyu; Kuk Il Hyun; Hofman, G.E.

1997-01-01

Research on the intermetallic compounds of uranium was revived in 1978 with the decision by the international research reactor community to develop proliferation-resistant fuels. The reduction of 93% 235 U (HEU) to 20% 235 U (LEU) necessitates the use of higher U-loading fuels to accommodate the addition 238 U in the LEU fuels. While the vast majority of reactors can be satisfied with U 3 Si 2 -Al dispersion fuel, several high performance reactors require high loadings of up to 8-9 g U cm -3 . Consequently, in the renewed fuel development program of the Reduced Enrichment for Research and Test Reactors (RERTR) Program, attention has shifted to high density uranium alloys. Early irradiation experiments with uranium alloys showed promise of acceptable irradiation behavior, if these alloys can be maintained in their cubic γ-U crystal structure. It has been reported that high density atomized U-Mo powders prepared by rapid cooling have metastable isotropic γ-U phase saturated with molybdenum, and good γ-U phase stability, especially in U-10wt.%Mo alloy fuel. If the alloy has good thermal compatibility with aluminium, and this metastable gamma phase can be maintained during irradiation, U-Mo alloy would be a prime candidate for dispersion fuel for research reactors. In this paper, U-2w.%Mo and U-10w.%Mo alloy powder which have high density (above 15 g-U/cm 3 ), are prepared by centrifugal atomization. The U-Mo alloy fuel meats are made into rods extruding the atomized powders. The characteristics related to the thermal compatibility of U-2w.%Mo and U-10w.%Mo alloy fuel meat at 400 o C for time up to 2000 hours are examined. (author)

The international law and the pacific uses of the atomic energy

International Nuclear Information System (INIS)

Mora, A.; Gutierrez, I.; Vargas, N.M.

1992-01-01

Contains information about: fundamental aspects of atomic energy; International Atomic Energy Agency; pacific uses of nuclear energy at national and international level; regulation for some risky activities in the pacific uses of radioactive materials; United Nations system for the secure use of atomic energy with pacific purposes; nuclear accidents; responsibility as fundamental element of nuclear law. 207 refs

International Reactor Physics Experiment Evaluation (IRPhE) Project

International Nuclear Information System (INIS)

2013-01-01

The International Reactor Physics Experiment Evaluation (IRPhE) Project aims to provide the nuclear community with qualified benchmark data sets by collecting reactor physics experimental data from nuclear facilities, worldwide. More specifically the objectives of the expert group are as follows: - maintaining an inventory of the experiments that have been carried out and documented; - archiving the primary documents and data released in computer-readable form; - promoting the use of the format and methods developed and seek to have them adopted as a standard. For those experiments where interest and priority is expressed by member countries or working parties and executive groups within the NEA provide guidance or co-ordination in: - compiling experiments into a standard international agreed format; - verifying the data, to the extent possible, by reviewing original and subsequently revised documentation, and by consulting with the experimenters or individuals who are familiar with the experimenters or the experimental facility; - analysing and interpreting the experiments with current state-of-the-art methods; - publishing electronically the benchmark evaluations. The expert group will: - identify gaps in data and provide guidance on priorities for future experiments; - involve the young generation (Masters and PhD students and young researchers) to find an effective way of transferring know-how in experimental techniques and analysis methods; - provide a tool for improved exploitation of completed experiments for Generation IV reactors; - coordinate closely its work with other NSC experimental work groups in particular the International Criticality Safety Benchmark Evaluation Project (ICSBEP), the Shielding Integral Benchmark Experiment Data Base (SINBAD) and others, e.g. knowledge preservation in fast reactors of the IAEA, the ANS Joint Benchmark Activities; - keep a close link with the working parties on scientific issues of reactor systems (WPRS), the expert

Installation of the Light-Water Breeder Reactor at the Shippingport Atomic Power Station (LWBR Development Program)

International Nuclear Information System (INIS)

Massimino, R.J.; Williams, D.A.

1983-05-01

This report summarizes the refueling operations performed to install a Light Water Breeder Reactor (LWBR) core into the existing pressurized water reactor vessel at the Shippingport Atomic Power Station. Detailed descriptions of the major installation operations (e.g., primary system preconditioning, fuel installation, pressure boundary seal welding) are included as appendices to this report; these operations are of technical interest to any reactor servicing operation, whether the reactor is a breeder or a conventional light water non-breeder core

Installation of the Light-Water Breeder Reactor at the Shippingport Atomic Power Station (LWBR Development Program)

Energy Technology Data Exchange (ETDEWEB)

Massimino, R.J.; Williams, D.A.

1983-05-01

This report summarizes the refueling operations performed to install a Light Water Breeder Reactor (LWBR) core into the existing pressurized water reactor vessel at the Shippingport Atomic Power Station. Detailed descriptions of the major installation operations (e.g., primary system preconditioning, fuel installation, pressure boundary seal welding) are included as appendices to this report; these operations are of technical interest to any reactor servicing operation, whether the reactor is a breeder or a conventional light water non-breeder core.

Degradation of fastener in reactor internal of PWR

Energy Technology Data Exchange (ETDEWEB)

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

2000-03-01

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

Comparison of advanced reactors program of different international vendors

International Nuclear Information System (INIS)

Agnihotri, N.K.

2001-01-01

The full text follows. Proposal for presenting a paper on Advanced Reactor Program Given below is the abstract for Track 6 session on Advanced Reactor at the ninth International Conference on Nuclear Engineering being held in Nice, France from April 8. through 12. 2001. This paper will provide an update on Advanced Reactor Program of different vendors in the United States, Japan, and Europe. Specifically the paper will look at the history of different Advanced Reactor Programs, international experience, aspect of economy due to standardization, and the highlights of technical specifications. The paper will also review aspects of Economy due to standardization, public acceptance, required construction time, and the experience of different vendors. The objective of the presentation is to underscore the highlights of the Reactor Program of different vendors in order to keep the attendees of the conference up-to-date. The presentation will be an impartial overview from an outsider's (not part of the Nuclear Steam Supply System's staff). (author)

Change in plan for installation of nuclear reactor in No.1 atomic powered vessel of Japan Atomic Energy Research Institute (change in purpose of use and in method for nuclear reactor installation and spent fuel disposal) (report)

International Nuclear Information System (INIS)

1987-01-01

This report, compiled by the Nuclear Safety Commission to be submitted to the Prime Minister, deals with studies concerning some changes in the plan for the installation of a nuclear reactor in the No.1 atomic powered vessel to be constructed under the Japan Atomic Energy Research Institute (changes in the purpose of its use and in the methods for the nuclear reactor installation and spent fuel disposal). The conclusions of and procedures for the examination and evaluation are presented and then detailes of the studies are described. The study on the location requirements for the incidental land facilities at Sekinehama covers various conditions concerning the location, geology, earthquakes, meteorology, hydrology and social environment. The study on the safety design of the nuclear reactor facilities deals with the reactor, fuel handling facilities and other auxiliary facilities, as well as various land facilities to be constructed at Sekinehama including the reactor facilities and other facilities for fuel handling, waste disposal and protection and management of radioactive rays. Evaluation of possible radiation emission is shown and the accident analysis is also addressed. (Nogami, K.)

International Conference 'Current Problems in Nuclear Physics and Atomic Energy'. May 29 - Jun 03 2006. Book of Abstracts

International Nuclear Information System (INIS)

Vyshnevskyi, I.M.

2006-01-01

The collective processes in atomic nuclei, nuclear reactions and processes with exotic nuclei, rare nuclear processes, relativistic nuclear physics, neutron physics, physics of nuclear reactors, problems of atomic energy and reactors of the future, applied nuclear physics and technique of experiments was discussed in this conference

Iser: an international inherently safe reactor concept

International Nuclear Information System (INIS)

Wakabayashi, Hiroaki

1988-01-01

Iser is a modular standardised 200-300 MWe power reactor based on the PIUS principle. It differs from PIUS in being simpler, and making full use of existing steel-vessel-based LWR technology. Iser is an inherently safe reactor concept under development in Japan. It is a generic concept, not a patented commodity, and it is expected that an international association to develop the concept will be formed. (U.K.)

Developments in the regulation of research reactors

International Nuclear Information System (INIS)

Loy, J.

2003-01-01

The International Atomic Energy Agency (IAEA) has data on over 670 research reactors in the world. Fewer than half of them are operational and a significant number are in a shutdown but not decommissioned state. The International Nuclear Safety Advisory Group (INSAG) has expressed concerns about the safety of many research reactors and this has resulted in a process to draw up an international Code of Conduct on the Safety of Research Reactors. The IAEA is also reviewing its safety standards applying to research reactors. On the home front, regulation of the construction of the Replacement Research Reactor continues. During the construction phase, regulation has centred around the consideration of Requests for Approval (RFA) for the manufacture and installation of systems, structures and components important for safety. Quality control of construction of systems, structures and components is the central issue. The process for regulation of commissioning is under consideration

Determination of internationally controlled materials according to provisions of the law for the regulations of nuclear source materials, nuclear fuel materials and reactors

International Nuclear Information System (INIS)

1984-01-01

The internationally controlled materials determined according to the law for nuclear source materials, etc. are the following: nuclear source materials, nuclear fuel materials, moderating materials, facilities including reactors, etc. sold, transferred, etc. to Japan according to the agreements for peaceful uses of atomic energy between Japan, and the United States, the United Kingdom, Canada, Australia and France by the respective governments and those organs under them; nuclear fuel materials resulting from usage of the above sold and transferred materials, facilities; nuclear fuel materials sold to Japan according to agreements set by the International Atomic Energy Agency; nuclear fuel materials involved with the safeguards in nuclear weapons non-proliferation treaty with IAEA. (Mori, K.)

Determination of internationally controlled materials according to provisions of the law for the regulations of nuclear source materials, nuclear fuel materials and reactors

International Nuclear Information System (INIS)

1977-01-01

According to the provisions of The Law, those stipulated as internationally controlled materials are nuclear source materials, nuclear fuel materials, moderating materials, reactors and facilities, transferred from such as the U.S.A., the U.K. and Canada on the agreements of peaceful uses of atomic energy, and nuclear fuel materials accruing therefrom. (Mori, K.)

Development of cutting technique of reactor core internals by CO laser

International Nuclear Information System (INIS)

Takano, G.; Beppu, S.; Matsumoto, O.; Sakamoto, N.; Onozawa, T.; Sugihara, M.; Miya, K.

1995-01-01

The CO laser is superior in the absorption characteristic to materials to the CO 2 laser due to its shorter wavelength. In consideration of this characteristic Nuclear Power Engineering Corporation is studying this applicability sponsored by the Ministry of International Trade Industry of Japan to cutting of reactor core internals of commercial nuclear power plant. In decommissioning of reactor core internals it is necessary to cut stainless steel plates of 305 mm thick. The authors cut stainless steel plates of up to 310mm thick in air and those of up to 150 mm thick underwater with a 20kW class laser. Further, models simulating key structural elements of PWR core internals were cut and secondary products to clarify the applicability of the CO laser cutting to reactor core internals were evaluated. (author)

Annual Report to Congress of the Atomic Energy Commission for 1966

Energy Technology Data Exchange (ETDEWEB)

Seaborg, Glenn T.

1967-01-31

The document represents the 1966 Annual Report of the Atomic Energy Commission (AEC) to Congress. The report opens with a Summary of Atomic Energy Programs in 1966 and includes 18 Chapters, 11 appendices and an index. Chapters are as follows: (1) Licensing and Regulating the Atom; (2) Reactor and Other Nuclear Facility Licensing; (3) The Regulation of Radioactive Materials; (4) Source and Special Nuclear Materials; (5) The Nuclear Defense Effort; (6) Naval Propulsion Reactors; (7) Reactor Development and Technology; (8) Space Nuclear Systems; (9) Isotopic Heat and Power Applications; (10) Isotopic Radiation Applications; (11) The Plowshare Program; (12) International Cooperation Activities; (13) Research Facilities and Projects; (14) Nuclear Education and Training; (15) Informational Activities; (16) Operational Safety; (17) Industrial Participation Aspects; and, (18) Administrative and Management Matters.

IAEA activities in gas-cooled reactor technology development

International Nuclear Information System (INIS)

Cleveland, J.; Kupitz, J.

1992-01-01

The International Atomic Energy Agency (IAEA) has the charter to ''foster the exchange of scientific and technical information'', and ''encourage and assist research on, and development and practical application of, atomic energy for peaceful uses throughout the world''. This paper describes the Agency's activities in Gas-cooled Reactor (GCR) technology development

International conference on fast reactors and related fuel cycles (FR09): Challenges and opportunities. CN-176 presentations

International Nuclear Information System (INIS)

2009-01-01

and development, as well as in design, has not been reported in a coordinated manner, which has made planning and implementation of expensive research and technology development programmes rather difficult. There is a perceived need for an appropriate forum to achieve the twin objectives of exchanging experience and innovative ideas among experts, and of sharing knowledge and mentoring, whereby experienced scientists and technologists, as well as fast reactor programme managers, would share their perspectives with the future generation of young scientists and technologists, helping them to choose research problems of eminence and pursue their careers to meet the challenges of the development of fast reactors with recycle. After almost 20 years, this is also the appropriate time, as fast reactor programmes are currently on an accelerated growth path in many countries of the world. It is in light of this that the IAEA convened on 7-11 December 2009 in Kyoto, Japan, the International Conference on 'Fast Reactors and Related Fuel Cycles - Challenges and Opportunities (FR09)', hosted by the Japan Atomic Energy Agency. This conference aimed at promoting the exchange of information on national and multinational programmes and new developments and experience, with the goal of identifying and critically reviewing problems of importance, and stimulating and facilitating cooperation, development and successful deployment of fast reactors in an expeditious manner

Research reactor core conversion programmes, Department of Research and Isotopes, International Atomic Energy Agency

International Nuclear Information System (INIS)

Muranaka, R.G.

1985-01-01

In order to put the problem of core conversion into perspective, statistical information on research reactors on a global scale is presented (from IAEA Research reactor Data Base). This paper describes the research reactor core conversion program of the Department of Research and Isotopes. Technical committee Meetings were held on the subject of research reactor core conversion since 1978, and results of these meetings are published in TECDOC-233, TECDOC-324, TECDOC-304. Additional publications are being prepared, several missions of experts have visited countries to discuss and help to plan core conversion programs; training courses and seminars were organised; IAEA has supported attendance of participants from developing countries to RERTR Meetings

Agreement of 10 September 1991 between the International Atomic Energy Agency and the Government of the Islamic Republic of Pakistan for the application of safeguards in connection with the supply of a miniature neutron source reactor from the People's Republic of China

International Nuclear Information System (INIS)

1991-10-01

The document reproduces the text of the Agreement of 10 September 1991, between the Government of the Islamic Republic of Pakistan and the International Atomic Energy Agency for the application of safeguards in connection with the supply of a miniature neutron source reactor from the People's Republic of China. The Agreement was approved by the Agency's Board of Governors on 20 February 1990 and entered into force upon signature on 10 September 1991

International bulletin on atomic and molecular data for fusion. No. 59

International Nuclear Information System (INIS)

Stephens, J.A.; Bannister, M.E.; Fuhr, J.; Gilbody, H.B.

2001-03-01

The International Bulletin on Atomic and Molecular Data for Fusion is prepared by the Atomic and Molecular Data Unit of the International Atomic Energy Agency. It is distributed free of charge by the IAEA to assist in the development of fusion research and technology. In part 1, the Atomic and Molecular Data Information System (AMDIS) is presented. In Part 2, the indexed papers are listed separately for structure and spectra, atomic and molecular collisions and surface interactions. Part 3 contains all the bibliographic data for both the indexed and non-indexed references. Finally, the Author Index (part 4) refers to the bibliographic references contained in part 3

International bulletin on atomic and molecular data for fusion. No. 58

International Nuclear Information System (INIS)

Stephens, J.; Bannister, M.E.; Fuhr, J.; Gilbody, H.B.

2000-06-01

The International Bulletin on Atomic and Molecular Data for Fusion is prepared by the Atomic and Molecular Data Unit of the International Atomic Energy Agency. It is distributed free of charge by the IAEA to assist in the development of fusion research and technology. In part 1, the Atomic and Molecular Data Information System (AMDIS) is presented. In Part 2, the indexed papers are listed separately for structure and spectra, atomic and molecular collisions and surface interactions. Part 3 contains all the bibliographic data for both the indexed and non-indexed references. Finally, the Author Index (part 4) refers to the bibliographic references contained in part 3

The law for the regulations of nuclear source materials, nuclear fuel materials and reactors

International Nuclear Information System (INIS)

1979-01-01

The law aims to perform regulations on enterprises of refining, processing and reprocessing of nuclear source and fuel materials and on establishment and operation of reactors to realize the peaceful and deliberate utilization of atomic energy according to the principle of the atomic energy basic law. Regulations of use of internationally regulated substances are also envisaged to observe international agreements. Basic concepts and terms are defined, such as: atomic energy; nuclear fuel material; nuclear source material; reactor; refining; processing; reprocessing and internationally regulated substance. Any person besides the Power Reactor and Nuclear Fuel Material Developing Corporation who undertakes refining shall be designated by the Prime Minister and the Minister of International Trade and Industry. An application shall be filed to the ministers concerned, listing name and address of the person, name and location of the refining works, equipment and method of refining, etc. The permission of the Prime Minister is necessary for any person who engages in processing. An application shall be filed to the Prime Minister, listing name and address of the person, name and location of the processing works and equipment and method of processing, etc. Permission of the Prime Minister, the Minister of International Trade and Industry or the Minister of Transport is necessary for any person who sets up reactors. An application shall be filed to the minister concerned, listing name and address of the person, purpose of operation, style, thermal output of reactor and number of units, etc. (Okada, K.)

Management of research reactor ageing

International Nuclear Information System (INIS)

1995-03-01

As of December 1993, about one quarter of the operating research reactors were over 30 years old. The long life of research reactors has raised some concern amongst research reactor operators, regulators and, to some extent, the general public. The International Atomic Energy Agency commenced activities on the topic of research reactor ageing by appointing an internal working group in 1988 and convening a Consultants Meeting in 1989. The subject was also discussed at an international symposium and a regional seminar held in 1989 and 1992 respectively. A draft document incorporating information and experience exchanged at the above meetings was reviewed by a Technical Committee Meeting held in Vienna in 1992. The present TECDOC is the outcome of this meeting and contains recommendations, guidelines and information on the management of research reactor ageing, which should be used in conjunction with related publications of the IAEA Research Reactor Safety Programme, which are referenced throughout the text. This TECDOC will be of interest to operators and regulators involved with the safe operation of any type of research reactor to (a) understand the behaviour and influence of ageing mechanisms on the reactor structures, systems and components; (b) detect and assess the effect of ageing; (c) establish preventive and corrective measures to mitigate these effects; and (d) make decisions aimed at the safe and continued operation of a research reactor. 32 refs, tabs

Management of research reactor ageing

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-03-01

As of December 1993, about one quarter of the operating research reactors were over 30 years old. The long life of research reactors has raised some concern amongst research reactor operators, regulators and, to some extent, the general public. The International Atomic Energy Agency commenced activities on the topic of research reactor ageing by appointing an internal working group in 1988 and convening a Consultants Meeting in 1989. The subject was also discussed at an international symposium and a regional seminar held in 1989 and 1992 respectively. A draft document incorporating information and experience exchanged at the above meetings was reviewed by a Technical Committee Meeting held in Vienna in 1992. The present TECDOC is the outcome of this meeting and contains recommendations, guidelines and information on the management of research reactor ageing, which should be used in conjunction with related publications of the IAEA Research Reactor Safety Programme, which are referenced throughout the text. This TECDOC will be of interest to operators and regulators involved with the safe operation of any type of research reactor to (a) understand the behaviour and influence of ageing mechanisms on the reactor structures, systems and components; (b) detect and assess the effect of ageing; (c) establish preventive and corrective measures to mitigate these effects; and (d) make decisions aimed at the safe and continued operation of a research reactor. 32 refs, tabs.

Method of starting internal pumps of a nuclear reactor

International Nuclear Information System (INIS)

Kumagami, Shoji.

1985-01-01

Purpose: To reduce the noise effects by decreasing the invading current into the main line upon starting an internal pump type nuclear reactor adapted to forcively recycle the reactor water by a plurality of internal pumps. Method: A plurality of internal pumps are divided into several groups and, upon starting pumps belonging to the individual unit group, the starting instances for the respective pumps are deviated to reduce the surges applied to the main line and suppress the invading current lower to reduce the earth noises. As a result, effects caused to other devices or equipments can be moderated to improve the reliability. Furthermore, by actuating the respective pumps on every group units in a starting pattern along the orthogonal line, flow rate distribution in the reactor can be balanced. Then, the instability region during low rotation of pumps, that is, instability of the flow rate near the resonance frequency can be decreased. (Kawakami, Y.)

Partial thorium loading in the initial core of Kakrapar atomic power reactor

International Nuclear Information System (INIS)

Balakrishnan, M.R.

1993-01-01

The first unit of Kakrapar nuclear power station has gone critical with some thorium oxide fuel bundles loaded in its core. The thorium helps to flatten the power by reducing neutron flux in the centre of the reactor. However, the placing of the thorium had to be planned with care, because if the neutron flux at a point where a safety rod is located is depressed, the reactivity worth of the safety rod gets reduced. Using a dynamic programing approach, the Reactor Engineering Division of Bhabha Atomic Research Centre worked out a satisfactory configuration for loading the thorium bundles

Atomic data for heavy element impurities in fusion reactors. Summary report of first IAEA research co-ordination meeting

International Nuclear Information System (INIS)

Clark, R.E.H.

2006-01-01

Twelve international experts discussed in detail the properties of heavy elements relevant to fusion energy research participated at the first Research Coordination Meeting (RCM) of the Coordinated Research Project (CRP) on 'Atomic data for heavy element impurities in fusion reactors' at IAEA Headquarters on 14-15 November 2005. The participants summarized all recent relevant developments in their research efforts. Detailed discussions took place to formulate specific objectives for the CRP. From a list of data needs and a review of current research capabilities, a detailed work plan was formulated for the first phase of the CRP. The discussions, conclusions and recommendations of the RCM are briefly described in this report. (author)

The low-temperature water-water reactor for district heating atomic power plant (DHPP)

International Nuclear Information System (INIS)

Skvortsov, S.A.; Sokolov, I.N.; Krauze, L.V.; Nikiporetz, Yu.G.; Philimonov, Yu.V.

1977-01-01

The district heating atomic power plant in the article is distinguished by the increased reliability and safety of operation that was provided by the use of following main principles: relatively low parameters of the coolant; the intergral arrangement of equipment and accordingly the minimum branching of the reactor circuit; the natural circulation of coolant of the primary circuit in the steady-state, transient and emergency regimes of reactor operation; the considerable reserves of cold water of the primary circuit in the reactor vessel, providing the emergency cooling; the application of two shells each of which is designed for the total working pressure, the second shell is made of prestressed reinforced concrete that eliminates its brittle failure. (M.S.)

Development of inertia-increased reactor internal pump

International Nuclear Information System (INIS)

Tanaka, Masaaki; Matsumura, Seiichi; Kikushima, Jun; Kawamura, Shinichi; Yamashita, Norimichi; Kurosaki, Toshikazu; Kondo, Takahisa

2000-01-01

The Reactor Internal Pump (RIP) was adopted for the Reactor Recirculation System (RRS) of Advanced Boiling Water Reactor (ABWR) plants, and ten RIPs are located at the bottom of the reactor pressure vessel. In order to simplify the power supply system for the RIPs, a new inertia-increased RIP was developed, which allows to eliminate the Motor-Generator (M-G) sets. The rotating inertia was increased approximately 2.5 times of current RIP inertia by addition of flywheel on its main shaft. A full scale proving test of the inertia-increased RIP under actual plant operating conditions using full scale test loop was performed to evaluate vibration characteristics and coast down characteristics. From the results of this proving test, the validity of the new inertia-increased RIP and its power supply system (without M-G sets) was confirmed. (author)

Cutting technique for reactor internals by laser beam

International Nuclear Information System (INIS)

Matsumoto, O.; Sugihara, M.; Matsuda, K.; Miya, K.

1990-01-01

At present in Japan the verification tests on the commercial nuclear power reactor decommissioning technology are being conducted as the project of The Ministry of International Trade and Industry by Nuclear Power Engineering Test Center. This paper summarizes the interim results of the verification test for the reactor core internals decommissioning technology, which is being conducted from 1986 as a theme of the above project. All core internals to be studied here are made of stainless steel, and the maximum wall thickness is about 500mm (the maximum one to be cut is about 300mm) for the PWR's, and about 100mm for the BWR's. Though the plasma cutting, arc saw cutting method, etc. have been studied u p to now as the cutting technology for decommissioning these core internals, the authors are carrying out the development and verification test of the cutting technology with the laser beam, which is expected to increase its power in future and can be applied to various materials

Physical model study of neutron noise induced by vibration of reactor internals

International Nuclear Information System (INIS)

Liu Jinhui; Gu Fangyu

1999-01-01

The author presents a physical model of neutron noise induced by reactor internals vibration in frequency domain. Based on system control theory, the reactor dynamic equations are coupled with random vibration equation, and non-linear terms are also taken into accounted while treating the random vibration. Experiments carried out on a zero-power reactor show that the model can be used to describe dynamic character of neutron noise induced by internals' vibration. The model establishes a method to help to determine internals'vibration features, and to diagnosis anomalies through neutron noise

Abstracts and papers of the 1996 International RERTR Meeting

International Nuclear Information System (INIS)

1996-10-01

The 1996 International Meeting on Reduced Enrichment for Research and Test Reactors (RERTR) was sponsored by the Korea Atomic Energy Research Institute and the International Atomic Energy Agency and was held in Seoul, Korea, on 6-10 October 1996. The abstracts and available papers that were presented at this meeting are provided. The most important subject dealt with was progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program. The major events, findings, and activities of 1996 are reviewed with the results which the RERTR Program has achieved by the end of 1995 in collaboration of the ANL and its many international partners

Abstracts and papers of the 1996 International RERTR Meeting

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-10-01

The 1996 International Meeting on Reduced Enrichment for Research and Test Reactors (RERTR) was sponsored by the Korea Atomic Energy Research Institute and the International Atomic Energy Agency and was held in Seoul, Korea, on 6-10 October 1996. The abstracts and available papers that were presented at this meeting are provided. The most important subject dealt with was progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program. The major events, findings, and activities of 1996 are reviewed with the results which the RERTR Program has achieved by the end of 1995 in collaboration of the ANL and its many international partners.

Mathematical models in Slowpoke reactor internal irradiation site

International Nuclear Information System (INIS)

Raza, J.

2007-01-01

The main objective is to build representative mathematical models of neutron activation analysis in a Slowpoke internal irradiation site. Another significant objective is to correct various elements neutron activation analysis measured mass using these models. The neutron flux perturbation is responsible for the measured under-estimation of real masses. We supposed that neutron flux perturbation measurements taken during the Ecole Polytechnique de Montreal Slowpoke reactor first fuel loading were still valid after the second fuelling. .We also supposed that the thermal neutrons spatial and kinetic energies distributions as well as the absorption microscopic cross section dependence on the neutrons kinetic energies were important factors to satisfactorily represent neutron activation analysis results. In addition, we assumed that the neutron flux is isotropic in the laboratory system. We used experimental results from the Slowpoke reactor internal irradiation sites, in order to validate our mathematical models. Our models results are in close agreement with these experimental results..We established an accurate global mathematical correlation of the neutron flux perturbation in function of samples volumes and macroscopic neutron absorption cross sections. It is applicable to sample volumes ranging from 0,1 to 1,3 ml and macroscopic neutron absorption cross section up to 5 moles-b for seven (7) elements with atomic numbers (Z) ranging from 5 to 79. We first came up with a heuristic neutron transport mathematical semi-analytical model, in order to better understand neutrons behaviour in presence of one of several different nuclei samples volumes and mass. In order to well represent the neutron flux perturbation, we combined a neutron transport solution obtained from the spherical harmonics method of a finite cylinder and a mathematical expression combining two cylindrical harmonic functions..With the help of this model and the least squares method, we made extensive

An overview of reactor vessel internals segmentation for nuclear plant decommissioning

International Nuclear Information System (INIS)

Litka, T.J.

1994-01-01

Several nuclear plants have undergone reactor vessel (RV) internals segmentation as part of or in preparation for decommissioning the plant. In addition, several other nuclear facilities are planning for similar work efforts. The primary technology used for segmentation of RV internals, whether in-air or underwater is Plasma Arc Cutting (PAC). Metal Disintegration Machining (MDM) is also used for difficult to make cuts. PAC and MDM are deployed by various means including Long Handled Tools (LHTs), fixtures, tracks, and multi-axis manipulators. These enable remote cutting due to the radiation and/or underwater environment. A Boiling Water Reactor (BWR), a Pressurized Water Reactor (PWR), and a High Temperature Gas Reactor (HTGR) have had their internals removed and segmented using PAC and MDM. The cutting technology used for each component, location of cut, cut geometry and environment had to be determined well before the actual cutting operations. This allowed for the design, fabrication, and testing of the delivery systems. The technologies, selection process, and methodology for RV internals segmentation will be discussed in this paper

5. International workshop on autoionization phenomena in atoms. Abstracts

International Nuclear Information System (INIS)

Balashov, V.V.

1995-01-01

Summaries of the reports presented at the 5 International Workshop on Autoionization Phenomena in Atoms (Dubna, 12-14 December 1995). The main topics of these 53 reports are the following ones: photoexcitation of autoionizing states in atoms and ions, autoionization in electron-atom collisions, autoionization in heavy particle collisions, coincidence experiments in autoionization studies, investigations of autoionizing states with lasers and wave functions and decay characteristics of autoionizing states

Determination of the neutral oxygen atom density in a plasma reactor loaded with metal samples

Science.gov (United States)

Mozetic, Miran; Cvelbar, Uros

2009-08-01

The density of neutral oxygen atoms was determined during processing of metal samples in a plasma reactor. The reactor was a Pyrex tube with an inner diameter of 11 cm and a length of 30 cm. Plasma was created by an inductively coupled radiofrequency generator operating at a frequency of 27.12 MHz and output power up to 500 W. The O density was measured at the edge of the glass tube with a copper fiber optics catalytic probe. The O atom density in the empty tube depended on pressure and was between 4 and 7 × 1021 m-3. The maximum O density was at a pressure of about 150 Pa, while the dissociation fraction of O2 molecules was maximal at the lowest pressure and decreased with increasing pressure. At about 300 Pa it dropped below 10%. The measurements were repeated in the chamber loaded with different metallic samples. In these cases, the density of oxygen atoms was lower than that in the empty chamber. The results were explained by a drain of O atoms caused by heterogeneous recombination on the samples.

Nitride Coating Effect on Oxidation Behavior of Centrifugally Atomized U-Mo Fuel

Energy Technology Data Exchange (ETDEWEB)

Jeong, Yong Jin; Cho, Woo Hyoung; Park, Jong Man; Lee, Yoon Sang; Yang, Jae Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2011-10-15

Uranium metal and uranium compounds are being used as nuclear fuel materials and generally known as pyrophoric materials. Nowadays the importance of nuclear fuel about safety is being emphasized due to the vigorous exchanges and co-operations among the international community. According to the reduced enrichment for research and test reactors (RERTR) program, the international research reactor community has decided to use low-enriched uranium instead of high-enriched uranium. As a part of the RERTR program, KAERI has developed centrifugally atomized U-Mo alloys as a promising candidate of research reactor fuel. Kang et al. studied the oxidation behavior of centrifugally atomized U-10wt% Mo alloy and it showed better oxidation resistance than uranium. In this study, the oxidation behavior of nitride coated U-7wt% Mo alloy is investigated to enhance the safety against pyrophoricity

An Overview of the International Reactor Physics Experiment Evaluation Project

International Nuclear Information System (INIS)

Briggs, J. Blair; Gulliford, Jim

2014-01-01

Interest in high-quality integral benchmark data is increasing as efforts to quantify and reduce calculational uncertainties associated with advanced modeling and simulation accelerate to meet the demands of next generation reactor and advanced fuel cycle concepts. Two Organization for Economic Cooperation and Development (OECD) Nuclear Energy Agency (NEA) activities, the International Criticality Safety Benchmark Evaluation Project (ICSBEP), initiated in 1992, and the International Reactor Physics Experiment Evaluation Project (IRPhEP), initiated in 2003, have been identifying existing integral experiment data, evaluating those data, and providing integral benchmark specifications for methods and data validation for nearly two decades. Data provided by those two projects will be of use to the international reactor physics, criticality safety, and nuclear data communities for future decades. An overview of the IRPhEP and a brief update of the ICSBEP are provided in this paper.

International Working Group on Fast Reactors Second Annual Meeting. Summary Report

International Nuclear Information System (INIS)

1969-01-01

The Agenda of the Meeting was as follows: Opening of the meeting. 2. Appraisal of the IWGFB's activity for the period from the first annual meeting of the Group. 3. Comments on national programmes on fast breeder reactors. 4. Presentation of general findings and conclusions of national and regional meetings on fast reactor problems held in represented countries and international organisations last year. 5. Comments on the programmes of international meetings on fast reactors to be held in 1969. 6. Consideration of a schedule for meetings on fast reactors in 1970. 7. Suggestions for the topics and location of specialists' meetings in 1969-1970. 8. Suggestions for reviews and studies in the field of fast reactors. 9. The time and place of the third annual meeting of the IWGFR. 10. Closing of the meeting

Atoms for peace - the work of the International Atomic Energy Agency

International Nuclear Information System (INIS)

Daglish, J.

1984-01-01

The article deals with the International Atomic Energy Agency (IAEA), which is situated in Vienna. The aims of the IAEA and its work are described. The safeguards system; promotional work; technical cooperation programme; and nuclear safety work concerned with basic safety standards for radiation protection; are all discussed. (U.K.)

Vibration behavior of PWR reactor internals Model experiments and analysis

International Nuclear Information System (INIS)

Assedo, R.; Dubourg, M.; Livolant, M.; Epstein, A.

1975-01-01

In the late 1971, the CEA and FRAMATOME decided to undertake a comprehensive joint program of studying the vibration behavior of PWR internals of the 900 MWe, 50 cycle, 3 loop reactor series being built by FRAMATOME in France. The PWR reactor internals are submitted to several sources of excitation during normal operation. Two main sources of excitation may effect the internals behavior: the large flow turbulences which could generate various instabilities such as: vortex shedding: the pump pressure fluctuations which could generate acoustic noise in the circuit at frequencies corresponding to shaft speed frequencies or blade passing frequencies, and their respective harmonics. The flow induced vibrations are of complex nature and the approach selected, for this comprehensive program, is semi-empirical and based on both theoretical analysis and experiments on a reduced scale model and full scale internals. The experimental support of this program consists of: the SAFRAN test loop which consists of an hydroelastic similitude of a 1/8 scale model of a PWR; harmonic vibration tests in air performed on full scale reactor internals in the manufacturing shop; the GENNEVILLIERS facilities which is a full flow test facility of primary pump; the measurements carried out during start up on the Tihange reactor. This program will be completed in April 1975. The results of this program, the originality of which consists of studying separately the effects of random excitations and acoustic noises, on the internals behavior, and by establishing a comparison between experiments and analysis, will bring a major contribution for explaining the complex vibration phenomena occurring in a PWR

Thermohydraulic relationships for advanced water cooled reactors and the role of the IAEA

International Nuclear Information System (INIS)

Badulescu, A.; Groeneveld, D.C.

2000-01-01

Under the auspices of the International Atomic Energy Agency (IAEA) a Coordinated Research Program (CRP) on Thermohydraulic Relationships for Advanced Water-Cooled Reactors was carried out from 1995-1998. It was included into the IAEA's Programme following endorsement in 1995 by the International Working Group on Advanced Technologies for Water Cooled Reactors. The overall goal was to promote International Information exchange and cooperation in establishing a consistent set of thermohydraulic relationships that are appropriate for use in analyzing the performance and safety of advanced water-cooled reactors. (authors)

Design of a multipurpose research reactor

International Nuclear Information System (INIS)

Sanchez Rios, A.A.

1990-01-01

The availability of a research reactor is essential in any endeavor to improve the execution of a nuclear programme, since it is a very versatile tool which can make a decisive contribution to a country's scientific and technological development. Because of their design, however, many existing research reactors are poorly adapted to certain uses. In some nuclear research centres, especially in the advanced countries, changes have been made in the original designs or new research prototypes have been designed for specific purposes. These modifications have proven very costly and therefore beyond the reach of developing countries. For this reason, what the research institutes in such countries need is a single sufficiently versatile nuclear plant capable of meeting the requirements of a nuclear research programme at a reasonable cost. This is precisely what a multipurpose reactor does. The Mexican National Nuclear Research Institute (ININ) plans to design and build a multipurpose research reactor capable at the same time of being used for the development of reactor design skills and for testing nuclear materials and fuels, for radioisotopes production, for nuclear power studies and basic scientific research, for specialized training, and so on. For this design work on the ININ Multipurpose Research Reactor, collaborative relations have been established with various international organizations possessing experience in nuclear reactor design: Atomehnergoeksport of the USSR: Atomic Energy of Canada Limited (AECL); General Atomics (GA) of the USA; and Japan Atomic Energy Research Institute

Vibration analysis of reactor assembly internals for Prototype Fast Breeder Reactor

International Nuclear Information System (INIS)

Chellapandi, P.; Jalaldeen, S.; Srinivasan, R.; Chetal, S.C.; Bhoje, S.B.

2003-01-01

Vibration analysis of the reactor assembly components of 500 MWe Prototype Fast Breeder Reactor (PFBR) is presented. The vibration response of primary pump as well as dynamic forces developed at its supports are predicted numerically. The stiffness properties of hydrostatic bearing are determined by formulating and solving governing fluid and structural mechanics equations. The dynamic forces exerted by pump are used as input data for the dynamic response of reactor assembly components, mainly inner vessel, thermal baffle and control plug. Dynamic response of reactor assembly components is also predicted for the pressure fluctuations caused by sodium free level oscillations. Thermal baffle (weir shell) which is subjected to fluid forces developed at the associated sodium free levels is analysed by formulating and solving a set of non-linear equations for fluids, structures and fluid structure interaction (FSI). The control rod drive mechanism is analysed for response under flow induced forces on the parts subjected to cross flow in the zone just above the core top, taking into account FSI between sheaths of control and safety rod and absorber pin bundle. Based on the analysis results, it is concluded that the reactor assembly internals are free from any risk of mechanical as well as flow induced vibrations. (author)

Agreement Between the International Atomic Energy Agency, the Government of Jamaica and the Government of the United States of America for Assistance in Securing Low Enriched Uranium for a Research Reactor

International Nuclear Information System (INIS)

2014-01-01

The text of the Agreement between the International Atomic Energy Agency, the Government of Jamaica and the Government of the United States of America for Assistance in Securing Low Enriched Uranium for a Research Reactor is reproduced in this document for the information of all Members of the Agency. The Agency's Board of Governors approved the text of the Agreement on 6 March 2013. The Agreement was signed by the authorized representatives of Jamaica on 25 November 2013, the United States on 2 May 2013 and the Director General of the IAEA on 16 December 2013. Pursuant to the Article XI of the Agreement, the Agreement entered into force on 16 December 2013, upon signature by the Director General of the IAEA and by the authorized representatives of Jamaica and the United States [fr

Agreement Between the International Atomic Energy Agency, the Government of Jamaica and the Government of the United States of America for Assistance in Securing Low Enriched Uranium for a Research Reactor

International Nuclear Information System (INIS)

2014-01-01

The text of the Agreement between the International Atomic Energy Agency, the Government of Jamaica and the Government of the United States of America for Assistance in Securing Low Enriched Uranium for a Research Reactor is reproduced in this document for the information of all Members of the Agency. The Agency's Board of Governors approved the text of the Agreement on 6 March 2013. The Agreement was signed by the authorized representatives of Jamaica on 25 November 2013, the United States on 2 May 2013 and the Director General of the IAEA on 16 December 2013. Pursuant to the Article XI of the Agreement, the Agreement entered into force on 16 December 2013, upon signature by the Director General of the IAEA and by the authorized representatives of Jamaica and the United States

Agreement Between the International Atomic Energy Agency, the Government of Jamaica and the Government of the United States of America for Assistance in Securing Low Enriched Uranium for a Research Reactor

International Nuclear Information System (INIS)

2014-01-01

The text of the Agreement between the International Atomic Energy Agency, the Government of Jamaica and the Government of the United States of America for Assistance in Securing Low Enriched Uranium for a Research Reactor is reproduced in this document for the information of all Members of the Agency. The Agency's Board of Governors approved the text of the Agreement on 6 March 2013. The Agreement was signed by the authorized representatives of Jamaica on 25 November 2013, the United States on 2 May 2013 and the Director General of the IAEA on 16 December 2013. Pursuant to the Article XI of the Agreement, the Agreement entered into force on 16 December 2013, upon signature by the Director General of the IAEA and by the authorized representatives of Jamaica and the United States [es

Status of international cooperation in nuclear technology on testing/research reactors between JAEA and INP-NNC

International Nuclear Information System (INIS)

Kawamura, Hiroshi; Tsuchiya, Kunihiko; Takemoto, Noriyuki; Kimura, Akihiro; Tanimoto, Masataka; Izumo, Hironobu; Chakrov, Petr; Gizatulin, Shamil; Chakrova, Yelena; Ludmila, Chkushuina; Asset, Shaimerdenov; Nataliya, Romanova

2012-02-01

Based on the implementing arrangement between National Nuclear Center of the Republic of Kazakhstan (NNC) and the Japan Atomic Energy Agency (JAEA) for 'Nuclear Technology on Testing/Research Reactors' in cooperation in Research and Development in Nuclear Energy and Technology, four specific topics of cooperation (STC) have been carried out from June, 2009. Four STCs are as follows; (1) STC No.II-1 : International Standard of Instrumentation. (2) STC No.II-2 : Irradiation Technology of RI Production. (3) STC No.II-3 : Lifetime Expansion of Beryllium Reflector. (4) STC No.II-4 : Irradiation Technology for NTD-Si. The information exchange, personal exchange and cooperation experiments are carried out under these STCs. The status in the field of nuclear technology on testing/research reactors in the implementing arrangement is summarized, and future plans of these specific topics of cooperation are described in this report. (author)

Boiling-Water Reactor internals aging degradation study

International Nuclear Information System (INIS)

Luk, K.H.

1993-09-01

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

Advances in fast reactor technology. Proceedings of the 30. meeting of the International Working Group on Fast Reactors

International Nuclear Information System (INIS)

1998-04-01

Individual States were largely responsible for early developments in experimental and prototype liquid metal fast reactors (LMFRs). However, for development of advanced LMFRs, international co-operation plays an important role. The IAEA seeks to promote such co-operation. For R and D incorporating innovative features, international co-operation allows pooling of resources and expertise in areas of common interest. Information on experience gained from R and D, and from the operation and construction of fast reactors, has been reviewed periodically by the International Working Group on Fast Reactors (IWGFR). These proceedings contain updated a new information on the status of LMFR development, as reported at the 30th meeting of the IWGFR, held in Beijing, China, from 13 to 16 May 1997

Advances in fast reactor technology. Proceedings of the 30. meeting of the International Working Group on Fast Reactors

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-04-01

Individual States were largely responsible for early developments in experimental and prototype liquid metal fast reactors (LMFRs). However, for development of advanced LMFRs, international co-operation plays an important role. The IAEA seeks to promote such co-operation. For R and D incorporating innovative features, international co-operation allows pooling of resources and expertise in areas of common interest. Information on experience gained from R and D, and from the operation and construction of fast reactors, has been reviewed periodically by the International Working Group on Fast Reactors (IWGFR). These proceedings contain updated a new information on the status of LMFR development, as reported at the 30th meeting of the IWGFR, held in Beijing, China, from 13 to 16 May 1997. Refs,figs,tabs.

International conference on Asian nuclear prospects 2010

International Nuclear Information System (INIS)

2010-10-01

The proceedings of the second international conference on Asian nuclear prospects was held during October 10-13, 2010. The topics covered were atomic minerals exploration, fuels and fuel fabrication, fast reactor engineering and technology, thermal reactors, fuel reprocessing, materials development, nuclear instrumentation and electronics, reactor and environmental safety, waste management, knowledge management and HRD. Papers relevant to INIS are indexed separately

Repair and preventive maintenance technology for BWR reactor internals and piping

International Nuclear Information System (INIS)

Ootsubo, Tooru; Itou, Takashi; Sakashita, Akihiro

2009-01-01

Stress corrosion cracking of welding portion has found in many domestic and foreign BWR reactor internals and Primary Loop Recirculation piping. Also, repair and preventive maintenance technologies for SCC has been developed and/or adopted to BWRs in recent years. This paper introduces the sample of these technologies, such as seal-welding for SCC on BWR reactor internals, preventive maintenance technology for PLR piping such as Corrosion Resistant Cladding, Internal Polishing and Induction Heating Stress Improvement. These technologies are introduced on 'E-Journal of Advanced Maintenance', which is an international journal on a exclusive website of Japan Society of Maintenology. (author)

The international thermonuclear reactor (ITER)

International Nuclear Information System (INIS)

Fowler, T.K.; Henning, C.D.

1987-01-01

Four governmental groups, representing Europe, Japan, USSR and U.S. met in March 1987 to consider a new international design of a magnetic fusion device for the 1990's. An interim group was appointed. The author gives a brief synopsis of what might be thought of as a draft charter. The starting point is the objective of the ITER device, which is summarized as demonstrating both scientific and technical feasibility of fusion. The paper presents an update on the current thinking and technical aspects for the International Thermonuclear Experimental Reactor (ITER). This covers not only what is happening in the U.S. but also some reports of preliminary thinking of the last technical work that occurred in Vienna

Operation and maintenance experience at the General Atomic Company's TRIGA reactor facility at San Diego, California

International Nuclear Information System (INIS)

Whittemore, W.L.; Stout, W.A.; Shoptaugh, J.R.; Chesworth, R.H.

1982-01-01

Since the startup of the original 250 kW TRIGA Mark I reactor in 1958, General Atomic Company has accumulated nearly 24 years of operation and maintenance experience with this type of reactor. In addition to the nearly 24 years of experience gained on the Mark I, GA has operated the 1.5 MW Advanced Prototype Test Reactor (Mark F) for 22 years and operated a 2 MW below-ground TRIGA Mark III for five years. Information obtained from normal and abnormal operation are presented. (author)

Maintenance technologies for reactor internals

Energy Technology Data Exchange (ETDEWEB)

Sato, Kenji [Nuclear Energy Systems and Services Div., Toshiba Corp., Tokyo (Japan); Kobayashi, Masahiro [Toshiba Corp., Yokohama (Japan). Keihin Product Operations; Sano, Yuji; Kimura, Seiichiro [Power and Industrial Systems Research and Development Center, Toshiba Corp., Tokyo(Japan)

2000-10-01

Toshiba places the highest priority on maintenance technologies for the reactor pressure vessel (RPV) and its internals in operating nuclear power plants. This paper summarizes the status of applied laser maintenance technologies, both preventive and repair. For laser peeing and laser desensitization treatment (LDT) technologies in particular, field applications are also described in detail. In the future, the area of field applications for preventive maintenance, repair, and inspection technologies will be further expanded. (author)

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

International Nuclear Information System (INIS)

Kim, Tae Ryong; Jeong, Seong Ho; Park, Jin Ho; Park, Jin Suk

1994-08-01

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

Status of research reactor spent fuel world-wide

International Nuclear Information System (INIS)

Ritchie, I.G.

2004-01-01

Results compiled in the research reactor spent fuel database are used to assess the status of research reactor spent fuel world-wide. Fuel assemblies, their types, enrichment, origin of enrichment and geological distribution among the industrialised and developed countries of the world are discussed. Fuel management practices in wet and dry storage facilities and the concerns of reactor operators about long-term storage of their spent fuel are presented and some of the activities carried out by the International Atomic Energy Agency to address the issues associated with research reactor spent fuel are outlined. (author)

Passive cooling of a fixed bed nuclear reactor

International Nuclear Information System (INIS)

Petry, V.J.; Bortoli, A.L. de; Sefidwash, F.

2005-01-01

Small nuclear reactors without the need for on-site refuelling have greater simplicity, better compliance with passive safety systems, and are more adequate for countries with small electric grids and limited investment capabilities. Here the passive cooling characteristic of the fixed bed nuclear reactor (FBNR), that is being developed under the International Atomic Energy Agency (IAEA) Coordinated Research Project, is studied. A mathematical model is developed to calculate the temperature distribution in the fuel chamber of the reactor. The results demonstrate the passive cooling of this nuclear reactor concept. (authors)

Fourteenth Annual Meeting of the International Working Group on Past Reactors. Summary Report. Part I

International Nuclear Information System (INIS)

1981-11-01

The Fourteenth Annual Meeting of the IAEA-IWGFR was held in accordance with the recommendations of the previous Annual Group Meeting, at the Vienna International Centre, Vienna from 31 March to 3 April 1981. All Member States of the group were represented at the meeting: France, the Federal Republic of Germany, India, Italy, Japan, the Union of Soviet Socialist Republics, the United Kingdom and the United States of America. The meeting was also attended by representatives from the Commission of European Communities, the Nuclear Energy Agency of the Organisation for Economic Co-operation and Development, the International Atomic Energy Agency and observers from Switzerland. The Agenda of the Meeting was as follows: 1. Review of IWGFR activities; 2. Consideration of future method of operation of the IWGFR; 3. Consideration of Conferences on Fast Reactors; 4. Consideration of the major recommendations of some of the IWGFR specialists' meetings for which the support of the IWGFR is requested; 5. Consideration of a schedule for specialists' meetings in 1981-1982; 6. Presentations and discussions on national programmes on fast breeder reactors.; 7. Recommendation of the IWGFR regarding a request of Switzerland concerning participation in the IWGFR; 8. The date and place of the Fifteenth Annual Meeting of the IWGFR

Baffle-former arrangement for nuclear reactor vessel internals

International Nuclear Information System (INIS)

Machado, O.J.; Berringer, R.T.

1978-01-01

A baffle-former arrangement for the reactor vessel internals of a nuclear reactor is described. The arrangement includes positioning of formers at the same elevations as the fuel assembly grids, and positioning flow holes in the baffle plates directly beneath selected former grid elevations. The arrangement reduces detrimental cross flows, maintains proper core barrel and baffle temperatures, and alleviates the potential of overpressurization within the baffle-former assembly under assumed major accident conditions

Sloshing response of a reactor tank with internals

International Nuclear Information System (INIS)

Ma, D.C.; Gvildys, J.; Chang, Y.W.

1984-01-01

The sloshing response of a large reactor tank with in-tank components is presented. The study indicates that the presence of the internal components can significantly change the dynamic characteristics of the sloshing motion. The sloshing frequency of a tank with internals is considerably higher than that of a tank without internal. The higher sloshing frequency reduces the sloshing wave height on the free-surface but increases the dynamic pressure in the fluid

Plant maintenance and advanced reactors, 2006

Energy Technology Data Exchange (ETDEWEB)

Agnihotri, Newal (ed.)

2006-09-15

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

Conversion, core redesign and upgrade of the Rhode Island Atomic Energy Commission Reactor

International Nuclear Information System (INIS)

DiMeglio, A.F.

1987-01-01

The 2 MW Rhode Island Atomic Energy Commission reactor is required to convert from the use of High Enriched Uranium (HEU) fuel to the use of Low Enriched Uranium (LEU) fuel using a standard LEU fuel plate which is thinner and contains more Uranium-235 than the current HEU plate. These differences, coupled with the fact that the conversion should be accomplished without serious degradation of reactor characteristics and capability, has resulted in core design studies and thermal hydraulic studies not only at the current 2 MW but also at the maximum power level of the reactor, 5 MW. In addition, during the course of its 23 years of operation, it has become clear that the main uses of the reactor are neutron scattering and neutron activation analysis. The requirement to convert to LEU presents an opportunity during the conversion to optimize the core for the utilization and to restudy the thermal hydraulics using modern techniques. This paper will present the preliminary conclusions of both aspects. (Author)

Development of inspection and evaluation guidelines for light water reactor internals

International Nuclear Information System (INIS)

Aoki, T.; Yamashita, H.; Sakai, K.

2002-01-01

Full text: In Japan, before a nuclear power plant reaches its 30 years of operation, the Japanese utilities carry out a 'study on plant life management'. Reflecting the results of that study into the maintenance, the utilities make efforts to maintain and improve the safety and reliability of their nuclear power plants. In this study, all safety related components are evaluated from the viewpoint of aging degradation, assuming a long-term operation. If a crack should be found at components such as reactor internals, which is deemed important for safety and are difficult to repair or replace may provide a serious impact on the plant operation and management. Reactor internals, for instance, made of austenitic stainless steel and nickel base alloy, are not completely free from aging degradation including stress corrosion cracking (SCC). Therefore, it is concluded in the study on plant life management that they are required continuous planned inspections to confirm their integrity while continuing plant operation. If an aging degradation such as SCC should be found at the reactor internals, a great amount of labor and time may be required for root cause investigation and analysis and subsequent repairs because it is very difficult to reach to the degraded portion due to structural, dimensional and environmental restrictions. Therefore, such situations may provide serious impact on plant operation and management. As to the reactor internals and other components with similar characteristics, it is strongly recommended that contingency plan should be prepared in advance. Thus, considering the significance of aging degradation, it is necessary to develop some standard rules for the inspection and evaluation of reactor internals. Such rules should specify when, where and how to inspect. They should also specify the evaluation method in case such degradation as a crack is found, and the repair method and extent if repairs are required. These standard rules must be

HTGR Metallic Reactor Internals Core Shell Cutting & Machining Antideformation Technique Study

International Nuclear Information System (INIS)

Xing Huiping; Xue Song

2014-01-01

The reactor shell assembly of HTGR nuclear power station demonstration project metallic reactor internals is key components of reactor, remains with high-precision large component with large-sized thin-walled straight cylinder-shaped structure, and is the first manufacture in China. As compared with other reactor shell, it has a larger ID (Φ5360mm), a longer length (19000mm), a smaller wall thickness (40mm) and a higher precision requirement. During the process of manufacture, the deformation due to cutting & machining will directly affect the final result of manufacture, the control of structural deformation and cutting deformation shall be throughout total manufacture process of such assembly. To realize the control of entire core shell assembly geometry, the key is to innovate and make breakthroughs on anti-deformation technique and then provide reliable technological foundations for the manufacture of HTGR metallic reactor internals. (author)

In-reactor behaviour of centrifugally atomized U3Si dispersion fuel irradiated at high temperature in HANARO

International Nuclear Information System (INIS)

Kim, Ki Hwan; Park, Jong Man; Yoo, Byeong Ok; Park, Dae Kyu; Lee, Choong Sung; Kim, Chang Kyu

2002-01-01

The irradiation test on full-size U 3 Si dispersion fuel elements, prepared by centrifugal atomization and conventional comminution method, has been performed up to about 77 at.% U-235 in maximum burn-up at CT hole position having the highest power condition in the HANARO reactor, in order to examine the irradiation performance of the atomized U 3 Si for the driver fuels of HANARO. The in-reactor interaction of the atomized U 3 Si dispersion fuel meats is generally assumed to be acceptable with the range of 5-15 μm in average thickness. The atomized spherical particles have more uniform and thinner reaction layer than the comminuted irregular particles. The U 3 Si particles have relatively fine and uniform size distribution of fission gas bubbles, irrespective of the powdering method. The bubble population in the atomized particles appears to be finer and more homogeneous with the characteristics of narrower bubble size distribution than that of the comminuted fuel. The atomized U 3 Si dispersion fuel elements exhibit sound swelling behaviours of 5 % in ΔV/V m even at ∼77 at.% U-235 burn-up, which meets with the safety criterion of the fuel rod, 20vol.% for HANARO. The atomized U3Si dispersion fuel elements show smaller swelling than the comminuted fuel elements

Role of plasma enhanced atomic layer deposition reactor wall conditions on radical and ion substrate fluxes

Energy Technology Data Exchange (ETDEWEB)

Sowa, Mark J., E-mail: msowa@ultratech.com [Ultratech/Cambridge NanoTech, 130 Turner Street, Building 2, Waltham, Massachusetts 02453 (United States)

2014-01-15

Chamber wall conditions, such as wall temperature and film deposits, have long been known to influence plasma source performance on thin film processing equipment. Plasma physical characteristics depend on conductive/insulating properties of chamber walls. Radical fluxes depend on plasma characteristics as well as wall recombination rates, which can be wall material and temperature dependent. Variations in substrate delivery of plasma generated species (radicals, ions, etc.) impact the resulting etch or deposition process resulting in process drift. Plasma enhanced atomic layer deposition is known to depend strongly on substrate radical flux, but film properties can be influenced by other plasma generated phenomena, such as ion bombardment. In this paper, the chamber wall conditions on a plasma enhanced atomic layer deposition process are investigated. The downstream oxygen radical and ion fluxes from an inductively coupled plasma source are indirectly monitored in temperature controlled (25–190 °C) stainless steel and quartz reactors over a range of oxygen flow rates. Etch rates of a photoresist coated quartz crystal microbalance are used to study the oxygen radical flux dependence on reactor characteristics. Plasma density estimates from Langmuir probe ion saturation current measurements are used to study the ion flux dependence on reactor characteristics. Reactor temperature was not found to impact radical and ion fluxes substantially. Radical and ion fluxes were higher for quartz walls compared to stainless steel walls over all oxygen flow rates considered. The radical flux to ion flux ratio is likely to be a critical parameter for the deposition of consistent film properties. Reactor wall material, gas flow rate/pressure, and distance from the plasma source all impact the radical to ion flux ratio. These results indicate maintaining chamber wall conditions will be important for delivering consistent results from plasma enhanced atomic layer deposition

PREFACE: Fourth International Symposium on Atomic Technology

Science.gov (United States)

Okada, Shigefumi

2010-04-01

The International Symposium on Atomic Technology (ISAT) is held every year. The 4th Symposium (ISAT-4) was held on November 18-19, 2009 at the Seaside Hotel MAIKO VILLA KOBE, Kobe City, Japan presided by the "Atomic Technology Project". The ISAT-4 symposium was intended to offer a forum for the discussion on the latest progress in the atomic technologies. The symposium was attended by 107 delegates. There were 10 invited and 6 oral presentations. The number of poster presentations was 69. From all the contributions, 22 papers selected through review process are contained in this volume. The "Atomic Technology Project" was started in 2006 as a joint project of three institutions; (1) the Center for Atomic and Molecular Technologies, Graduate School of Engineering, Osaka University (CAMT), (2) the Tsukuba Research Center for Interdisciplinary Materials Science, Graduate School of Pure and Applied Sciences, University of Tsukuba (TIMS) and (3) the Polyscale Technology Research Center, Research Institute for Science and Technology, Tokyo University of Science (PTRC), each of which were independently pursuing nano-technologies and was developing atomic scale operation and diagnostics, functional materials, micro processing and device. The project is funded by the Ministry of Education, Culture, Sports, Science and Technology of Japan. The goal of the project is to contribute to the development of atomic-scale science and technologies such as functional molecules, biomaterials, and quantum functions of atomic-scale structures. Shigefumi Okada Conference Chair Center for Atomic and Molecular Technologies, Graduate School of Engineering, Osaka University, 2-1 Yamadaoka, Suita-city, Osaka 565-0871, Japan. Conference photograph Kobe photograph

Annual Report to Congress of the Atomic Energy Commission for 1967

Energy Technology Data Exchange (ETDEWEB)

Seaborg, Glenn T.

1968-01-31

The document represents the 1967 Annual Report of the Atomic Energy Commission (AEC) to Congress. The report opens with a Summary of Atomic Energy Programs in 1967 and includes 17 Chapters, 11 appendices and an index. Chapters are as follows: (1) Source and Special Nuclear Materials; (2) Safeguards and Materials Management; (3) The Nuclear Defense Effort; (4) Naval Propulsion Reactors; (5) Reactor Development and Technology; (6) Licensing and Regulating the Atom; (7) Operational Safety; (8) Nuclear Rocket Propulsion; (9) Specialized Nuclear Power Units; (10) Isotopic Radiation Applications; (11) The Plowshare Program; (12) International Cooperation Activities; (13) Informational Activities; (14) Nuclear Education and Training; (15) Basic Research; (16) Industrial Participation Aspects; and, (17) Administrative and Management Matters.

Dynamic Modeling for the Design and Cyclic Operation of an Atomic Layer Deposition (ALD Reactor

Directory of Open Access Journals (Sweden)

Curtisha D. Travis

2013-08-01

Full Text Available A laboratory-scale atomic layer deposition (ALD reactor system model is derived for alumina deposition using trimethylaluminum and water as precursors. Model components describing the precursor thermophysical properties, reactor-scale gas-phase dynamics and surface reaction kinetics derived from absolute reaction rate theory are integrated to simulate the complete reactor system. Limit-cycle solutions defining continuous cyclic ALD reactor operation are computed with a fixed point algorithm based on collocation discretization in time, resulting in an unambiguous definition of film growth-per-cycle (gpc. A key finding of this study is that unintended chemical vapor deposition conditions can mask regions of operation that would otherwise correspond to ideal saturating ALD operation. The use of the simulator for assisting in process design decisions is presented.

Experience relevant to safety obtained from reactor decommissioning operations in the French Atomic Energy Commission

International Nuclear Information System (INIS)

Giraudel, B.; Langlois, G.

1979-01-01

From among the nuclear facilities constructed in France the authors cite eight large reactors, ranging from critical assemblies to power reactors, that have been finally shut-down since 1965. A brief account is given of the way in which the various operations were carried out after the final control rod drop, a distinction being drawn between the shut-down proper and the containment and dismantling work. A description is also given, from the technical and regulatory standpoint, of the final stage attained, and mention is made of French safety arrangements and of the part played by the safety services during decommissioning operations. Among the lessons derived from French experience, the authors mention the completion of operations without any serious safety problems, and with guarantees for the protection of personnel and the population as a whole, by conventional techniques; the advantage of planning decommissioning operations from the very beginning of construction of the facilities; and the importance of filing descriptive documents. In view of the experience gained, the French Atomic Energy Commission has devised internal procedures for facilitating the application of regulations governing the shut-down and decommissioning phases, which are aimed at preserving surveillance procedures similar to those in force during normal operation. (author)

International Reactor Physics Experiment Evaluation (IRPhE) Project. IRPhE Handbook - 2015 edition

International Nuclear Information System (INIS)

Bess, John D.; Gullifor, Jim

2015-03-01

The purpose of the International Reactor Physics Experiment Evaluation (IRPhE) Project is to provide an extensively peer-reviewed set of reactor physics-related integral data that can be used by reactor designers and safety analysts to validate the analytical tools used to design next-generation reactors and establish the safety basis for operation of these reactors. This work of the IRPhE Project is formally documented in the 'International Handbook of Evaluated Reactor Physics Benchmark Experiments', a single source of verified and extensively peer-reviewed reactor physics benchmark measurements data. The evaluation process entails the following steps: Identify a comprehensive set of reactor physics experimental measurements data, Evaluate the data and quantify overall uncertainties through various types of sensitivity analysis to the extent possible, verify the data by reviewing original and subsequently revised documentation, and by talking with the experimenters or individuals who are familiar with the experimental facility, Compile the data into a standardized format, Perform calculations of each experiment with standard reactor physics codes where it would add information, Formally document the work into a single source of verified and peer reviewed reactor physics benchmark measurements data. The International Handbook of Evaluated Reactor Physics Benchmark Experiments contains reactor physics benchmark specifications that have been derived from experiments that were performed at nuclear facilities around the world. The benchmark specifications are intended for use by reactor designers, safety analysts and nuclear data evaluators to validate calculation techniques and data. Example calculations are presented; these do not constitute a validation or endorsement of the codes or cross-section data. The 2015 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments contains data from 143 experimental series that were

International cooperation on breeder reactors

International Nuclear Information System (INIS)

Gray, J.E.; Kratzer, M.B.; Leslie, K.E.; Paige, H.W.; Shantzis, S.B.

1978-01-01

In March 1977, as the result of discussions which began in the fall of 1976, the Rockefeller Foundation requested International Energy Associates Limited (IEAL) to undertake a study of the role of international cooperation in the development and application of the breeder reactor. While there had been considerable international exchange in the development of breeder technology, the existence of at least seven major national breeder development programs raised a prima facie issue of the adequacy of international cooperation. The final product of the study was to be the identification of options for international cooperation which merited further consideration and which might become the subject of subsequent, more detailed analysis. During the course of the study, modifications in U.S. breeder policy led to an expansion of the analysis to embrace the pros and cons of the major breeder-related policy issues, as well as the respective views of national governments on those issues. The resulting examination of views and patterns of international collaboration emphasizes what was implicit from the outset: Options for international cooperation cannot be fashioned independently of national objectives, policies and programs. Moreover, while similarity of views can stimulate cooperation, this cannot of itself provide compelling justification for cooperative undertakings. Such undertakings are influenced by an array of other national factors, including technological development, industrial infrastructure, economic strength, existing international ties, and historic experience

Experience with mechanical segmentation of reactor internals

International Nuclear Information System (INIS)

Carlson, R.; Hedin, G.

2003-01-01

Operating experience from BWE:s world-wide has shown that many plants experience initial cracking of the reactor internals after approximately 20 to 25 years of service life. This ''mid-life crisis'', considering a plant design life of 40 years, is now being addressed by many utilities. Successful resolution of these issues should give many more years of trouble-free operation. Replacement of reactor internals could be, in many cases, the most favourable option to achieve this. The proactive strategy of many utilities to replace internals in a planned way is a market-driven effort to minimize the overall costs for power generation, including time spent for handling contingencies and unplanned outages. Based on technical analyses, knowledge about component market prices and in-house costs, a cost-effective, optimized strategy for inspection, mitigation and replacements can be implemented. Also decommissioning of nuclear plants has become a reality for many utilities as numerous plants worldwide are closed due to age and/or other reasons. These facts address a need for safe, fast and cost-effective methods for segmentation of internals. Westinghouse has over the last years developed methods for segmentation of internals and has also carried out successful segmentation projects. Our experience from the segmentation business for Nordic BWR:s is that the most important parameters to consider when choosing a method and equipment for a segmentation project are: - Safety, - Cost-effectiveness, - Cleanliness, - Reliability. (orig.)

Hydrodynamic study of an internal airlift reactor for microalgae culture.

Science.gov (United States)

Rengel, Ana; Zoughaib, Assaad; Dron, Dominique; Clodic, Denis

2012-01-01

Internal airlift reactors are closed systems considered today for microalgae cultivation. Several works have studied their hydrodynamics but based on important solid concentrations, not with biomass concentrations usually found in microalgae cultures. In this study, an internal airlift reactor has been built and tested in order to clarify the hydrodynamics of this system, based on microalgae typical concentrations. A model is proposed taking into account the variation of air bubble velocity according to volumetric air flow rate injected into the system. A relationship between riser and downcomer gas holdups is established, which varied slightly with solids concentrations. The repartition of solids along the reactor resulted to be homogenous for the range of concentrations and volumetric air flow rate studied here. Liquid velocities increase with volumetric air flow rate, and they vary slightly when solids are added to the system. Finally, liquid circulation time found in each section of the reactor is in concordance with those employed in microalgae culture.

Development of an internally cooled annular fuel bundle for pressurized heavy water reactors

Energy Technology Data Exchange (ETDEWEB)

Hamilton, H.; Armstrong, J.; Kittmer, A.; Zhuchkova, A.; Xu, R.; Hyland, B.; King, M.; Nava-Dominguez, A.; Livingstone, S.; Bergeron, A. [Atomic Energy of Canada, Ltd., Chalk River Laboratories, Chalk River, ON (Canada)

2013-07-01

A number of preliminary studies have been conducted at Atomic Energy of Canada Limited to explore the potential of using internally cooled annular fuel (ICAF) in CANDU reactors including finite element thermo-mechanical modelling, reactor physics, thermal hydraulics, fabrication and mechanical design. The most compelling argument for this design compared to the conventional solid-rod design is the significant reduction in maximum fuel temperature for equivalent LERs (linear element ratings). This feature presents the potential for power up-rating or higher burnup and a decreased defect probability due to in-core power increases. The thermal-mechanical evaluation confirmed the significant reduction in maximum fuel temperatures for ICAF fuel compared to solid-rod fuel for equivalent LER. The maximum fuel temperature increase as a function of LER increase is also significantly less for ICAF fuel. As a result, the sheath stress induced by an equivalent power increase is approximately six times less for ICAF fuel than solid-rod fuel. This suggests that the power-increase thresholds to failure (due to stress-corrosion cracking) for ICAF fuel should be well above those for solid-rod fuel, providing improvement in operation flexibility and safety.

The qualification of reactor operators

International Nuclear Information System (INIS)

Lima, J.M. de; Soares, H.V.

1981-01-01

The qualification and performance of nuclear power personnel have an important influence on the availability and safety operation of these plants. This paper describes the Brazilian rules and norms established by the CNEN-Brazilian Atomic Energy Comission, as well as policy of other countries concerning training requirements and experiences of nuclear power reactor operators. Some coments are made about the im pact of the march 1979 Three Mile Island accident on upgrading the reactor training requirements in U.S.A. and its international implication. (Author) [pt

IAEA activities on research reactor safety

International Nuclear Information System (INIS)

Alcala-Ruiz, F.

1995-01-01

Since its inception in 1957, the International Atomic Energy Agency (IAEA) has included activities in its programme to address aspects of research reactors such as safety, utilization and fuel cycle considerations. These activities were based on statutory functions and responsibilities, and on the current situation of research reactors in operation around the world; they responded to IAEA Member States' general or specific demands. At present, the IAEA activities on research reactors cover the above aspects and respond to specific and current issues, amongst which safety-related are of major concern to Member States. The present IAEA Research Reactor Safety Programme (RRSP) is a response to the current situation of about 300 research reactors in operation in 59 countries around the world. (orig.)

DESIGN OF A VIBRATION AND STRESS MEASUREMENT SYSTEM FOR AN ADVANCED POWER REACTOR 1400 REACTOR VESSEL INTERNALS COMPREHENSIVE VIBRATION ASSESSMENT PROGRAM

OpenAIRE

KO, DO-YOUNG; KIM, KYU-HYUNG

2013-01-01

In accordance with the US Nuclear Regulatory Commission (US NRC), Regulatory Guide 1.20, the reactor vessel internals comprehensive vibration assessment program (RVI CVAP) has been developed for an Advanced Power Reactor 1400 (APR1400). The purpose of the RVI CVAP is to verify the structural integrity of the reactor internals to flow-induced loads prior to commercial operation. The APR1400 RVI CVAP consists of four programs (analysis, measurement, inspection, and assessment). Thoughtful prepa...

Specifications of the International Atomic Energy Agency's international project on safety assessment driven radioactive waste management solutions

International Nuclear Information System (INIS)

Ghannadi, M.; Asgharizadeh, F.; Assadi, M. R.

2008-01-01

Radioactive waste is produced in the generation of nuclear power and the production and use of radioactive materials in the industry, research, and medicine. The nuclear waste management facilities need to perform a safety assessment in order to ensure the safety of a facility. Nuclear safety assessment is a structured and systematic way of examining a proposed facility, process, operation and activity. In nuclear waste management point of view, safety assessment is a process which is used to evaluate the safety of radioactive waste management and disposal facilities. In this regard the International Atomic Energy Agency is planed to implement an international project with cooperation of some member states. The Safety Assessment Driving Radioactive Waste Management Solutions Project is an international programme of work to examine international approaches to safety assessment in aspects of p redisposal r adioactive waste management, including waste conditioning and storage. This study is described the rationale, common aspects, scope, objectives, work plan and anticipated outcomes of the project with refer to International Atomic Energy Agency's documents, such as International Atomic Energy Agency's Safety Standards, as well as the Safety Assessment Driving Radioactive Waste Management Solutions project reports

International collaboration between nuclear research centres and the role of research reactors

International Nuclear Information System (INIS)

Dodd, B.

2001-01-01

A research reactor is a core facility in many nuclear research centres (NRCs) of Member States and it is logical that it should be the focus of any international collaboration between such centres. There are several large and sophisticated research reactors in operation in both developed and developing Member States, such as Belgium, China, Egypt, France, Hungary, Indonesia, India, Japan, ROK, Netherlands, South Africa and the USA. There are also several new, large reactors under construction or being planned such as those in Australia, Canada, China, France, Germany, and Thailand. It is felt that the utilization of these reactors can be enhanced by international co-operation to achieve common goals in research and applications. (author)

International standardization of nuclear reactor designs - the way forward

International Nuclear Information System (INIS)

Raetzke, Christian

2010-01-01

The concept of 'International Standardization of Nuclear Reactor Designs' means that vendors could build their designs in every country without having to adapt it specifically to national safety requirements. Such standardization would have two main effects. It would greatly facilitate nuclear new build worldwide by giving greater efficiency and certainty to the national licensing procedures; by taking into account the fact that vendors, and nowadays also utilities, are active across borders; by helping developing countries to establish their nuclear new build programmes; and by reducing the strain on human resources on both the regulators' and the industry's side. The second valuable effect of standardization would be to further enhance safety by improving the exchange of construction and operating experience among a number of reactors belonging to fleets of the same design. The World Nuclear Association's CORDEL (Cooperation in Reactor Design Evaluation and Licensing) Group has developed a concept for implementation of international standardization of reactor designs. It has defined a number of steps to be taken by industry. At the same time, possibilities offered by national and international regulatory mechanisms would have to be fully made use of, and some changes in regulatory frameworks might be necessary. Some steps especially towards greater cooperation of regulators have already been taken; however, much still remains to be done. The concept of deploying standardized reactor designs across a number of countries supposes an alignment and, if possible, harmonization of national safety standards; a streamlining of national licensing procedures, making them more efficient and predictable; and the willingness of national regulators to take into account licensing done in other countries. In the end, this should lead to a mutual acceptance of design approvals or, in a more distant future, even to a multinational design approval process. All in all, the concept

Moderator behaviour and reactor internals integrity at Atucha I NPP

International Nuclear Information System (INIS)

Berra, S.; Guala, M.; Herzovich, P.; Chocron, M.; Lorenzo, A.; Raffo Calderon, Ma. C. del; Urrutia, G.

1996-01-01

Atucha I is a Pressure Vessel Heavy Water Cooled Heavy Water Moderator Reactor. In this kind of reactor the moderator tank is physically connected to the primary coolant. Since neutron economy requires the moderator to be as cold as possible, it is necessary that even when physically connected, it should have a separated cooling system, which in this case is also used as a feed-water preheater, and also heat mass transfer with primary coolant should be minimized. This condition requires that some reactor internals are designed in principle to last the whole life of the plant. However, in 1988 the failure of one internal produced a 16 month shut down. This incident could have been prevented but the idea that reactor internals would not have failures due to aging was dominant at that time avoiding the early detection of the failure. However, the analysis of the records after the incident showed that some process variables had changed previously to the incident, i.e., power exchanged at the moderator heat exchanger had increased. Since the station restart up some changes in the moderator process variables and a flow rate reduction of about 10% through the primary side of one moderator cooler were observed. In order to understand the flow reduction and the overall behaviour of moderators parameters, two models were developed that predict moderator and moderator cooler behavior under the new conditions. The present paper refers to these models, which together with the improvement of process variables measurements mentioned in another paper presented at this meeting permits to understand current moderator behaviour and helps to early diagnostic of an eventual reactor internal failure. (author). 2 refs, 4 figs, 1 tab

Moderator behaviour and reactor internals integrity at Atucha I NPP

Energy Technology Data Exchange (ETDEWEB)

Berra, S; Guala, M; Herzovich, P [Central Nuclear Atucha I, Nucleoelectrica Argentina, Lima, Buenos Aires (Argentina); Chocron, M; Lorenzo, A; Raffo Calderon, Ma. C. del; Urrutia, G [Comision Nacional de Energia Atomica, Buenos Aires (Argentina). Centro Atomico Constituyentes

1997-12-31

Atucha I is a Pressure Vessel Heavy Water Cooled Heavy Water Moderator Reactor. In this kind of reactor the moderator tank is physically connected to the primary coolant. Since neutron economy requires the moderator to be as cold as possible, it is necessary that even when physically connected, it should have a separated cooling system, which in this case is also used as a feed-water preheater, and also heat mass transfer with primary coolant should be minimized. This condition requires that some reactor internals are designed in principle to last the whole life of the plant. However, in 1988 the failure of one internal produced a 16 month shut down. This incident could have been prevented but the idea that reactor internals would not have failures due to aging was dominant at that time avoiding the early detection of the failure. However, the analysis of the records after the incident showed that some process variables had changed previously to the incident, i.e., power exchanged at the moderator heat exchanger had increased. Since the station restart up some changes in the moderator process variables and a flow rate reduction of about 10% through the primary side of one moderator cooler were observed. In order to understand the flow reduction and the overall behaviour of moderators parameters, two models were developed that predict moderator and moderator cooler behavior under the new conditions. The present paper refers to these models, which together with the improvement of process variables measurements mentioned in another paper presented at this meeting permits to understand current moderator behaviour and helps to early diagnostic of an eventual reactor internal failure. (author). 2 refs, 4 figs, 1 tab.

Application of Code Of Conduct on the Safety of Research Reactor (RTP)

International Nuclear Information System (INIS)

Ligam, A.S.; Ahmad Nabil Abd Rahim; Zarina Masood

2014-01-01

The implementation and the practices of the effective safety system at research reactors are important to ensure that the worker, public and environment do not receive any abnormal causes. Many international safety related support agencies for research reactor such as International Atomic Energy Agency (IAEA) providing guidelines that can be applied to enhance and strengthen the enforcement of safety namely Code of Conduct on the Safety of Research Reactor (IAEA/CODEOC/RR/2006). The excellent safety management, reliability, and maintainability of RTP reactor structures, coupled with personnel numerous lessons and experiences learned, Reactor TRIGA PUSPATI research reactor providing Nuclear Malaysia personnel and visitor the very safe working and visiting environment. This paper will discuss the status, practices and improvement strategies over the past few years. (author)

Internal conversion theory of gamma radiation in unfilled atomic shells

International Nuclear Information System (INIS)

Anderson, Eh.M.; Trusov, V.F.; Ehglajs, M.O.

1980-01-01

The internal conversion theory of gamma radiation in unfilled shells, when the atom is in a state with certain energy and momentum, is considered. A formula for the conversion coefficient between the atom and ion levels is obtained. This coefficient turns to be dependent on genealogic characteristics of the atom. It is discussed when the conversion coefficients are proportional to the numbers of filling subshells in the atom. Exact calculations have been carried out in the multiconfigurational approximation taking into account intermediate coupling for the d-shell of the Fe atom Single-electron radial wave functions have been calculated on the basis of the relativistic method of the Hartree-Fock-Dirak self-consistent field. Conversion coefficients on certain subshells as well as submatrix elements of the production operator are calculated. The electric coefficient of internal conversion (CIC) in the calculation for one electron does not depend on spin orientation. That is why the electric CIC from the level will not depend on filling number distribution by subshells. For magnetic CIC the dependence on the atom state is significant. Using multiconfiguration basis for calculating energy matrix and its succeeding diagonalization means the account of the intermediate coupling type, which takes place for the unfilled shells

Atomic layer deposition on nanoparticles in a rotary reactor

Science.gov (United States)

McCormick, Jarod Alan

Challenges are encountered during atomic layer deposition (ALD) on large quantities of nanoparticles. The particles must be agitated or vigorously mixed to perform the ALD surface reactions in reasonable times and to prevent the particles from being agglomerated by the ALD film. The high surface area of nanoparticles also demands efficient reactant usage because large quantities of reactant are required for the surface reactions to reach completion. To address these challenges, a novel rotary reactor was developed to achieve constant particle agitation during static ALD reactant exposures. In the design of this new reactor, a cylindrical drum with porous metal walls was positioned inside a vacuum chamber. The porous cylindrical drum was rotated by a magnetically coupled rotary feedthrough. By rotating the cylindrical drum to obtain a centrifugal force of less than one gravitational force, the particles were agitated by a continuous "avalanche" of particles. The effectiveness of this rotary reactor was demonstrated by Al 2O3 ALD on ZrO2 particles. A number of techniques including transmission electron microscopy, Fourier transform infrared spectroscopy, scanning Auger spectroscopy and x-ray photoelectron spectroscopy confirmed that the Al2O3 ALD film conformally coats the ZrO 2 particles. Combining static reactant exposures with a very high surface area sample in the rotary reactor also provides unique opportunities for studying the surface chemistry during ALD. Sequential, subsaturating doses can be used to examine the self-limiting behavior of the ALD reactions in the rotary reactor. This dosing method is the first demonstration of self-limiting ALD on bulk quantities of nanoparticles. By combining these sequential, subsaturating doses with quadrupole mass spectrometry, ALD reactions can be analyzed from the gas phase using full mass spectrum analysis. The reaction products are present in a high enough concentration to discern a gas phase mechanism for reactions

The Atomic energy basic law

International Nuclear Information System (INIS)

1979-01-01

The law aims to secure future energy resources, push forward progress of science and advancement of industry for welfare of the mankind and higher standard of national life by helping research, development and utilization of atomic power. Research, development and utilization of atomic power shall be limited to the peaceful purpose with emphasis laid on safety and carried on independently under democratic administration. Basic concepts and terms are defined, such as: atomic power; nuclear fuel material; nuclear raw material; reactor and radiation. The Atomic Energy Commission and the Atomic Energy Safety Commission shall be set up at the Prime Minister's Office deliberately to realize national policy of research, development and utilization of atomic power and manage democratic administration for atomic energy. The Atomic Energy Commission shall plan, consider and decide matters concerning research, development and utilization of atomic energy. The Atomic Energy Safety Commission shall plan, consider and decide issues particularly concerning safety securing among such matters. The Atomic Energy Research Institute shall be founded under the governmental supervision to perform research, experiment and other necessary affairs for development of atomic energy. The Power Reactor and Nuclear Fuel Development Corporation shall be established likewise to develop fast breeding reactor, advanced thermal reactor and nuclear fuel materials. Development of radioactive minerals, control of nuclear fuel materials and reactors and measures for patent and invention concerning atomic energy, etc. are stipulated respectively. (Okada, K.)

The influence of atomizer internal design and liquid physical properties on effervescent atomizing of coal-water slurry

Energy Technology Data Exchange (ETDEWEB)

Liu, Meng; Duan, Yufeng [Southeast Univ., Nanjing (China). Inst. of Thermal Engineering

2013-07-01

This study investigated the dependence of effervescent atomizing of coal-water slurry (CWS) on atomizer internal design and fluid properties. Results demonstrate that internal design of atomizer and fluid properties directly affect the two-phase flow pattern inside the atomizer which consequently affects the spray quality. The influence of mixing chamber length on spray quality is not significant at the ALR of 0.15 except for spray 0.75 glycerol/0.248 water/0.002 xanthan mixture. The same trend also found in the effect of angle of aeration holes at ALR of 0.15. Large diameter of the inclined aeration holes shows small SMD for water. The consistency index of fluids has no effect on the spray quality and Sauter Mean Diameter (SMD) increases when polymer additions were added to the glycerin-water mixture. The radial profile of SMD for spray water are almost flat, however, the largest SMD can be obtained at the edge of spray for three other fluids.

Annual Report to Congress of the Atomic Energy Commission for 1968

Energy Technology Data Exchange (ETDEWEB)

Seaborg, Glenn T.

1969-01-31

The document represents the 1968 Annual Report of the Atomic Energy Commission (AEC) to Congress. The report opens with ''An Introduction to the Atomic Energy Programs during 1968'' followed by 17 Chapters, 8 appendices and an index. Chapters are as follows: (1) Source, Special, and Nuclear Byproduct Materials; (2) Nuclear Materials Safeguards; (3) The Nuclear Defense Effort; (4) Naval Propulsion Reactors; (5) Reactor Development and Technology; (6) Licensing and Regulating the Atom; (7) Operational and Public Safety; (8) Nuclear Rocket Propulsion; (9) Specialized Nuclear Power; (10) Isotopic Radiation Applications; (11) Peaceful Nuclear Explosives; (12) International Affairs and Cooperation; (13) Informational and Related Activities; (14) Nuclear Education and Training; (15) Biomedical and Physical Research; (16) Industrial Participation Aspects; and, (17) Administrative and Management Matters.

Annual Report to Congress of the Atomic Energy Commission for 1969

Energy Technology Data Exchange (ETDEWEB)

Seaborg, Glenn T.

1970-01-31

The document represents the 1969 Annual Report of the Atomic Energy Commission (AEC) to Congress. The report opens with ''An Introduction to the Atomic Energy Programs during 1969'' followed by 17 Chapters, 8 appendices and an index. Chapters are as follows: (1) Source, Special, and Byproduct Nuclear Materials; (2) Nuclear Materials Safeguards; (3) The Nuclear Defense Effort; (4) Naval Propulsion Reactors; (5) Reactor Development and Technology; (6) Licensing and Regulating the Atom; (7) Operational and Public Safety; (8) Space Nuclear Propulsion; (9) Specialized Nuclear Power; (10) Isotopic Radiation Applications; (11) Peaceful Nuclear Explosives; (12) International Affairs and Cooperation; (13) Informational and Related Activities; (14) Nuclear Education and Training; (15) Biomedical and Physical Research; (16) Industrial Participation Aspects; and, (17) Administrative and Management Matters.

International Atomic Energy Agency activities in decommissioning

International Nuclear Information System (INIS)

Reisenweaver, D W.; )

2005-01-01

Full text: The International Atomic Energy Agency (IAEA) has been addressing the safety and technical issues of decommissioning for over 20 years, but their focus has been primarily on planning. Up to know, the activities have been on an ad hoc basis and sometimes, important issues have been missed. A new Action Plan on the Decommissioning of Nuclear Facilities has recently been approved by the Agency's board of Governors which will focus the Agency's efforts and ensure that our Member States' concerns are addressed. The new initiatives associated with this Action Plan will help ensure that decommissioning activities in the future are performed in a safe and coherent manner. The International Atomic Energy Agency (IAEA) has been preparing safety and technical documents concerning decommissioning since the mid-1980's. There have been over 30 documents prepared that provide safety requirements, guidance and supporting technical information. Many of these documents are over 10 years old and need updating. The main focus in the past has been on planning for decommissioning. During the past five years, a set of Safety Standards have been prepared and issued to provide safety requirements and guidance to Member States. However, decommissioning was never a real priority with the Agency, but was something that had to be addressed. To illustrate this point, the first requirements documents on decommissioning were issued as part of a Safety Requirements [1] on pre-disposal management of radioactive waste. It was felt that decommissioning did not deserve its own document because it was just part of the normal waste management process. The focus was mostly on waste management. The Agency has assisted Member States with the planning process for decommissioning. Most of these activities have been focused on nuclear power plants and research reactors. Now, support for the decommissioning of other types of facilities is being requested. The Agency is currently providing technical

Kazakhstan participation in International Experimental Reactor ITER Construction project. Work status and prospects

International Nuclear Information System (INIS)

Tazhibayeva, I.L.; Tukhvatullin, Sh.T.; Shestakov, V.P.; Kuznetsov, B.A.

2002-01-01

these works is also a part of further Kazakhstan contribution in ITER project. JSC 'UMP' provides feasibility study of manufacturing possibility of complete details from beryllium together with superconductor components taking into consideration ITER project requirements. It is supposed that new grades of beryllium, such as SS65C or DSHG-200, will be developed or alternative grade of beryllium with its further verification will be used. As a support for project ITER participation as well as for development of new contemporary trends in science and technology development, and for training the high skilled personnel a unique and the only one in the world KTM experimental tokamak installation is being developed in the country. It will be used for testing structural materials and important elements of thermonuclear reactors. At present a draft of scientific program for KTM reactor researches has been worked out; also feasibility study has been conducted, preliminary design is completed and the work of design documentation development and modeling of the installation is got under way. KTM tokamak, which is constructed in NNC RK in partnership with Russian organizations, will allow to perform predicative tests of future ITER reactor assemblies being produced at JSC 'UMP'. It will promote international cooperation in the field of thermonuclear material science and new material production. It will also contribute greatly in new technologies and modern designs of thermonuclear fusion reactors debugging. ITER and KTM projects are the examples of Successful Eurasia cooperation in the field of peaceful atomic energy use

The ReactorSTM: Atomically resolved scanning tunneling microscopy under high-pressure, high-temperature catalytic reaction conditions

Energy Technology Data Exchange (ETDEWEB)

Herbschleb, C. T.; Tuijn, P. C. van der; Roobol, S. B.; Navarro, V.; Bakker, J. W.; Liu, Q.; Stoltz, D.; Cañas-Ventura, M. E.; Verdoes, G.; Spronsen, M. A. van; Bergman, M.; Crama, L.; Taminiau, I.; Frenken, J. W. M., E-mail: frenken@physics.leidenuniv.nl [Huygens-Kamerlingh Onnes Laboratory, Leiden University, P.O. box 9504, 2300 RA Leiden (Netherlands); Ofitserov, A.; Baarle, G. J. C. van [Leiden Probe Microscopy B.V., J.H. Oortweg 21, 2333 CH Leiden (Netherlands)

2014-08-15

To enable atomic-scale observations of model catalysts under conditions approaching those used by the chemical industry, we have developed a second generation, high-pressure, high-temperature scanning tunneling microscope (STM): the ReactorSTM. It consists of a compact STM scanner, of which the tip extends into a 0.5 ml reactor flow-cell, that is housed in a ultra-high vacuum (UHV) system. The STM can be operated from UHV to 6 bars and from room temperature up to 600 K. A gas mixing and analysis system optimized for fast response times allows us to directly correlate the surface structure observed by STM with reactivity measurements from a mass spectrometer. The in situ STM experiments can be combined with ex situ UHV sample preparation and analysis techniques, including ion bombardment, thin film deposition, low-energy electron diffraction and x-ray photoelectron spectroscopy. The performance of the instrument is demonstrated by atomically resolved images of Au(111) and atom-row resolution on Pt(110), both under high-pressure and high-temperature conditions.

International tokamak reactor conceptual design overview

International Nuclear Information System (INIS)

Stacey, W.M. Jr.

1983-01-01

The International Tokamak Reactor (INTOR) Workshop is an unique collaborative effort among Euratom, Japan, the USA and the USSR, under the auspices of the IAEA, to assess, define, design, construct and operate the next major experiment in the World Tokamak Program beyond the TFTR, JET, JT-60, T-15 generation. During the Zero-Phase (1979), a technical data base assessment was performed, leading to a positive assessment of feasibility. During Phase-I (1/80-6/81), a conceptual design was developed to define the concept. The programmatic objectives are that INTOR should: (1) be the maximum reasonable step beyond the TFTR, JET, JT-60, T-15 generation of tokamaks, (2) demonstrate the plasma performance required for tokamak DEMOs, (3) test the development and integration into a reactor system of those technologies required for a DEMO, (4) serve as a test facility for blanket, tritium production, materials, and plasma engineering technology, (5) test fusion reactor component reliability, (6) test the maintainability of a fusion reactor, and (7) test the factors affecting the reliability, safety and environmental acceptability of a fusion reactor. A conceptual design has been developed to define a device which is consistent with these objectives. The design concept could, with a reasonable degree of confidence, be developed into a workable engineering design of a tokamak that met the performance objectives of INTOR. There is some margin in the design to allow for uncertainty. While design solutions have been found for all of the critical issues, the overall design may not yet be optimal. (author)

Atomic nanoscale technology in the nuclear industry

CERN Document Server

Woo, Taeho

2011-01-01

Developments at the nanoscale are leading to new possibilities and challenges for nuclear applications in areas ranging from medicine to international commerce to atomic power production/waste treatment. Progress in nanotech is helping the nuclear industry slash the cost of energy production. It also continues to improve application reliability and safety measures, which remain a critical concern, especially since the reactor disasters in Japan. Exploring the new wide-ranging landscape of nuclear function, Atomic Nanoscale Technology in the Nuclear Industry details the breakthroughs in nanosca

Reduced enrichment for research and test reactors: Proceedings

International Nuclear Information System (INIS)

1993-07-01

The 15th annual Reduced Enrichment for Research and Test Reactors (RERTR) international meeting was organized by Ris oe National Laboratory in cooperation with the International Atomic Energy Agency and Argonne National Laboratory. The topics of the meeting were the following: National Programs, Fuel Fabrication, Licensing Aspects, States of Conversion, Fuel Testing, and Fuel Cycle. Individual papers have been cataloged separately

Reduced enrichment for research and test reactors: Proceedings

Energy Technology Data Exchange (ETDEWEB)

1993-07-01

The 15th annual Reduced Enrichment for Research and Test Reactors (RERTR) international meeting was organized by Ris{o} National Laboratory in cooperation with the International Atomic Energy Agency and Argonne National Laboratory. The topics of the meeting were the following: National Programs, Fuel Fabrication, Licensing Aspects, States of Conversion, Fuel Testing, and Fuel Cycle. Individual papers have been cataloged separately.

Proceedings of 18th international conference on structural mechanics in reactor technology

International Nuclear Information System (INIS)

2005-07-01

The 18th International Conference on Structural Mechanics in Reactor Technology was held on August 7-12, 2005 in Beijing, China, and Sponsored by International Association for Structural Mechanics in Reactor Technology, Chinese Nuclear Society, Chinese Society of Theoretical and Applied Mechanics, and Tsinghua University. 486 abstracts are Collected. The contents includes: opening, plenary and keynote presentations; computational mechanics; fuel and core structures; aging, life extension, and license renewal; design methods and rules for components; fracture mechanics; concrete material, containment and other structures; analysis and design for dynamic and extreme loads; seismic analysis, design and qualification; structural reliability and probabilistic safety assessment (PSA); operation, inspection and maintenance; severe accident management and structural evaluation; advanced reactors and generation IV reactors; decommissioning of nuclear facilities and waste management.

FFTF scale-model characterization of flow-induced vibrational response of reactor internals

International Nuclear Information System (INIS)

Ryan, J.A.; Julyk, L.J.

1977-01-01

As an integral part of the Fast Test Reactor Vibration Program for Reactor Internals, the flow-induced vibrational characteristics of scaled Fast Test Reactor core internal and peripheral components were assessed under scaled and simulated prototype flow conditions in the Hydraulic Core Mockup. The Hydraulic Core Mockup, a 0.285 geometric scale model, was designed to model the vibrational and hydraulic characteristics of the Fast Test Reactor. Model component vibrational characteristics were measured and determined over a range of 36 percent to 111 percent of the scaled prototype design flow. Selected model and prototype components were shaker tested to establish modal characteristics. The dynamic response of the Hydraulic Core Mockup components exhibited no anomalous flow-rate dependent or modal characteristics, and prototype response predictions were adjudged acceptable

FFTF scale-model characterization of flow induced vibrational response of reactor internals

Energy Technology Data Exchange (ETDEWEB)

Ryan, J A; Julyk, L J [Hanford Engineering Development Laboratory, Richland, WA (United States)

1977-12-01

As an integral part of the Fast Test Reactor Vibration Program for Reactor Internals, the flow-induced vibrational characteristics of scaled Fast Test Reactor core internal and peripheral components were assessed under scaled and simulated prototype flow conditions in the Hydraulic Core Mockup. The Hydraulic Core Mockup, a 0.285 geometric scale model, was designed to model the vibrational and hydraulic characteristics of the Fast Test Reactor. Model component vibrational characteristics were measured and determined over a range of 36% to 111% of the scaled prototype design flow. Selected model and prototype components were shaker tested to establish modal characteristics. The dynamic response of the Hydraulic Core Mockup components exhibited no anomalous flow-rate dependent or modal characteristics, and prototype response predictions were adjudged acceptable. (author)

FFTF scale-model characterization of flow induced vibrational response of reactor internals

International Nuclear Information System (INIS)

Ryan, J.A.; Julyk, L.J.

1977-01-01

As an integral part of the Fast Test Reactor Vibration Program for Reactor Internals, the flow-induced vibrational characteristics of scaled Fast Test Reactor core internal and peripheral components were assessed under scaled and simulated prototype flow conditions in the Hydraulic Core Mockup. The Hydraulic Core Mockup, a 0.285 geometric scale model, was designed to model the vibrational and hydraulic characteristics of the Fast Test Reactor. Model component vibrational characteristics were measured and determined over a range of 36% to 111% of the scaled prototype design flow. Selected model and prototype components were shaker tested to establish modal characteristics. The dynamic response of the Hydraulic Core Mockup components exhibited no anomalous flow-rate dependent or modal characteristics, and prototype response predictions were adjudged acceptable. (author)

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

Maintenance experience on reactor recirculation pumps at Tarapur Atomic Power Station

International Nuclear Information System (INIS)

Singh, A.K.

1995-01-01

Reactor recirculation pumps at Tarapur Atomic Power Station (TAPS) are vertical, single stage centrifugal pumps having mechanical shaft seals and are driven by vertical mounted 3.3 kV, 3 phase, 1500 h.p. electric motors. During these years of operation TAPS has gained enough experience and expertise on the maintenance of reactor recirculation pumps which are dealt in this article. Failure of mechanical shaft seals, damage on pump carbon bearings, motor winding insulation failures and motor shaft damage have been the main areas of concern on recirculation pump. A detailed procedure step by step with component sketches has helped in eliminating errors during shaft seal assembly and installation. Pressure breakdown devices in seal assembly were rebuilt. Additional coolant water injection for shaft seal cooling was provided. These measures have helped in extending the reactor recirculation pump seal life. Pump bearing problems were mainly due to failure of anti-rotation pins and dowel pins of bearing assembly. These pins were redesigned and strengthened. Motor stator winding insulation failures were detected. Stator winding replacement program has been taken up on regular basis to avoid winding insulation failure due to aging. 3 refs., 2 tabs., 7 figs

Procedures of ASME code case N-201 for KALIMER. Reactor internal structures

International Nuclear Information System (INIS)

Koo, Gyeong Hoi; Yoo, B.

2001-02-01

The main objective of this report is to describe the design procedure of ASME Boiler and Pressure Vessel Code, Code Case N-201-4, which is an elevated temperature structural design code of the Nuclear reactor internal structures, checking the criteria of stress limit, accumulated inelastic strain and deformation, creep-fatigue damage, and buckling limit. As one of examples, the creep-fatigue damage evaluations are carried out for the KALIMER reactor internal structures of baffle annulus. This report is expected to be very useful in evaluating the structural integrity of the liquid metal reactor operating under an elevated temperature

TIBER (Tokamak Ignition/Burn Experimental Reactor) II as a precursor to an international thermonuclear experimental reactor

International Nuclear Information System (INIS)

Henning, C.D.; Gilleland, J.R.

1988-01-01

The Tokamak Ignition/Burn Experimental Reactor (TIBER) was pursued in the US as one option for an International Thermonuclear Experimental Reactor (ITER). This concept evolved from earlier work on the Tokamak Fusion Core Experiment (TFCX) to develop a small, ignited tokamak. While the copper-coil versions of TFCX became the short-pulsed, 1.23-m radius, Compact Ignition Tokamak (CIT), the superconducting TIBER with long pulse or steady state and a 2.6-m radius was considered for international collaboration. Recently the design was updated to TIBER II, to accommodate more conservative confinement scaling, double-poloidal divertors for impurity control, steady-state current drive, and nuclear testing. 18 refs., 1 fig

Annual report of the Department of Atomic Energy 1977-78

International Nuclear Information System (INIS)

1978-01-01

The activities during the financial year 1977-78 of the research organizations and laboratories, various projects underway and public sector undertakings of the Department of Atomic Energy (India) have been reported. The R and D Work of the Bhabha Atomic Research Centre, Bombay and Reactor Research Centre, Kalpakkam in the fields of nuclear physics, radio- and radiation chemistry and other physical sciences; biological sciences including nuclear medicine, food irradiation and plant breeding by radiation mutation; reactor engineering and application of radiation and radioisotopes has been surveyed. The progress of heavy water projects, MHD project, nuclear power plant projects and 100 Mw thermal research reactor R-5 project has been described. Performance of Tarapur and Rajasthan Atomic Power Stations, Nuclear Fuel Complex and Electronics Corporation of India Ltd., both at Hyderabad, Uranium Corporation of India Ltd. at Jaduguda and Indian Rare Earths Ltd. has been reported. Major achievements during the period of report are : (1) completion of construction work of the Power Reactor Fuel Reprocessing Plant at Tarapur and (2) Commissioning of the Variable Energy Cyclotron, Calcutta for the internal circulating beam of alpha particles. (M.G.B.)

Ratcheting problems for ITER [International Thermonuclear Experimental Reactor

International Nuclear Information System (INIS)

Majumdar, S.

1991-01-01

Because of the presence of high cyclic thermal stress, pressure-induced primary stress, and disruption-induced high cyclic primary stress, ratcheting of the first wall poses a serious challenge to the designers of ITER (International Thermonuclear Experimental Reactor). Existing design tools such as the Bree diagram in the ASME Boiler and Pressure Vessels Code, are not directly applicable to ITER, because of important differences in geometry and loading modes. Available alternative models for ratcheting are discussed and new Bree diagrams, that are more relevant for fusion reactor applications, are proposed. 9 refs., 17 figs

International Conference on Physics and Technology of Reactors and Applications

International Nuclear Information System (INIS)

2011-01-01

Full text : The international conference on physics and technology of reactors is organized by the Moroccan Association for Nuclear enggineering and Reactor Technology (GMTR) with the collaboration of the Centre for Energy and Nuclear Sciences and Techniques (CNESTEN) and under the auspices of the ministry of Energy, Mining, Water and Environment. The programme of the PHYTRA2 conference covers a wide variety of topics. The conference is organised in one plenary session, eight oral technical sessions and one poster session. The oral and poster technical sessions covers the usual topics of nuclear engineering including one session on research reactors utilisation and computational methods for research reactors

Proceedings of the nineteenth symposium of atomic energy research on WWER reactor physics and reactor safety

International Nuclear Information System (INIS)

Vidovszky, I.

2009-10-01

The present volume contains 55 papers, presented on the nineteenth symposium of atomic energy research, held in Varna, Bulgaria, 21-25 September 2009. The papers are presented in their original form, i. e. no corrections or modifications were carried out. The content of this volume is divided into thematic groups: Fuel Management, Spectral and Core Calculations, Core Surveillance and Monitoring, CFD Analysis, Reactor Dynamics Thermal Hydraulics and Safety Analysis, Physical Problems of Spent Fuel Decommissioning and Radwaste, Actinide Transmutation and Spent Fuel Disposal, Core Operation, Experiments and Code Validation - according to the presentation sequence on the Symposium. (Author)

Water cooled reactor technology: Safety research abstracts no. 1

International Nuclear Information System (INIS)

1990-01-01

The Commission of the European Communities, the International Atomic Energy Agency and the Nuclear Energy Agency of the OECD publish these Nuclear Safety Research Abstracts within the framework of their efforts to enhance the safety of nuclear power plants and to promote the exchange of research information. The abstracts are of nuclear safety related research projects for: pressurized light water cooled and moderated reactors (PWRs); boiling light water cooled and moderated reactors (BWRs); light water cooled and graphite moderated reactors (LWGRs); pressurized heavy water cooled and moderated reactors (PHWRs); gas cooled graphite moderated reactors (GCRs). Abstracts of nuclear safety research projects for fast breeder reactors are published independently by the Nuclear Energy Agency of the OECD and are not included in this joint publication. The intention of the collaborating international organizations is to publish such a document biannually. Work has been undertaken to develop a common computerized system with on-line access to the stored information

About connection between atomic and hydrogen energy power

International Nuclear Information System (INIS)

Avdeeva, M.Zh.; Vecher, A.A.; Pan'kov, V.V.

2008-01-01

Possible interaction between atomic and hydrogen energy power has been discussed. The analysis of the result held shows that the electrical energy produced by the atomic reactor during the of-load hours can be involved into the process of obtaining hydrogen by electrolysis. In order to optimize the transportation and storage of hydrogen it is proposed to convert it into ammonia. The direct uses of ammonia as a fuel into the internal combustion engine and fuel cells are examined. (authors)

Process and apparatus for adjusting a new upper reactor internals in a reactor vessel of a PWR

International Nuclear Information System (INIS)

Frizot, A.; Cadaureille, G.; Lalere, C.; Machuron, J.Y.

1987-01-01

On the new upper reactor internals is mounted devices for alignment and clearances, before introducing in the reactor vessel. After introducing alignment and clearances are measured. Adjustment pieces are provided for optimum clearances and alignment and fixed after removal from vessel. Decontamination is made by using water jets prior to fitting recess parts [fr

Fast nuclear reactors. Associated international projects. State of the art and assessment of the concepts

International Nuclear Information System (INIS)

Azpitarte, O.; Ramilo, L.

2013-01-01

The recognition of the strategic importance of nuclear energy as a source of sustainable energy may be perceived in the continuous development, in many countries, of the technology of fast nuclear reactors with an associated closed fuel cycle, assuming that these Generation IV innovative systems will be required in the future. These reactors fulfill international requirements for safety and reliability, economic competitiveness, sustainability and proliferation resistance. They have the potential of using more efficiently the natural resources of Uranium and of reducing the volume and radiotoxicity of the nuclear waste by partitioning and transmutation of Minor Actinides. The national and international programs being carried out today are concentrated in the following concepts: Sodium Fast Reactor (SFR), Lead Fast Reactor (LFR), Gas Fast Reactor (GFR), Super Critical Water Reactor (SCWR) and Molten Salt Reactor (MSR). This article presents a short review of the technology of the mentioned concepts and details the current state of the main national and international related projects. (author)

Dismantling of JPDR reactor internals by underwater plasma arc cutting technique using robotic manipulator

International Nuclear Information System (INIS)

Yokota, M.

1988-01-01

The actual dismantling of JPDR started on December 4, 1986. As of now, equipment that surrounds the reactor has mostly been removed to provide working space in reactor containment prior to the dismantling of reactor internals. Some reactor internals have been successfully dismantled using the underwater arc cutting system with a robotic manipulator during the period of January to March 1988. The cutting system is composed of an underwater plasma arc cutting device and a robotic manipulator. The cut off reactor internals were core spray block, feedwater sparger and stabilizers for fuel upper grid tube. The plasma arc cutting device was developed to dismantle the reactor internals underwater. It mainly consists of a plasma torch, power and gas supply systems for the torch, and by-product treatment systems. It has the cutting ability of 130 mm thickness stainless steel underwater. The robotic manipulator has seven degrees of freedom of movement, enabling it to move in almost the same way as the arm of a human being. The arm of the robot is mounted on a supporting device which is suspended by three chains from the support structure set on a service floor. A plasma torch is griped by the robotic hand; its position to the structure to be cut is controlled from a remote control room, about 100 meters outside the reactor containment

Review of the general atomic experimental fusion power reactor initial conceptual design

International Nuclear Information System (INIS)

Baker, C.C.; Sager, P.H. Jr.; Harder, C.R.

1976-01-01

The primary objective of the Experimental Power Reactor (EPR) is to provide the necessary interface between physics experiments and the first demonstration power plants. Since economically viable tokamak-type reactors may well have to be very high Q devices (ratio of fusion power out to power into the plasma), it will be essential for a tokamak demonstration reactor to operate at or near ignition conditions. Thus, it is believed that one of the primary objectives of the EPR must be to fully model the behavior of a D-T burning plasma required in the reactor of a demonstration plant. Therefore, a major objective of the EPR should be to achieve ignition conditions. In addition to demonstrating the ability to ignite and control a D-T plasma, it is also desirable that the EPR should produce, or at least demonstrate the ability to produce, a small amount of net electrical power. These objectives should be accomplished at a reasonable cost; this implies achieving a sufficiently high β (ratio of plasma pressure to magnetic field pressure). It is believed that noncircular cross section tokamaks offer the best chance of realizing these objectives. Consequently, noncircular cross sections are a major design feature of the General Atomic EPR

International Nuclear Information System. 1988-2002. International Atomic Energy Agency publications

International Nuclear Information System (INIS)

2002-08-01

This catalogue lists all sales publications and products of the International Atomic Energy Agency dealing with the International Nuclear Information System (INIS), and issued during the period 1 January 1990 - 31 July 2002. Most publications are issued in English, though some are also available in other languages. This is noted as E for English, F for French, G for German, R for Russian and S for Spanish before the relevant ISBN number. Some INIS Reference Series publications are available in electronic form from the INIS Clearinghouse. For more details on the INIS publications programme, please visit the INIS web site mentioned above

Annual report of the Japan Atomic Energy Research Institute, for fiscal 1989

International Nuclear Information System (INIS)

1990-01-01

Japan Atomic Energy Research Institute has promoted the research on nuclear safety, the research and development of high temperature engineering and nuclear fusion which are the leading projects bringing about the breakthrough in atomic energy technology, the research on radiation utilization and the research and development of nuclear-powered ships, following the 'Plan of development and long term utilization of atomic energy' decided in 1987, as the central, general research institute in atomic energy field in Japan. Also the advanced basic research for opening atomic energy frontier and various international cooperation as well as the cooperation in Japan have been promoted. The engineering safety of nuclear facilities and environmental safety, the construction of the Nuclear Fuel Cycle Safety Engineering Research Facility, the design of the High Temperature Engineering Test Reactor and the various tests related to it, the reconstruction of JT-60 for increasing the current, the design of a nuclear fusion reactor, the high utilization of radiation using ion beam, the construction of Sekinehama Port for the nuclear-powered ship 'Mutsu', the power increasing test of the reactor of the Mutsu, the reconstruction of JRR-3 and others are reported. (K.I.)

Choosing a standard reactor: International competition and domestic politics in Chinese nuclear policy

International Nuclear Information System (INIS)

Ramana, M.V.; Saikawa, Eri

2011-01-01

China has ambitious plans to expand its nuclear power capacity. One of the policy goals that high-level policymakers have desired is to base the nuclear program on a standardized reactor design. However, this has not materialized so far. By examining its nuclear reactor choices for individual projects, we argue that China’s policymaking process has been greatly influenced by international competition and domestic politics. Multiple international nuclear vendors are intent upon maintaining their respective niches in the expanding Chinese reactor market, and they have used various forms of economic and political pressure to achieve their objectives. On the other hand, China’s policymaking process is fragmented and the shifting power balances among powerful domestic actors do not allow a fixed path to be followed. Further, because of the high costs and potential profits involved, nuclear reactor choices in China have been driven not just by technical considerations but also by foreign and trade policy objectives. All of these make it unlikely that China will standardize the reactor type it constructs in the near future. -- Highlights: ► China’s nuclear power policymaking has been fragmented and without central control. ► Multiple domestic actors have pursued independent agendas. ► International nuclear vendors have intensely competed for Chinese reactor contracts. ► Economic, political and foreign policy goals have driven reactor contract decisions. ► China is unlikely to construct only a standardized reactor design.

Decommissioning: Strategies and programmes at the International Atomic Energy Agency (IAEA)

International Nuclear Information System (INIS)

Laraia, M.

2003-01-01

The International Atomic Energy Agency (IAEA) has included decommissioning in its regular programmes since 1985. Until a few years ago, attention was focused on the decommissioning of nuclear power plants, and to a lesser extent, research reactors. Some countries, however, are now devoting greater attention to the decommissioning of non-reactor facilities, with implementation of these programmes being seen as a high priority. This demanded equal attention in IAEA's programmes. In recent years, the IAEA has expanded its programmes to include guidance on decommissioning of small medical, industrial and research facilities which are prevailing in most of its over 130 Member States. By 2010-2020, a significant number of nuclear power plants, research reactors, fuel cycle and non-reactor facilities will have exceeded their normal design lifetimes. Many of these facilities are already shutdown and are awaiting decommissioning. In 1996, the IAEA organized the decommissioning programme along two directions. A first direction focuses on the safety of management of radioactive waste including decommissioning. A second direction focuses on the technology and strategies to support waste management and decommissioning activities. This split of activities was instituted in order to keep the regulatory aspects separated from the strategic and technology-related activities. The focus of this paper will be on current and foreseen activities related to strategies and technologies of decommissioning, but other activities will be touched upon as well. All technical divisions of the IAEA provide technical support for Technical Co-operation (TC) projects with developing countries. TC projects in the field of decommissioning are given separate coverage in this paper. The IAEA documents on decommissioning strategies and technologies are presented in Section 2. Technical Co-operation Programme concerning Decommissioning is discussed focusing the objectives, the specific projects and the

Atomic and plasma-material interaction data for fusion. Vol.1

International Nuclear Information System (INIS)

1991-01-01

The International Atomic Energy Agency, through its Atomic and Molecular Data Unit, coordinates a wide spectrum of programmes for the compilation, evaluation, and generation of atomic, molecular, and plasma-wall interaction data for fusion research. The present, first, volume of Atomic and Plasma-Material Interaction Data for Fusion, contains extended versions of the reviews presented at the IAEA Advisory Group Meeting on Particle-Surface Interaction Data for Fusion, held 19-21 April 1989 at the IAEA Headquarters in Vienna, The plasma-wall interaction processes covered here are those considered most important for the operational performance of magnetic confinement fusion reactors. In addition to processes due to particle impact under normal operation, plasma-wall interaction effects due to off-normal plasma events (disruptions, electron runaway bombardment) are covered, and a summary of the status of data information on these processes is given from the point of view of magnetic fusion reactor design. Refs, figs and tabs

Status of national programmes on fast reactors 1997/98. 31. annual meeting of the International Working Group on Fast Reactors

International Nuclear Information System (INIS)

1998-01-01

The objective of the meeting was to co-ordinate the exchange of information on the status of fast reactor development and operational experience, including experience with experimental types of reactor; to consider meeting arrangements for 1998 and 1999; and to review the IAEA co-ordinated research activities in the field of fast reactor, as well as co-ordination of the International Working Group on Fast Reactors activities with other organizations

Report on application results of the nuclear reactor in Atomic Energy Research Laboratory, Rikkyo University. April 1994 - March 1995

International Nuclear Information System (INIS)

1995-01-01

This report is on researching action state, application state, management state, and others of 1994 fiscal year at the Atomic Energy Research Laboratory, Rikkyo University. The experimental reactor has been used for the studies such as application of neutron radioactivity analysis to multi fields, application of fission and alpha track method to age determination and metallurgy, hot atom chemistry, neutron radiation effect on semiconductors and others, nuclear data measurement, organism, materials and products using neutron radiography, and development and application to inspection of radiation detectors such as neutron detector. This report was a report shown as a shape of research results of actions of the researchers. And, another report of colaborate research results using the Rikkyo University reactor was also published from the Atomic Energy Center, the University of Tokyo begun since April, 1974. (G.K.)

Review of the current status of linear hybrid reactor concepts

International Nuclear Information System (INIS)

Schultz, K.R.

1977-07-01

A review was made of the current status of linear fusion-fission hybrid reactor design studies in the USA. The linear hybrid reactor concepts reviewed include the linear theta-pinch hybrid reactor being studied at Los Alamos Scientific Laboratory, the electron beam-heated solenoid hybrid reactor under development at Physics International Co., the laser-heated solenoid hybrid reactor being investigated at Mathematical Sciences Northwest, Inc., and the linear fusion waste burning reactor being studied at General Atomic Company. The discussion addresses confinement and heating mechanisms for each concept, as well as the hybrid blanket designs. The current state of the four reactor designs is summarized and the performance of the various concepts compared

Study of the strength of the internal can for internally and externally cooled fuel elements intended for gas graphite reactors

International Nuclear Information System (INIS)

Boudouresque, B.; Courcon, P.; Lestiboubois, G.

1964-01-01

The cartridge of an internally and externally cooled annular fuel element used in gas-graphite reactors is made up of an uranium fuel tube, an external can and an internal can made of magnesium alloy. For the thermal exchange between the internal can and the fuel to be satisfactory, it is necessary for the can to stay in contact with the uranium under all temperature conditions. This report, based on a theoretical study, shows how the internal can fuel gap varies during the processes of canning, charging into the reactor and thermal cycling. The following parameters are considered: tube diameter, pressure of the heat carrying gas, gas entry temperature, plasticity of the can alloy. It is shown that for all operating conditions the internal can of a 77 x 95 element, planned for a gas-graphite reactor with a 40 kg/cm 2 gas pressure, should remain in contact with the fuel. (authors) [fr

The gas turbine modular helium reactor. An international project to develop a safe, efficient, flexible product

International Nuclear Information System (INIS)

Silberstein, A.J.

1998-01-01

As originally scheduled, the Conceptual Design Report of the 600 Mwt Gas Turbine Modular Helium Reactor has been issued in October 1997 by OKBM in Nizhny Novgorod, a keystone Russian Engineering Institute fully involved in the realization of this International Project. The plutonium burning, graphite moderated helium cooled reactor design results from the work done on the basis of General Atomics original concept combined with the goal of optimizing safety power and efficiency with multi contributions in specific fields from the Russian organizations: MINATOM, OKBM, VNIINM, Lutch, Kurchatov Institute, Seversk Chemical Combinat, Fuji Electric and FRAMATOME. The objective to concentrate the engineering work in Russia has met a full success due principally to the quality and experience of the people, to the international support and to the progressive integration of new techniques of communication, of project management culture and utilization of modern computerized design tools and methods. To day the best international standard of quality is reached in the engineering activity and expected to stay at this level for future developments, when including experimental facilities operation and components manufacturing activities, thanks to the diffusion of the common culture, acquired by the main actors during the conceptual design phase, that will be exported to Russian third parties. At this stage we are planning to start design verification and sensitive components and systems qualification, with the same original actors. The European Commission has already shown some significant interest through the MICHELANGELO Initiative in supporting the HTR concepts assessment and identification of the R and D needs. We are looking forward for further support from the International Community and particularly from European Institutions in the frame of the 5th PCRD to pursue the GT MHR R and D program. Furthermore we are looking for funding the building of a prototype in Russia

Agreement among the Government of the Republic of Poland, the Government of the United States of America and the International Atomic Energy Agency for assistance in securing nuclear fuel for a research reactor

International Nuclear Information System (INIS)

2007-01-01

The text of the Project and Supply Agreement among the Government of the Republic of Poland, the Government of the United States of America and the International Atomic Energy Agency for Assistance in Securing Nuclear Fuel for a Research Reactor is reproduced in this document for the information of all Members of the Agency. The Agency's Board of Governors approved the above mentioned Project and Supply Agreement on 14 June 2006. The Agreement was signed by the authorized representatives of Poland on 8 January 2007, the United States on 12 January 2007 and by the Director General of the IAEA on 16 January 2007. Pursuant to the Article XII of the Agreement, the Agreement entered into force on 16 January 2007, upon signature by the representatives of Poland, the United States and the Director General of the IAEA

White paper on atomic energy in 1979

International Nuclear Information System (INIS)

1980-01-01

In Japan, there are currently 21 nuclear power plants in operation with a total capacity of 15,000MW. Under the present situation of the so-called second energy crisis, the role of nuclear power is assuming increasingly more importance. The white paper is presented covering the one year period from October 1978; statistics, however, are for fiscal 1978. Contents are the following: part I general ''world nuclear power situation, advances in nuclear energy, the outlook for 1980s''; part II the status of nuclear power ''nuclear power generation, nuclear power safety, nuclear fuel cycle, international activities, safeguards, development of power reactors, nuclear fusion/nuclear powered ship/high-temperature gas cooled reactor, radiation utilization, basic research, nuclear power industry''; part III references (organization/plans of Atomic Energy Commission etc., atomic energy budgets, nuclear energy statistics, etc.). (J.P.N.)

Nuclear measurements, techniques and instrumentation, industrial applications, plasma physics and nuclear fusion 1986-1996. International Atomic Energy Agency publications

International Nuclear Information System (INIS)

1997-03-01

This catalogue lists all sales publications of the International Atomic Energy Agency dealing with Nuclear Measurements, Techniques, and Instrumentation, Industrial Applications, Plasma Physics and Nuclear Fusion, issued during the period 1986-1996. Most publications are in English. Proceedings of conferences, symposia and panels of experts may contain some papers in languages other than English (French, Russian or Spanish), but all of these papers have abstracts in English. Contents cover the three main areas of (i) Nuclear Measurements, Techniques and Instrumentation (Physics, Dosimetry Techniques, Nuclear Analytical Techniques, Research Reactor and Particle Accelerator Applications, and Nuclear Data), (ii) Industrial Applications (Radiation Processing, Radiometry, and Tracers), and (iii) Plasma Physics and Controlled Thermonuclear Fusion

International bulletin on atomic and molecular data for fusion. No. 53

International Nuclear Information System (INIS)

Stephens, J.A.

1997-11-01

The International Bulletin on Atomic and Molecular Data for Fusion is presented in four parts: 1) The Atomic and Molecular Data Information System (AMDIS) of the IAEA; 2) the indexed papers listed separately for structure and spectra, atomic and molecular collisions, and surface interactions; 3) all bibliographic data for both the indexed and non-indexed references; 4) the Author Index refers to the bibliographic references contained in Part 3

Agreement of 23 February 1989 Between the Government of the Democratic People's Republic of Algeria and the International Atomic Energy Agency for the Application of Safeguards in Connection With the Supply of a Research Reactor From the Republic of Argentina. Definitive Entry into Force

International Nuclear Information System (INIS)

1990-07-01

Pursuant to Section 28 of the Agreement between the Government of the Democratic People's Republic of Algeria and the International Atomic Energy Agency for the Application of Safeguards in Connection with the Supply of a Research Reactor from the Republic of Argentina, which entered into force provisionally on 23 February 1989, the Agreement entered into force definitively on 9 April 1990, the date on which the Agency received from the Government of Algeria written notification that the statutory and constitutional requirements of the Government of Algeria for the entry into force had been met.

Small intrinsically safe reactor implications

International Nuclear Information System (INIS)

Wakabayashi, Hiroaki

1985-01-01

Reviewing the history of nuclear power, it is found that peaceful uses of nuclear power are children of the war-like atom. Importance of special growth in a shielded environment is emphasized to exploit fully the advantages of nuclear power. Nuclear power reactors must be safe for their assimilation into society from the points of view of both technology and social psychology. ISR/ISER is identified as a missing link in the development of nuclear power reactors from this perspective and advocated for international development and utilization, being unleashed from the concerns of politicization, safety, and proliferation

Environment. 1990-2001. International Atomic Energy Agency publications

International Nuclear Information System (INIS)

2001-05-01

This catalog lists all sales publications of the International Atomic Energy Agency dealing with the Environment, and issued during the period 1 January 1990 - 30 April 2001. Most publications are issued in English, though some are also available in other languages. These are noted in the catalogue

Design and analysis of reactor headers for Narora Atomic Power Project

International Nuclear Information System (INIS)

Danak, M.R.

1975-01-01

Reactor header for Narora Atomic Power Reactor is a 400 mm O.D. 10 metres long pressure vessel in the primary coolant circuit connecting 153 feeders to PHT pumps or steam generators. The vessel dimensions are restricted are by containment philosophy. The outlet connections for pumps or steam generators are to be of the size of vessel diameter and DO/t ratio for the vessel is approximately 10. The design and stresses induced meet the code requirements except that at times it is difficult to get precise stress values in absence of certain data and lack of code or available literature giving practical approach to the problem. It can be seen that the 400 mm equal tees used as part of the vessel cannot be penetrated in the light of code reinforcement requirements. However if the tees have to penetrated to retain established feeder layout, it should be established experimentally or by some detailed stress analysis that it will meet the intent of code. (author)

The IAEA's WorldAtom Internet site: International news and information services

International Nuclear Information System (INIS)

Kyd, D.R.

2000-01-01

The International Atomic Energy Agency (IAEA) provides news and public information services via the Internet through its WorldAtom home page. The page is accessible at www.iaea.org/worldatom. Following are brief highlights of the items available on the site by clicking Press Centre, Reference Centre, or other links: Daily Press Review: Summaries of selected news items pertaining to global nuclear developments and the IAEA's work are provided each day, drawing upon a wide range of global media sources. IAEA NewsBriefs: Regularly featured are updates about IAEA activities related to areas of safety, technology transfer, and nuclear safeguards. Meetings and training courses: News about IAEA-sponsored symposia, seminars, and other meetings, as well as information about international meetings on atomic energy sponsored by other organizations, are updated on a daily basis. Press releases and statements: All IAEA press releases and media advisories since 1995 are accessible on the site. Topical and feature pages: In-depth coverage and links to information resources within and outside the IAEA are regularly given to selected topics of high international interest involving the IAEA. IAEA publications: listings and overviews of IAEA technical reports, safety standards, and other publications are updated as they are issued. Scientific and technical information: WorldAtom includes links (Reference Centre) to the International Nuclear Information System, IAEA's extensive bibliographic database of references and resources, to the nuclear database, and to departmental pages at IAEA that focus on IAEA programs and activities. IAEA documents: Electronic versions of official IAEA documents are added as they are issued. These documents include the texts and status lists of international conventions under IAEA auspices; IAEA information circulars to member states; IAEA annual reports (since 1995); and background reports and documents for the IAEA General Conference related to

Internal Spin Control, Squeezing and Decoherence in Ensembles of Alkali Atomic Spins

Science.gov (United States)

Norris, Leigh Morgan

Large atomic ensembles interacting with light are one of the most promising platforms for quantum information processing. In the past decade, novel applications for these systems have emerged in quantum communication, quantum computing, and metrology. Essential to all of these applications is the controllability of the atomic ensemble, which is facilitated by a strong coupling between the atoms and light. Non-classical spin squeezed states are a crucial step in attaining greater ensemble control. The degree of entanglement present in these states, furthermore, serves as a benchmark for the strength of the atom-light interaction. Outside the broader context of quantum information processing with atomic ensembles, spin squeezed states have applications in metrology, where their quantum correlations can be harnessed to improve the precision of magnetometers and atomic clocks. This dissertation focuses upon the production of spin squeezed states in large ensembles of cold trapped alkali atoms interacting with optical fields. While most treatments of spin squeezing consider only the case in which the ensemble is composed of two level systems or qubits, we utilize the entire ground manifold of an alkali atom with hyperfine spin f greater than or equal to 1/2, a qudit. Spin squeezing requires non-classical correlations between the constituent atomic spins, which are generated through the atoms' collective coupling to the light. Either through measurement or multiple interactions with the atoms, the light mediates an entangling interaction that produces quantum correlations. Because the spin squeezing treated in this dissertation ultimately originates from the coupling between the light and atoms, conventional approaches of improving this squeezing have focused on increasing the optical density of the ensemble. The greater number of internal degrees of freedom and the controllability of the spin-f ground hyperfine manifold enable novel methods of enhancing squeezing. In

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)

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

Proceedings of the international conference on mathematics and computations, reactor physics, and environmental analyses. Volume 1 and 2

International Nuclear Information System (INIS)

Anon.

1995-01-01

The International Conference on Mathematics and Computations, Reactor Physics, and Environmental Analyses marks the sixteenth biennial topical meeting of the Mathematics and Computation (M ampersand C) Division of the American Nuclear Society (ANS). This conference combines many traditional features of M ampersand C conferences with several new aspects. The meeting is, for the first time, being held in Portland, Oregon and sponsored by the ANS Eastern Washington Section. Three of the cosponsors - the ANS Reactor Physics Division, the European Nuclear Society, and the Atomic Energy Society of Japan - have participated in a series of such meetings, with very successful results. The fourth cosponsor, the ANS Environmental Science Division, is participating for the first time as a cosponsor of a M ampersand C topical meeting, as a result of the M ampersand C Division's decision to formally include the area of environmental analyses as a major focus of the conference, another 'first.' Separate abstracts have been submitted to the energy database for contributions to this conference

Under-Sodium Inspection Techniques for Reactor Internals of KALIMER-600 using Ultrasonic Waveguide Sensor

International Nuclear Information System (INIS)

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

2005-01-01

KALIMER-600 is a pool type liquid metal reactor (LMR) which is operated with a sodium coolant. The reactor internals of KALIMER-600 are submerged in a liquid sodium pool. As the liquid sodium is opaque to the light, a conventional visual inspection can not be used for observing the internal structures under a sodium condition. An under-sodium viewing (USV) technique using an ultrasonic wave should be applied for the observation of the refueling maneuver and the in-service inspection of the reactor internals. Under-sodium inspection technology utilizing ultrasonic waves has been widely developed for a visualization of the reactor core and internal components of LMR. Immersion sensors and waveguide sensors have been applied to the USV inspection. The immersion sensor has a precise imaging capability, but may have high temperature restrictions and an uncertain life. The waveguide sensor has the advantages of simplicity and reliability, but limited in its movement. The new plate-type waveguide sensor has been developed as a useful alternative to immersion sensors for USV applications. In the viewing and monitoring applications, a beam steering function of a waveguide sensor might be required. A new waveguide sensor and technique are being developed to overcome the limitations of a waveguide ultrasonic sensor. In this study, the under-sodium inspection techniques using the newly developed waveguide sensor for the reactor internal structures of KALIMER-600 is proposed

Status of research reactor spent fuel world-wide: Database summary

International Nuclear Information System (INIS)

Ritchie, I.G.

1996-01-01

Results complied in the research reactor spent fuel database are used to assess the status of research reactor spent fuel world-wide. Fuel assemblies, their types, enrichment, origin of enrichment and geological distribution among the industrialized and developed countries of the world are discussed. Fuel management practices in wet and dry storage facilities and the concerns of reactor operators about long-term storage of their spent fuel are presented and some of the activities carried out by the International Atomic Energy Agency to address the issues associated with research reactor spent fuel are outlined. (author). 4 refs, 17 figs, 4 tabs

SSC RIAR is the largest centre of research reactors

International Nuclear Information System (INIS)

Kalygin, V.V.

1997-01-01

The State Scientific Centre (SSC) ''Research Institute of Atomic Reactors'' (RIAR) is situated 100 km to the south-east from Moscow, in Dimitrovgrad, the Volga Region of the Russian Federation. SSC RIAR is the largest centre of research reactors in Russia. At present there are 5 types of reactor facilities in operation, including two NPP. One of the main tasks the Centre is the investigations on safety increase for power reactors. Broad international connections are available at the Institute. On the basis of the SSC RIAR during 3 years work has been done on the development of the branch training centre (TC) for the training of operation personnel of research and pilot reactors in Russia. (author). 3 tabs

SSC RIAR is the largest centre of research reactors

Energy Technology Data Exchange (ETDEWEB)

Kalygin, V V [State Scientific Centre, Research Inst. of Atomic Reactors (Russian Federation)

1997-10-01

The State Scientific Centre (SSC) ``Research Institute of Atomic Reactors`` (RIAR) is situated 100 km to the south-east from Moscow, in Dimitrovgrad, the Volga Region of the Russian Federation. SSC RIAR is the largest centre of research reactors in Russia. At present there are 5 types of reactor facilities in operation, including two NPP. One of the main tasks the Centre is the investigations on safety increase for power reactors. Broad international connections are available at the Institute. On the basis of the SSC RIAR during 3 years work has been done on the development of the branch training centre (TC) for the training of operation personnel of research and pilot reactors in Russia. (author). 3 tabs.

Improving nuclear safety at international research reactors: The Integrated Research Reactor Safety Enhancement Program (IRRSEP)

International Nuclear Information System (INIS)

Huizenga, David; Newton, Douglas; Connery, Joyce

2002-01-01

Nuclear energy continues to play a major role in the world's energy economy. Research and test reactors are an important component of a nation's nuclear power infrastructure as they provide training, experiments and operating experience vital to developing and sustaining the industry. Indeed, nations with aspirations for nuclear power development usually begin their programs with a research reactor program. Research reactors also are vital to international science and technology development. It is important to keep them safe from both accident and sabotage, not only because of our obligation to prevent human and environmental consequence but also to prevent corresponding damage to science and industry. For example, an incident at a research reactor could cause a political and public backlash that would do irreparable harm to national nuclear programs. Following the accidents at Three Mile Island and Chernobyl, considerable efforts and resources were committed to improving the safety posture of the world's nuclear power plants. Unsafe operation of research reactors will have an amplifying effect throughout a country or region's entire nuclear programs due to political, economic and nuclear infrastructure consequences. (author)

International centres of excellence based on research reactors

International Nuclear Information System (INIS)

Alldred, K.; Tozser, S.M.; Adelfang, P.

2013-01-01

A number of high flux research reactors were, or will be constructed. Each of these high flux facilities has the potential to be an important regional or International Centre of Excellence based on Research Reactors (ICERR) and scientific hub for research and materials investigations. Some are so organized currently, but for many there is a strongly national focus and scope for a significant expansion of their international role. There are manifold benefits of an expanded international role both for the ICERR's themselves and for the institutes that affiliate with them. These benefits include increased utilization and financial stability, increased international prestige, and enhanced scientific resources and capabilities. There are significant hurdles to obtaining the benefits from an expanded international role. For example, to achieve its full potential an ICERR must accommodate scientists from other nations, and include the plans and aspirations of the international community in the ICERR governance. The ICERR must also fully meet the national responsibilities for safety and security. Balancing these potentially conflicting requirements and finding a path through the organisational and legal issues is a significant challenge for any institute. The existing ICERR's therefore provide important case studies and examples of best practice that could inform the actions of other potential ICERR's. This paper describes an IAEA initiative to encourage and support the formation of new ICERR's, strengthen existing ones, and increase training resources available to Member States. The initiative will seek to share best practice and facilitate meetings and technical exchanges between the existing and potential ICERRs, and between the potential ICERR's and potential subscribing or affiliating institutes. (orig.)

International Centers of Excellence based on Research Reactors

International Nuclear Information System (INIS)

Alldred, K.; Tozser, S. M.; Adelfang, P.

2012-01-01

A number of high flux research reactors were, or will be constructed. Each of these high flux facilities has the potential to be an important regional or International Centre of Excellence based on Research Reactors (ICERR) and scientific hub for research and materials investigations. Some are so organized currently, but for many there is a strongly national focus and scope for a significant expansion of their international role. There are manifold benefits of an expanded international role both for the ICERR's themselves and for the institutes that affiliate with them. These benefits include increased utilization and financial stability, increased international prestige, and enhanced scientific resources and capabilities. There are significant hurdles to obtaining the benefits from an expanded international role. For example, to achieve its full potential an ICERR must accommodate scientists from other nations, and include the plans and aspirations of the international community in the ICERR governance. The ICERR must also fully meet the national responsibilities for safety and security. Balancing these potentially conflicting requirements and finding a path through the organisational and legal issues is a significant challenge for any institute. The existing ICERR's therefore provide important case studies and examples of best practice that could inform the actions of other potential ICERR's. This paper describes an IAEA initiative to encourage and support the formation of new ICERR's, strengthen existing ones, and increase training resources available to Member States. The initiative will seek to share best practice and facilitate meetings and technical exchanges between the existing and potential ICERRs, and between the potential ICERR's and potential subscribing or affiliating institutes. (authors)

ENS RRFM 2005: 9th international topical meeting on research reactor fuel management. Transactions

International Nuclear Information System (INIS)

2005-01-01

The ENS topical meeting on research reactor fuel management is an annual conference launched successfully in 1997. It has since then grown into well established international forum for the exchange and expertise on all significant aspects of the nuclear fuel cycle of research reactors. Oral presentations at this meeting were divided in the following four sessions: International Topics; Fuel Development, Qualification, Fabrication and Licensing; Reactor Operation, Fuel Safety and Core Conversion; Spent Fuel Management, Back-end Options, Transportation. The three poster sessions were devoted to fuel development, qualification, fabrication and licensing; reactor operation, fuel safety, core conversion, spent fuel; spent fuel management, fuel cycle back-end options, transportation

Gamma dose from activation of internal shields in IRIS reactor.

Science.gov (United States)

Agosteo, Stefano; Cammi, Antonio; Garlati, Luisella; Lombardi, Carlo; Padovani, Enrico

2005-01-01

The International Reactor Innovative and Secure is a modular pressurised water reactor with an integral design. This means that all the primary system components, such as the steam generators, pumps, pressuriser and control rod drive mechanisms, are located inside the reactor vessel, which requires a large diameter. For the sake of better reliability and safety, it is desirable to achieve the reduction of vessel embrittlement as well as the lowering of the dose beyond the vessel. The former can be easily accomplished by the presence of a wide downcomer, filled with water, which surrounds the core region, while the latter needs the presence of additional internal shields. An optimal shielding configuration is under investigation, for reducing the ex-vessel dose due to activated internals and for limiting the amount of the biological shielding. MCNP 4C calculations were performed to evaluate the neutron and the gamma dose during operation and the 60Co activation of various shields configurations. The gamma dose beyond the vessel from activation of its structural components was estimated in a shutdown condition, with the Monte Carlo code FLUKA 2002 and the MicroShield software. The results of the two codes are in agreement and show that the dose is sufficiently low, even without an additional shield.

Gamma dose from activation of internal shields in IRIS reactor

International Nuclear Information System (INIS)

Agosteo, S.; Cammi, A.; Garlati, L.; Lombardi, C.; Padovani, E.

2005-01-01

The International Reactor Innovative and Secure is a modular pressurised water reactor with an integral design. This means that all the primary system components, such as the steam generators, pumps, pressurizer and control rod drive mechanisms, are located inside the reactor vessel, which requires a large diameter. For the sake of better reliability and safety, it is desirable to achieve the reduction of vessel embrittlement as well as the lowering of the dose beyond the vessel. The former can be easily accomplished by the presence of a wide downcomer, filled with water, which surrounds the core region, while the latter needs the presence of additional internal shields. An optimal shielding configuration is under investigation, for reducing the ex-vessel dose due to activated internals and for limiting the amount of the biological shielding. MCNP 4C calculations were performed to evaluate the neutron and the gamma dose during operation and the 60 Co activation of various shields configurations. The gamma dose beyond the vessel from activation of its structural components was estimated in a shutdown condition, with the Monte Carlo code FLUKA 2002 and the MicroShield software. The results of the two codes are in agreement and show that the dose is sufficiently low, even without an additional shield. (authors)

Preliminary irradiation test results from the Yankee Atomic Electric Company reactor vessel test irradiation program

International Nuclear Information System (INIS)

Biemiller, E.C.; Fyfitch, S.; Campbell, C.A.

1993-01-01

The Yankee Atomic Electric Company test irradiation program was implemented to characterize the irradiation response of representative Yankee Rowe reactor vessel beltline plate materials and to remove uncertainties in the analysis of existing irradiation data on the Yankee Rowe reactor vessel steel. Plate materials each containing 0.24 w/o copper, but different nickel contents at 0.63 w/o and 0.19 w/o, were heat treated to simulate the Yankee vessel heat treatment (austenitized at 1800 deg F) and to simulate Regulatory Guide 1.99 database materials (austenitized at 1600 deg. F). These heat treatments produced different microstructures so the effect of microstructure on irradiation damage sensitivity could be tested. Because the nickel content of the test plates varied and the copper level was constant, the effect of nickel on irradiation embrittlement was also tested. Correlation monitor material, HSST-02, was included in the program to benchmark the Ford Nuclear Reactor (U. of Michigan Test Reactor) which had never been used for this type of irradiation program. Materials taken from plate surface locations (vs. 1/4T) were included to test whether or not the improved toughness properties of the plate surface layer, resulting from the rapid quench, is maintained after irradiation. If the improved properties are maintained, pressurized thermal shock calculations could utilize this margin. Finally, for one experiment, irradiations were conducted at two irradiation temperatures (500 deg. F and 550 deg. F) to determine the effect of irradiation temperature on embrittlement. The preliminary results of the irradiation program show an increase in T 30 shift of 69 deg. F for a decrease in irradiation temperature of 50 deg. F. The results suggest that for nickel bearing steels, the superior toughness of plate surface material is maintained after irradiation and for the copper content tested, nickel had no apparent effect on irradiation response. No apparent microstructure

Summary of the fifth international conference on reactor shielding

International Nuclear Information System (INIS)

Roussin, R.W.; Abbott, L.S.; Bartine, D.E.

1977-01-01

The Fifth International Conference on Reactor Shielding was held April 18-23, 1977 in Knoxville, Tennessee. The meeting was the largest in the series and attracted participants from 34 countries. The 10 invited papers and 10 of the contributed papers, selected as being representative of the Conference by the Technical Program Committee, are published in this issue of ATOMKERNENERGIE. This collection of papers demonstrates that the field of nuclear reactor shielding has developed into a mature discipline while retaining a definite vitality. (orig.) [de

International Atomic Energy Agency Publications. Catalogue 1986-1999

International Nuclear Information System (INIS)

2000-11-01

This catalogue lists all sales publications of the International Atomic Energy Agency issued from 1986 up to the end of 1999 and still available. Some earlier titles which form part of an established series or are still considered important have also been included. The catalogue is in CD-ROM format

Rotary reactor for atomic layer deposition on large quantities of nanoparticles

International Nuclear Information System (INIS)

McCormick, J. A.; Cloutier, B. L.; Weimer, A. W.; George, S. M.

2007-01-01

Challenges are encountered during atomic layer deposition (ALD) on large quantities of nanoparticles. The particles must be agitated or fluidized to perform the ALD surface reactions in reasonable times and to prevent the particles from being agglomerated by the ALD film. The high surface area of nanoparticles also demands efficient reactant usage because large quantities of reactant are required for the surface reactions to reach completion. The residence time of the reactant in a fluidized particle bed reactor may be too short for high efficiency if the ALD surface reactions have low reactive sticking coefficients. To address these challenges, a novel rotary reactor was developed to achieve constant particle agitation during static ALD reactant exposures. In the design of this new reactor, a cylindrical drum with porous metal walls was positioned inside a vacuum chamber. The porous cylindrical drum was rotated by a magnetically coupled rotary feedthrough. By rotating the cylindrical drum to obtain a centrifugal force of less than one gravitational force, the particles were agitated by a continuous 'avalanche' of particles. In addition, an inert N 2 gas pulse helped to dislodge the particles from the porous walls and provided an efficient method to purge reactants and products from the particle bed. The effectiveness of this rotary reactor was demonstrated by Al 2 O 3 ALD on ZrO 2 particles. A number of techniques including transmission electron microscopy, Fourier transform infrared spectroscopy, and x-ray photoelectron spectroscopy confirmed that the Al 2 O 3 ALD film conformally coats the ZrO 2 particles. Combining static reactant exposures with a very high surface area sample in the rotary reactor also provides unique opportunities for studying the surface chemistry during ALD

Annual Report to Congress of the Atomic Energy Commission for 1965

Energy Technology Data Exchange (ETDEWEB)

Seaborg, Glenn T.

1966-01-31

The document represents the 1965 Annual Report of the Atomic Energy Commission (AEC) to Congress. The report opens with a Foreword - a letter from President Lyndon B. Johnson. The main portion is divided into 3 major sections for 1965, plus 10 appendices and the index. Section names and chapters are as follows. Part One reports on Developmental and Promotional Activities with the following chapters: (1) The Atomic Energy Program - 1965; (2) The Industrial Base ; (3) Industrial Relations; (4) Operational Safety; (5) Source and Special Nuclear Materials Production; (6) The Nuclear Defense Effort; (7) Civilian Nuclear Power; (8) Nuclear Space Applications; (9) Auxiliary Electrical Power for Land and Sea; (10) Military Reactors; (11) Advanced Reactor Technology and Nuclear Safety Research; (12) The Plowshare Program; (13) Isotopes and Radiation Development; (14) Facilities and Projects for Basic Research; (15) International Cooperation; and, (16) Nuclear Education and Information. Part Two reports on Regulatory Activities with the following chapters: (1) Licensing and Regulating the Atom; (2) Reactors and other Nuclear Facilities; and, (3) Control of Radioactive Materials. Part Three reports on Adjudicatory Activities.

PREFACE: 8th Asian International Seminar on Atomic and Molecular Physics (AISAMP)

Science.gov (United States)

Williams, Jim F.; Buckman, Steve; Bieske, Evan J.

2009-09-01

These proceedings arose from the 8th Asian International Seminar on Atomic and Molecular Physics (AISAMP) which was held at the University of Western Australia 24-28 November 2008. The history of AISAMP (Takayanagi and Matsuzawa 2002) recognizes its origin from the Japan-China meeting of 1985, and the first use of the name 'The First Asian International Seminar on Atomic and Molecular Physics (AISAMP)' in 1992. The initial attendees, Japan and China, were joined subsequently by scientists from Korea, Taiwan, India, Australia and recently by Malaysia, Thailand, Vietnam, Turkey Iran, UK and USA. The main purpose of the biennial AISAMP series is to create a wide forum for exchanging ideas and information among atomic and molecular scientists and to promote international collaboration. The scope of the AISAMP8 meeting included pure, strategic and applied research involving atomic and molecular structure and processes in all forms of matter and antimatter. For 2008 the AISAMP conference incorporated the Australian Atomic and Molecular Physics and Quantum Chemistry meeting. The topics for AISAMP8 embraced themes from earlier AISAMP meetings and reflected new interests, in atomic and molecular structures, spectroscopy and collisions; atomic and molecular physics with laser or synchrotron radiation; quantum information processing using atoms and molecules; atoms and molecules in surface physics, nanotechnology, biophysics, atmospheric physics and other interdisciplinary studies. The implementation of the AISAMP themes, as well as the international representation of research interests, is indicated both in the contents list of these published manuscripts as well as in the program for the meeting. Altogether, 184 presentations were made at the 8th AISAMP, including Invited Talks and Contributed Poster Presentations, of which 60 appear in the present Proceedings after review by expert referees in accordance with the usual practice of Journal of Physics: Conference Series of

Annual Report to Congress of the Atomic Energy Commission for 1971

Energy Technology Data Exchange (ETDEWEB)

Schlesinger, James R.

1971-01-31

The document represents the 1971 Annual Report of the Atomic Energy Commission (AEC) to Congress. The report opens with ''An Introduction to the Atomic Energy Programs during 1971'' followed by 3 Parts, each with various chapters, plus a final index. Parts and Chapters are as follows. Part One, Regulatory Activities, has Chapters (1) Licensing and Regulation; (2) Reactor Licensing; and (3) Materials Control. Part Two, Environmental Safety, has Chapters (4) Environmental Considerations; (5) Radioactive Wastes; and (6), Operational Safety. Part Three, Operating and Developmental Functions, has Chapters (7) National Defense; (8) Reactor Technology; (9) Nuclear Materials; (10) Applications Research; (11) Basic Research; (12) International Affairs; and, (13) Educational and Administrative.

Training and research reactor facility longevity extension program

International Nuclear Information System (INIS)

Carriveau, G.W.

1991-01-01

Since 1943, over 550 training and research reactors have been in operation. According to statistics from the International Atomic Energy Agency, ∼325 training and research reactors are currently in service. This total includes a wide variety of designs covering a range of power and research capabilities located virtually around the world. A program has been established at General Atomics (GA) that is dedicated to the support of extended longevity of training and research reactor facilities. Aspects of this program include the following: (1) new instrumentation and control systems; (2) improved and upgraded nuclear monitoring and control channels; (3) facility testing, repair and upgrade services that include (a) pool or tank integrity, (b) cooling system, and (c) water purification system; (4) fuel element testing procedures and replacement; (5) control rod drive rebuilding and upgrades; (6) control and monitoring system calibration and repair service; (7) training services, including reactor operations, maintenance, instrumentation calibration, and repair; and (8) expanded or new uses such as neutron radiography and autoradiography, isotope production, nuclear medicine, activation analysis, and material properties modification

Aging management of PWR reactor internals in U.S. plants

International Nuclear Information System (INIS)

Amberge, K.J.; Demma, A.

2015-01-01

This paper describes the development, aging management strategies and inspection results of the Pressurized Water Reactor (PWR) vessel internals inspection and evaluation guidelines. The goal of these guidelines is to provide PWR owners with robust aging management strategies to monitor degradation of internals components to support life extension as well as the current period of operation and power up-rate activities. The implementation of these guidelines began in 2010 within the U.S. PWR fleet and several examinations have been performed since. Examples of inspection results are presented for selected vessel internals components and are compared with simulation results. In summary, to date there have been no observations of austenitic stainless steel stress corrosion cracking (SCC), which is consistent with expectations based on the current understanding of the mechanism. Observations of irradiation assisted stress corrosion cracking (IASCC) have been limited and only found in baffle former bolting. Additionally, no macroscopic effects or global observations of void swelling impacts on general conditions of reactor internal hardware have been observed. (authors)

Comparative study WWER-440/230 reactors and the Juragua-based WWER-440/318 project

International Nuclear Information System (INIS)

1994-01-01

The work consists in a comparative analysis between WWER-440/230 and WWER-440/318 type reactors on the basis of the major findings and recommendations given in the extra budgetary Programme of the International Atomic Energy Agency on the safety of these reactors and summarized in the document 'The safety of WWER-440 model 230 nuclear power plant'. The fundamental emphasis was done in the design concept of these reactors

Process for dissolving the radioactive corrosion products from internal surfaces in nuclear reactors

International Nuclear Information System (INIS)

Brown, W.W.

1976-01-01

This invention concerns a process for dissolving in the coolant flowing in a reactor the radioactive substances from the corrosion of the internal surfaces of the reactor to which they cling. When a reactor is operating, the fission occurring in the fuel generates gases and fission substances, such as iodine 131 and 133, cesium 134 and 137, molybdenum 99, xenon 133 and activates the structural materials of the reactor such as nickel by giving off cobalt 58 and similar substances. Under this invention an oxygen rich solution is injected in the reactor coolant after the temperature and pressure reduction stage, during the preparation prior to refuelling and repairs. The oxygen in the solution speeds up the release of cobalt 58 and other radioactive substances from the internal surfaces of the reactor and their dissolving in the oxygenated cold coolant at the start of the cooling procedures of the installation. This allows them to be removed by an ion exchanger before the reactor is emptied. By utilising this process, about half a day may be gained in refuelling time when this has to be done once a week [fr

International Nuclear Information System 1983-1996. International Atomic Energy Agency Publications

International Nuclear Information System (INIS)

1997-03-01

This catalogue lists all sales publications of the International Atomic Energy Agency dealing with the International Nuclear Information System (INIS). INIS was established in 1969 to announced the scientific literature published worldwide on the peaceful uses of nuclear energy. All books are published in English. It should be noted that prices of books are quoted in Austrian Schillings. The prices do not include local taxes and are subject to change without notice. All books in this catalogue are 21 x 30 cm, paper-bound, unless otherwise stated. In addition all books in this catalogue, except for the INIS Input Training Kit, are available on microfiche. For information on the microfiche versions, contact the INIS Clearinghouse of the IAEA

Department of Atomic Energy: Annual report, 1983-84

International Nuclear Information System (INIS)

1984-01-01

The annual report of the Department of Atomic Energy for the financial year 1983-84 describes its activities under the headings: Nuclear Power, Research and Development, Public Sector Undertakings, and Other Activities. The report surveys: (1) the performance of nuclear power plants at Tarapur, Kota and Kalpakkam, heavy water plants, fuel fabrication and reprocessing plants, and waste management facilities, (2) the research and development activities of Bhabha Atomic Research Centre at Bombay and its constituent units at various locations in the country, Reactor Research Centre at Kalpakkam, the aided institutes, namely, Tata Institute of Fundamental Research and Tata Memorial Centre, both at Bombay, and Saha Institute of Nuclear Physics at Calcutta, (3) performance of public sector undertakings: Indian Rare Earths Ltd., Uranium Corporation of India Ltd., and Electronics Corporation of India Ltd., (4) progress of nuclear power projects at Narora and Kakrapar, Orissa Sand Complex Project, MHD project at Tiruchirapalli, DHRUVA (formerly known as R-5) project at Bombay, Fast Breeder Test Reactor and 500 MW Prototype Fast Breeder Reactor projects at Kalpakkam, and heavy water projects at Thal-Vaishet and Manuguru, and (5) other activities including technology transfer; training; service to industry, agriculture and medicine in use of radioisotopes and radiation, export of radioisotopes, allied products and nuclear instruments; international relations; countrywide radiation safety programme, exploration of atomic minerals; information and publicity etc. An Atomic Energy Regulatory Board was established during the report year for the special purpose of carrying out regulatory and safety functions specified in the Atomic Energy Act of the Government of India. (M.G.B.)

Elevated temperature design of KALIMER reactor internals accounting for creep and stress-rupture effects

International Nuclear Information System (INIS)

Koo, Gyeong Hoi; Yoo, Bong

2000-01-01

In most LMFBR (Liquid Metal Fast Breed Reactor) design, the operating temperature is very high and the time-dependent creep and stress-rupture effects become so important in reactor structural design. Therefore, unlike with conventional PWR, the normal operating conditions can be basically dominant design loading because the hold time at elevated temperature condition is so long and enough to result in severe total creep ratcheting strains during total service lifetime. In this paper, elevated temperature design of the conceptually designed baffle annulus regions of KALIMER (Korea Advanced Liquid Metal Reactor) reactor internal structures is carried out for normal operating conditions which have the operating temperature 530 deg. C and the total service lifetime of 30 years. For the elevated temperature design of reactor internal structures, the ASME Code Case N-201-4 is used. Using this code, the time-dependent stress limits, the accumulated total inelastic strain during service lifetime, and the creep-fatigue damages are evaluated with the calculation results by the elastic analysis under conservative assumptions. The application procedures of elevated temperature design of the reactor internal structures using ASME code case N-201-4 with the elastic analysis method are described step by step in detail. This paper will be useful guide for actual application of elevated temperature design of various reactor types accounting for creep and stress-rupture effects. (author)

IAEA Launches Expert Advisory Service for Research Reactor Infrastructure, First Mission to Nigeria

International Nuclear Information System (INIS)

2018-01-01

The International Atomic Energy Agency (IAEA) has launched a new peer review service to assist Member States in the development of infrastructure for nuclear research reactors, expanding the range of its expert advisory missions. The first Integrated Nuclear Infrastructure Review for Research Reactors (INIR-RR) was conducted this week in Nigeria at the invitation of the Government, which is planning to build the country’s second research reactor. Research reactors are used for research, development, education and training. They play a vital role across several fields, producing radioisotopes used in research, medicine, industry and agriculture. Operation of a research reactor requires a national infrastructure — including a legal and regulatory framework — to ensure that national and international obligations are met during planning, design, construction, operation and decommissioning.

Internally Heated Screw Pyrolysis Reactor (IHSPR) heat transfer performance study

Science.gov (United States)

Teo, S. H.; Gan, H. L.; Alias, A.; Gan, L. M.

2018-04-01

1.5 billion end-of-life tyres (ELT) were discarded globally each year and pyrolysis is considered the best solution to convert the ELT into valuable high energy-density products. Among all pyrolysis technologies, screw reactor is favourable. However, conventional screw reactor risks plugging issue due to its lacklustre heat transfer performance. An internally heated screw pyrolysis reactor (IHSPR) was developed by local renewable energy industry, which serves as the research subject for heat transfer performance study of this particular paper. Zero-load heating test (ZLHT) was first carried out to obtain the operational parameters of the reactor, followed by the one dimensional steady-state heat transfer analysis carried out using SolidWorks Flow Simulation 2016. Experiments with feed rate manipulations and pyrolysis products analyses were conducted last to conclude the study.

Use and Storage of Test and Operations Data from the High Temperature Test Reactor Acquired by the US Government from the Japan Atomic Energy Agency

International Nuclear Information System (INIS)

Gougar, Hans

2010-01-01

This document describes the use and storage of data from the High Temperature Test Reactor (HTTR) acquired from the Japan Atomic Energy Agency (JAEA) by the U.S. Government for high temperature reactor research under the Next Generation Nuclear Plant (NGNP) Project. Data acquired from foreign entities by the United States Government is subject to protections and restrictions that depend upon the nature of the data and its intended use. The storage and use of nuclear data is subject to international agreements, federal laws and Department of Energy regulations. Security systems and personnel are deployed at Department of Energy national laboratories in order to ensure that these laws are observed. Laboratory employees undergo extensive training in operational security to protect all sensitive and classified information. This document describes the facilities at which data from the High Temperature Test Reactor will be used. It also described the procedures and policies that ensure that this data is used only by authorized personnel and only for the purposes for which it is intended.

An investigation of structural design methodology for HTGR reactor internals with ceramic materials (Contract research)

International Nuclear Information System (INIS)

Sumita, Junya; Shibata, Taiju; Nakagawa, Shigeaki; Iyoku, Tatsuo; Sawa, Kazuhiro

2008-03-01

To advance the performance and safety of HTGR, heat-resistant ceramic materials are expected to be used as reactor internals of HTGR. C/C composite and superplastic zirconia are the promising materials for this purpose. In order to use these new materials as reactor internals in HTGR, it is necessary to establish a structure design method to guarantee the structural integrity under environmental and load conditions. Therefore, C/C composite expected as reactor internals of VHTR is focused and an investigation on the structural design method applicable to the C/C composite and a basic applicability of the C/C composite to representative structures of HTGR were carried out in this report. As the results, it is found that the competing risk theory for the strength evaluation of the C/C composite is applicable to design method and C/C composite is expected to be used as reactor internals of HTGR. (author)

IAEA activities in nuclear reactor simulation for educational purposes

International Nuclear Information System (INIS)

Badulescu, A.; Lyon, R.

2001-01-01

The International Atomic Energy Agency (IAEA) has established a programme in nuclear reactor simulation computer programs to assist its Member States in education and training. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the supply or development of simulation programs and training material, sponsors training courses and workshops, and distributes documentation and computer programs. Currently, the IAEA has simulation programs available for distribution that simulate the behaviour of BWR, PWR and HWR reactor types. (authors)

Application of JAERI research reactors to education

International Nuclear Information System (INIS)

Ogawa, Shigeru; Morozumi, Minoru

1987-01-01

At the dawning of the atomic age in Japan, training on reactor operation and reactor engineering experiments has been started in 1958 using JRR-1 (a 50 kW water boiler type reactor with liquid fuel), which was the first research reactor in Japan. The role of the training has been transferred to JRR-4 (a 3500 kW swimming pool type reactor with ETR type fuel) since 1969 due to the decommission of JRR-1. The training courses which have been held are: JRR-1 Short-Term Course for Operation (1958 ∼ 1963) General Course (1961 ∼ ) Reactor Engineering Course (1976 ∼ ) Training Course in Nuclear Technology (International course)(1986 ∼ ). And individual training concerning research reactors for the participants of scientist exchange program sponsored by Science and Technology Agency and of bilateral agreement have been initiated in 1985. The graduates of these courses work as staff members in various fields in nuclear industry. (author)

Design features to facilitate IAEA safeguards at light water reactors

International Nuclear Information System (INIS)

Pasternak, T.; Glancy, J.; Goldman, L.; Swartz, J.

1981-01-01

Several studies have been performed recently to identify and analyze light water reactor (LWR) features that, if incorporated into the facility design, would facilitate the implementation of International Atomic Energy Agency (IAEA) safeguards. This paper presents results and conclusions of these studies. 2 refs

Borated stainless steel storage project to the spent fuel of the IEA-R1 reactor

International Nuclear Information System (INIS)

Rodrigues, Antonio Carlos Iglesias; Madi Filho, Tufic; Ricci Filho, Walter

2013-01-01

The IEA-R1 research reactor operates in a regimen of 64h weekly, at the power of 4.5 MW. In these conditions, the racks to the spent fuel elements have less than half of its initial capacity. Thus, maintaining these operating circumstances, the storage will have capacity for approximately six years. Whereas the estimated useful life of the IEA-R1 is around twenty years, it will be necessary to increase the storage capacity for the spent fuel. Dr. Henrik Grahn, expert of the International Atomic Energy Agency on wet storage, visiting the IEA-R1 Reactor (September/2012) made some recommendations: among them, the design and installation of racks made with borated stainless steel and internally coated with an aluminum film, so that corrosion of the fuel elements would not occur. This work objective is the project of high capacity storage for spent fuel elements, using borated stainless steel, to answer the Reactor IEA-R1 demand and the security requirements of the International Atomic Energy Agency. (author)

Borated stainless steel storage project to the spent fuel of the IEA-R1 reactor

Energy Technology Data Exchange (ETDEWEB)

Rodrigues, Antonio Carlos Iglesias; Madi Filho, Tufic; Ricci Filho, Walter, E-mail: acirodri@ipen.br, E-mail: tmfilho@ipen.br, E-mail: wricci@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2013-07-01

The IEA-R1 research reactor operates in a regimen of 64h weekly, at the power of 4.5 MW. In these conditions, the racks to the spent fuel elements have less than half of its initial capacity. Thus, maintaining these operating circumstances, the storage will have capacity for approximately six years. Whereas the estimated useful life of the IEA-R1 is around twenty years, it will be necessary to increase the storage capacity for the spent fuel. Dr. Henrik Grahn, expert of the International Atomic Energy Agency on wet storage, visiting the IEA-R1 Reactor (September/2012) made some recommendations: among them, the design and installation of racks made with borated stainless steel and internally coated with an aluminum film, so that corrosion of the fuel elements would not occur. This work objective is the project of high capacity storage for spent fuel elements, using borated stainless steel, to answer the Reactor IEA-R1 demand and the security requirements of the International Atomic Energy Agency. (author)

Inspection methods for physical protection Task III review of other agencies' physical security activities for research reactors

International Nuclear Information System (INIS)

In Task I of this project, the current Nuclear Regulatory Commission (NRC) position-on physical security practices and procedures at research reactors were reviewed. In the second task, a sampling of the physical security plans was presented and the three actual reactor sites described in the security plans were visited. The purpose of Task III is to review other agencies' physical security activities for research reactors. During this phase, the actions, procedures and policies of two domestic and two foreign agencies other than the NRC that relate to the research reactor community were examined. The agencies examined were: International Atomic Energy Agency; Canadian Atomic Energy Control Board; Department of Energy; and American Nuclear Insurers

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

Energy Technology Data Exchange (ETDEWEB)

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

2008-09-01

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

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

International Nuclear Information System (INIS)

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

2008-01-01

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

Environment. 1980-1994. International Atomic Energy Agency publications

International Nuclear Information System (INIS)

1995-06-01

The catalogue lists all publications of the International Atomic Energy Agency dealing with the Environment issued during the period 1980-1993. The major subjects covered include: effect of agrochemical residues on soils and aquatic ecosystems, application of radioisotopes in conservation of the environment, siting of nuclear power plants, environmental isotope data and environmental contamination due to nuclear accidents

Effects of neutron radiation and residual stresses on the corrosion of welds in light water reactor internals

International Nuclear Information System (INIS)

Schaaf, Bob van der; Gavillet, Didier; Lapena, Jesus; Ohms, Carsten; Roth, Armin; Dyck, Steven van

2006-01-01

After many years of operation in Light Water Reactors (LWR) Irradiation Assisted Stress Corrosion Cracking (IASCC) of internals has been observed. In particular the heat-affected zone (HAZ) has been associated with IASCC attack. The welding process induces residual stresses and micro-structural modifications. Neutron irradiation affects the materials response to mechanical loading. IASCC susceptibility of base materials is widely studied, but the specific conditions of irradiated welds are rarely assessed. Core component relevant welds of Type 304 and 347 steels have been fabricated and were irradiated in the High Flux Reactor (HFR) in Petten to 0.3 and 1 dpa (displacement per atom). In-service welds were cut from the thermal shield of the decommissioned BR-3 reactor. Residual stresses, measured using neutron diffraction, ring core tests and X-ray showed residual stress levels up to 400 MPa. Micro-structural characterization showed higher dislocation densities in the weld and HAZ. Neutron radiation increased the dislocation density, resulting in hardening and reduced fracture toughness. The sensitization degree of the welds, measured with the electrochemical potentio-dynamic reactivation method, was negligible. The Slow Strain Rate Tensile (SSRT) tests, performed at 290 deg. C in water with 200 ppb dissolved oxygen, (DO), did not reveal inter-granular cracking. Inter-granular attack of in-service steel is observed in water with 8 ppm (DO), attributed not only to IASCC, but also to IGSCC from thermal sensitization during fabrication. Stress-relieve annealing has caused Cr-grain boundary precipitation, indicating the sensitization. The simulated internal welds, irradiated up to 1.0 dpa, did not show inter-granular cracking with 8 ppm DO. (authors)

International bulletin on atomic and molecular data for fusion. No. 52

Energy Technology Data Exchange (ETDEWEB)

Stephens, J A [ed.

1997-08-01

This bulletin is published by the International Atomic Energy Agency to provide atomic and molecular data relevant to fusion research and technology. In part 1 the indexed papers are listed separately for (i) structure and spectra (energy levels, wavelengths, transition probabilities, oscillator strengths, interatomic potentials); (ii) atomic and molecular collisions (photon collisions, electron collisions, heavy-particle collisions); and (iii) surface interactions (sputtering, chemical reactions, trapping and detrapping, adsorption, desorption, reflection, and secondary electron emission). Part 2 contains the bibliographic data, essentially for the above listed topics.

International bulletin on atomic and molecular data for fusion. No. 52

International Nuclear Information System (INIS)

Stephens, J.A.

1997-08-01

This bulletin is published by the International Atomic Energy Agency to provide atomic and molecular data relevant to fusion research and technology. In part 1 the indexed papers are listed separately for (i) structure and spectra (energy levels, wavelengths, transition probabilities, oscillator strengths, interatomic potentials); (ii) atomic and molecular collisions (photon collisions, electron collisions, heavy-particle collisions); and (iii) surface interactions (sputtering, chemical reactions, trapping and detrapping, adsorption, desorption, reflection, and secondary electron emission). Part 2 contains the bibliographic data, essentially for the above listed topics

The international dimensions of nuclear safety standards

International Nuclear Information System (INIS)

Reed, J.M.

1992-01-01

The paper reviews the activities of the major international organisations in the field of nuclear safety standards; the International Atomic Energy Agency (IAEA), the OECD's Nuclear Energy Agency (NEA) and the Commission of the European Communities. Each organisation encourages the concept of international nuclear safety standards. After Chernobyl, there were calls for some form of binding international nuclear safety standards. Many Member States of IAEA accepted these Codes as a suitable basis for formulating their national safety standards, but the prevailing view was that voluntary compliance with the Codes was the preferred path. With few reactor vendors in a limited international market, the time may be approaching when an internationally licensable nuclear reactor is needed. Commonly accepted safety standards would be a prerequisite. The paper discusses the issues involved and the complexities of standards making in the international arena. (author)

Determination of internationally controlled materials according to provisions of the law for the regulations of nuclear source materials, nuclear fuel materials and reactors

International Nuclear Information System (INIS)

1981-01-01

This rule is established under the provisions of the law concerning the regulation of nuclear raw materials, nuclear fuel materials and reactors, and the former notification No. 26, 1961, is hereby abolished. Internationally regulated goods under the law are as follows: nuclear raw materials, nuclear fuel materials and moderator materials transferred by sale or other means from the governments of the U.S., U.K., Canada, Australia and France or the persons under their jurisdictions according to the agreements concluded between the governments of Japan and these countries, respectively, the nuclear fuel materials recovered from these materials or produced by their usage, nuclear reactors, the facilities and heavy water transferred by sale or other means from these governments or the persons under their jurisdictions, the nuclear fuel materials produced by the usage of such reactors, facilities and heavy water, the nuclear fuel materials sold by the International Atomic Energy Agency under the contract between the Japanese government and the IAEA, the nuclear fuel materials recovered from these materials or produced by their usage, the heavy water produced by the facilities themselves transferred from the Canadian government, Canadian governmental enterprises or the persons under the jurisdiction of the Canadian government or produced by the usage of these facilities, etc. (Okada, K.)

Lattice cell and full core physics of internally cooled annular fuel in heavy water moderated reactors

Energy Technology Data Exchange (ETDEWEB)

Armstrong, J.; Hamilton, H.; Hyland, B. [Atomic Energy of Canada Limited, Chalk River Laboratories, Chalk River, Ontario, K0J 1J0 (Canada)

2013-07-01

A program is underway at Atomic Energy of Canada Limited (AECL) to develop a new fuel bundle concept to enable greater burnups for PT-HWR (pressure tube heavy water reactor) cores. One option that AECL is investigating is an internally cooled annular fuel (ICAF) element concept. ICAF contains annular cylindrical pellets with cladding on the inner and outer diameters. Coolant flows along the outside of the element and through the centre. With such a concept, the maximum fuel temperature as a function of linear element rating is significantly reduced compared to conventional, solid-rod type fuel. The preliminary ICAF bundle concept considered in this study contains 24 half-metre long internally cooled annular fuel elements and one non-fuelled centre pin. The introduction of the non-fuelled centre pin reduces the coolant void reactivity (CVR), which is the increase in reactivity that occurs on voiding the coolant in accident scenarios. Lattice cell and full core physics calculations of the preliminary ICAF fuel bundle concept have been performed for medium burnups of approximately 18 GWd/tU using WIMS-AECL and reactor fuel simulation program (RFSP). The results will be used to assist in concept configuration optimization. The effects of radial and axial core power distributions, linear element power ratings, refuelling rates and operational power ramps have been analyzed. The results suggest that burnups of greater than 18 GWd/tU can be achieved in current reactor designs. At approximately 18 GWd/tU, expected maximum linear element ratings in a PT-HWR with online-refuelling are approximately 90 kW/m. These conditions would be prohibitive for solid-rod fuel, but may be possible in ICAF fuel given the reduced maximum fuel temperature as a function of linear element rating. (authors)

Lattice cell and full core physics of internally cooled annular fuel in heavy water moderated reactors

International Nuclear Information System (INIS)

Armstrong, J.; Hamilton, H.; Hyland, B.

2013-01-01

A program is underway at Atomic Energy of Canada Limited (AECL) to develop a new fuel bundle concept to enable greater burnups for PT-HWR (pressure tube heavy water reactor) cores. One option that AECL is investigating is an internally cooled annular fuel (ICAF) element concept. ICAF contains annular cylindrical pellets with cladding on the inner and outer diameters. Coolant flows along the outside of the element and through the centre. With such a concept, the maximum fuel temperature as a function of linear element rating is significantly reduced compared to conventional, solid-rod type fuel. The preliminary ICAF bundle concept considered in this study contains 24 half-metre long internally cooled annular fuel elements and one non-fuelled centre pin. The introduction of the non-fuelled centre pin reduces the coolant void reactivity (CVR), which is the increase in reactivity that occurs on voiding the coolant in accident scenarios. Lattice cell and full core physics calculations of the preliminary ICAF fuel bundle concept have been performed for medium burnups of approximately 18 GWd/tU using WIMS-AECL and reactor fuel simulation program (RFSP). The results will be used to assist in concept configuration optimization. The effects of radial and axial core power distributions, linear element power ratings, refuelling rates and operational power ramps have been analyzed. The results suggest that burnups of greater than 18 GWd/tU can be achieved in current reactor designs. At approximately 18 GWd/tU, expected maximum linear element ratings in a PT-HWR with online-refuelling are approximately 90 kW/m. These conditions would be prohibitive for solid-rod fuel, but may be possible in ICAF fuel given the reduced maximum fuel temperature as a function of linear element rating. (authors)

The atomic energy basic law

International Nuclear Information System (INIS)

1977-01-01

The law establishes clearly the principles that Japan makes R and D, and utilizations of atomic energy only for the peaceful purposes. All the other laws and regulations concerning atomic energy are based on the law. The first chapter lays down the above mentioned objective of the law, and gives definitions of basic concepts and terms, such as atomic energy, nuclear fuel material, nuclear source material, nuclear reactor and radiation. The second chapter provides for the establishment of Atomic Energy Commission which conducts plannings and investigations, and also makes decisions concerning R and D, and utilizations of atomic energy. The third chapter stipulates for establishment of two government organizations which perform R and D of atomic energy developments including experiments and demonstrations of new types of reactors, namely, Atomic Energy Research Institute and Power Reactor and Nuclear Fuel Development Corporation. Chapters from 4th through 8th provide for the regulations on development and acquisition of the minerals containing nuclear source materials, controls on nuclear fuel materials and nuclear reactors, administrations of the patents and inventions concerning atomic energy, and also prevention of injuries due to radiations. The last 9th chapter requires the government and its appointee to compensate the interested third party for damages in relation to the exploitation of nuclear source materials. (Matsushima, A.)

The Iranian Atomic program - Energy- versus Safety policy? On the problem of international non-proliferation regime

International Nuclear Information System (INIS)

Klimas, Mirko

2007-01-01

The Iranian atomic program: An important part of national energy policy or military way out of the international safety dilemma? In detail the author investigates - against background of Neorealism Theory - the coherences between international non-proliferation regimes - from Atoms for Peace program of the 1950 years up to actual measures of the IAEA - and the development of the Iranian atomic program. Off from superficial discussions on intentions of the Iranian government the book lightens the background of an international policy, which not only favours Iran during production of nuclear weapons as also invites the country to a military utilization of its atomic program. (orig./GL)

Annual report of the Department of Atomic Energy 1975-76

International Nuclear Information System (INIS)

1976-01-01

The activities of the various constituent units of the Department of Atomic Energy such as the Bhabha Atomic Research Centre, Reactor Research Centre, Variable Energy Cyclotron, the power stations and a few others during the year 1975-76 are reported. The progress achieved in the field of atomic minerals, nuclear medicine, nuclear power, development of radioisotopes etc. are presented in detail. The responsibilities and achievements of the public sector undertakings under Department of Atomic Energy such as the Indian Rare Earth Ltd., Electronics Corporation of India Ltd., Uranium Corporation of India Ltd., are highlighted. Other activities such as planning and execution, economic and personnel health aspects, international relations etc. are also mentioned. (A.K.)

Plant maintenance and advanced reactors issue, 2008

Energy Technology Data Exchange (ETDEWEB)

Agnihotri, Newal [ed.

2009-09-15

The focus of the September-October issue is on plant maintenance and advanced reactors. Major articles/reports in this issue include: Technologies of national importance, by Tsutomu Ohkubo, Japan Atomic Energy Agency, Japan; Modeling and simulation advances brighten future nuclear power, by Hussein Khalil, Argonne National Laboratory, Energy and desalination projects, by Ratan Kumar Sinha, Bhabha Atomic Research Centre, India; A plant with simplified design, by John Higgins, GE Hitachi Nuclear Energy; A forward thinking design, by Ray Ganthner, AREVA; A passively safe design, by Ed Cummins, Westinghouse Electric Company; A market-ready design, by Ken Petrunik, Atomic Energy of Canada Limited, Canada; Generation IV Advanced Nuclear Energy Systems, by Jacques Bouchard, French Commissariat a l'Energie Atomique, France, and Ralph Bennett, Idaho National Laboratory; Innovative reactor designs, a report by IAEA, Vienna, Austria; Guidance for new vendors, by John Nakoski, U.S. Nuclear Regulatory Commission; Road map for future energy, by John Cleveland, International Atomic Energy Agency, Vienna, Austria; and, Vermont's largest source of electricity, by Tyler Lamberts, Entergy Nuclear Operations, Inc. The Industry Innovation article is titled Intelligent monitoring technology, by Chris Demars, Exelon Nuclear.

Safe operation of critical assemblies and research reactors

Energy Technology Data Exchange (ETDEWEB)

NONE

1960-09-15

Some countries have accumulated considerable experience in the operation of these reactors and have in the process developed safe practices. On the other hand, other countries which have recently acquired, or will soon acquire, such reactors do not have sufficient background of experience with them to have developed full knowledge regarding their safe operation. In this situation, the International Atomic Energy Agency has considered that it would be useful to make available to all its Member States a set of recommendations on the safe operation of these reactors, based on the accumulated experience and best practices. The Director General accordingly nominated a Pane Ion Safe Operation of Critical Assemblies and Research Reactors to assist the Agency's Secretariat in drafting such recommendations

Amendment of Atomic Energy Basic Law and the development of Atomic Energy Administration

International Nuclear Information System (INIS)

Ochi, Kenji

1978-01-01

This article explains the key points of the major development of Atomic Energy Administration recently made by amendments of Atomic Energy Basic Law and other two relating laws. These amendments passed through the Diet and were enacted on 7th, June, 1978. The aim of them is focussed on reinforcement and rearrangement of safety controls on nuclear reactors. Previously, although the approval of the installation plan with basic designs of a nuclear reactor has been done by Prime Minister, further approvals of detailed designs and process of construction works, as well as inspections before and after operation have been conducted by each responsible minister, respectively. That is, those controls for power reactors have been within jurisdiction of minister of Trade and Industry, and for nuclear ships' reactors minister of Transportation has been responsible. Under the new system, above mentioned ministers continue to exercise almost same controls over reactors within their jurisdiction respectively, however the new laws have established so-called ''double check'' principle in that: when each responsible minister approves the installation, detailed designs and further stages of construction and operation of the reactor, he should hear and pay a great regard for opinions of Atomic Energy Commission and Atomic Energy Safety Commission. The latter is newly established organization which has similar status and authority to the former. (J.P.N.)

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

ITER at the international conference on fusion reactor materials

International Nuclear Information System (INIS)

Kalinin, G.; Barabash, V.; Matera, R.

1998-01-01

The reports summarizes the topics of the eighth International Conference on Fusion Reactor Materials (ICFRM-8) which was held in Sendai, Japan, on 26-31 October 1997. The ICFRM is focused on the whole spectrum of materials and technologies to be applied in fusion reactors and related facilities. The total number of conference participants was over 500, representing 24 countries and about 600 oral and poster papers were presented at the conference. Three sessions were devoted to ITER materials: (i) Design-Materials Interface and ITER (oral session); (ii) ITER, Irradiation Facility and Technology, (poster session); (iii) ITER and Beyond (discussion session)

Load bearing capacities and elastic-plastic behavior of reactor vessel internals

International Nuclear Information System (INIS)

Watanabe, Keita; Nagase, Ryuichi

2017-01-01

Radial Support Keys (RSKs) are installed at the bottom of Reactor Vessel Internal (RVI) of Pressurized Water Reactor (PWR) and fit into Core Support Lugs of Reactor Pressure Vessel (RPV). This structure provides reactor core horizontal support and transmits the loads between RVI and RPV. RSK is one of the critical parts of RVI from the view point of earthquake-proof safety. In order to assure the structural integrity of Nuclear Reactor in case of massive earthquake, load bearing capacities of RSK are confirmed by static loading tests with reduced-scale mockups. In addition, collapse loads of actual components calculated by Limit Analyses are conservative enough compared to the load bearing capacities confirmed by the test. Thus, the methodology to calculate collapse load by Limit Analysis is applicable to evaluation of structural integrity for RSK. (author)

Progress in liquid metal fast reactor technology. Proceedings of the 28th meeting of the International Working Group on Fast Reactors

International Nuclear Information System (INIS)

1996-04-01

The key objectives and activities of Member State liquid metal fast reactor (LMFR) programmes are: Demonstration of effective designs; demonstration of system safety; demonstration of economic competitiveness with other power generation systems. The International Working Group on Fast Reactors (IWGFR) at its 1995 meeting observed that while some countries (as a result of static or falling power demand) are reducing the research and development programmes or delaying the commercial deployment of fast reactors, other countries are planning to introduce these reactors and are embarking on their own development programmes. In these circumstances the international exchange of information and experience is of increasing importance. These proceedings contain updated information from long standing members of the IWGFR and new information on the status of LMFR research and development from new members of the Group: Brazil, China, Republic of Kazakhstan and the Republic of Korea. Refs, figs, tabs

Progress in liquid metal fast reactor technology. Proceedings of the 28th meeting of the International Working Group on Fast Reactors

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-04-01

The key objectives and activities of Member State liquid metal fast reactor (LMFR) programmes are: Demonstration of effective designs; demonstration of system safety; demonstration of economic competitiveness with other power generation systems. The International Working Group on Fast Reactors (IWGFR) at its 1995 meeting observed that while some countries (as a result of static or falling power demand) are reducing the research and development programmes or delaying the commercial deployment of fast reactors, other countries are planning to introduce these reactors and are embarking on their own development programmes. In these circumstances the international exchange of information and experience is of increasing importance. These proceedings contain updated information from long standing members of the IWGFR and new information on the status of LMFR research and development from new members of the Group: Brazil, China, Republic of Kazakhstan and the Republic of Korea. Refs, figs, tabs.

Radiation therapy. 1990-2001. International Atomic Energy Agency publications

International Nuclear Information System (INIS)

2001-04-01

This catalog lists all sales publications of the International Atomic Energy Agency dealing with Radiation Therapy, and issued during the period 1 January 1990 - 30 April 2001. Most publications are issued in English, though some are also available in other languages. These are noted in the catalogue

The changing structure of the international commercial nuclear power reactor industry

International Nuclear Information System (INIS)

Forsberg, C.W.; Hill, L.J.; Reich, W.J.; Rowan, W.J.

1992-12-01

The objective of this report is to provide an understanding of the international commercial nuclear power industry today and how the industry is evolving. This industry includes reactor vendors, product lines, and utility customers. The evolving structure of the international nuclear power reactor industry implies different organizations making decisions within the nuclear power industry, different outside constraints on those decisions, and different priorities than with the previous structure. At the same time, cultural factors, technical constraints, and historical business relationships allow for an understanding of the organization of the industry, what is likely, and what is unlikely. With such a frame of reference, current trends and future directions can be more readily understood

Scale-model characterization of flow-induced vibrational response of FFTF reactor internals

International Nuclear Information System (INIS)

Ryan, J.A.; Mahoney, J.J.

1980-10-01

Fast Test Reactor core internal and peripheral components were assessed for flow-induced vibrational characteristics under scaled and simulated prototype flow conditions in the Hydraulic Core Mockup as an integral part of the Fast Test Reactor Vibration Program. The Hydraulic Core Mockup was an 0.285 geometric scale model of the Fast Test Reactor internals designed to simulate prototype vibrational and hydraulic characteristics. Using water to simulate sodium coolant, vibrational characteristics were measured and determined for selected model components over the scaled flow range of 36 to 110%. Additionally, in-situ shaker tests were conducted on selected Hydraulic Core Mockup outlet plenum components to establish modal characteristics. Most components exhibited resonant response at all test flow rates; however, the measured dynamic response was neither abnormal nor anomalously flow-rate dependent, and the predicted prototype components' response were deemed acceptable

Utilization of research reactors

International Nuclear Information System (INIS)

1962-01-01

About 200 research reactors are now in operation in different parts of the world, and at least 70 such facilities, which are in advanced stages of planning and construction, should be critical within the next two or three years. In the process of this development a multitude of problems are being encountered in formulating and carrying out programs for the proper utilization of these facilities, especially in countries which have just begun or are starting their atomic energy work. An opportunity for scientific personnel from different Member States to discuss research reactor problems was given at an international symposium on the Programing and Utilization of Research Reactors organized by the Agency almost immediately after the General Conference session. Two hundred scientists from 35 countries, as well as from the European Nuclear Energy Agency and EURATOM, attended the meeting which was held in Vienna from 16 to 21 October 1961

International Reactor Innovative and Secure (IRIS) summary

International Nuclear Information System (INIS)

Carelli, Mario D.

2001-01-01

The IRIS (International Reactor Innovative and Secure) reactor is described in the first part of the presentation. IRIS is a light water cooled reactor with an integral configuration, where steam generators, pumps and pressurizer are inside the reactor vessel. Partially funded by the DOE NERI program, IRIS is being developed by an international consortium of 16 organizations from seven countries. A key IRIS characteristic is its 'safety by design' approach which strives to eliminate, by design, as many accidents as possible rather than coping with their consequences. Initial returns are very positive; out of the eight Class IV accidents considered in the AP600 only one remains as a Class IV in IRIS, and at much reduced probability. Small-to-medium LOCAs have minimal consequences as the core remains safely under water for days, without the need for safety injection or water makeup. In spite of its novelty IRIS is firmly grounded on proven LWR technology and therefore a prototype is not needed to assure design certification. Rather, very extensive scaled tests will be performed to investigate the performance of in-vessel components such as steam generators and pumps, both individually and as interactive systems. Accident sequences will also be simulated and tested to prove IRIS safety by design claims. The first core fuel is less than 5% enriched and the fuel assembly is very similar to existing PWR assemblies, so there is no licensing challenge regarding the fuel. Because of the safety by design approach, yielding simplifications In design and accident management (e.g., IRIS does not have an emergency core cooling system), some accident scenarios are eliminated and others have lesser consequences. Thus, simplification and streamlining of the regulatory process might be possible. Risk informed regulation will be coupled with safety by design to show lower accident and damage probabilities. This could lead to a relaxation of siting regulatory requirements. It is

Department of Atomic Energy, annual report, 1980-81

International Nuclear Information System (INIS)

1981-01-01

The annual report of the Department of Atomic Energy (DAE) of the Government of India for the period of the fiscal year 1980-81 surveys the work of DAE, its various constituent units and aided institutions. The main thrust of the DAE's programme in the country is directed towards peaceful uses of atomic energy - primarily for generation of electric power and also for application of radioisotopes and radiation in medicine, agriculture, and industry. The research and development (R and D) activities of the Bhabha Atomic Research Centre (BARC) at Bombay, the major R and D establishment of DAE, in the fields of nuclear physics, solid state physics, chemistry and materials science, isotope and radiation applications, reactor technology and radioactive waste management are described in detail. The R and D activities of the Reactor Research Centre at Kalpakkam and the aided institutions such as the Tata Institute of Fundamental Research and the Tata Memorial Centre, both at Bombay, and the Saha Institute of Nuclear Physics at Calcutta are reviewed in brief. Progress of the MHD project, the heavy water plant projects, the thermal research reactor R-5 project at BARC and nuclear power plant projects at Narora and Kalpakkam is surveyed. Performance of industrial production units such as nuclear power stations at Tarapur and Kota, the Nuclear Fuel Complex at Hyderabad, Atomic Minerals Division, ISOMED - the radiation sterilisation plant for medical products, the Indian Rare Earths Ltd., the Electronics Corporation of India Ltd., and the Uranium Corporation of India Ltd., is reported. India's participation in the activities of the International Atomic Energy Agency and collaboration with other countries are also mentioned. (M.G.B.)

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

Energy Technology Data Exchange (ETDEWEB)

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

2011-09-01

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

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

International Nuclear Information System (INIS)

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

2016-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-04-22

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-07-01

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

Creating an international forum - The success story of the International Forum for Reactor Aging Management (IFRAM)

International Nuclear Information System (INIS)

Brenchley, D.L.; Bond, L.J.; Carpenter, C.E.; Hwang, I.S.; Martin, O.; Reister, R.; Shoji, T.; Tilley, R.

2012-01-01

As the benefits of extending the safe operating life of nuclear power plants has become more evident, so has the need to increase international cooperation in reactor aging management research. This paper describes how individuals and organizations from Asia, Europe, and North America teamed to create the International Forum for Reactor Aging Management (IFRAM). The mission of IFRAM is to facilitate the appropriate exchange of information among those parties and organizations around the world that are presently, or are planning to, address issues of nuclear power plant (NPP) systems, structures and components aging management. The main purpose of this paper is to describe the steps to success in creating IFRAM. The desired attributes, charter, and operational methods of IFRAM are described and the key contributors for creating IFRAM are identified. (author)

IAEA activities in nuclear reactors simulation for educational purposes

International Nuclear Information System (INIS)

Lyon, R.B.

1998-01-01

The International Atomic Energy Agency (IAEA) has established a programme in nuclear reactor simulation computer programs to assist its Member States in education and training. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the supply or development of simulation programs and training material, sponsors training courses and workshops, and distributes documentation and computer programs. Currently, the IAEA has two simulation programs: the Classroom-based Advanced Reactor Demonstrators package, and the Advanced Reactor Simulator. Both packages simulate the behaviour of BWR, PWR and HWR reactor types. For each package, the modeling approach and assumptions are broadly described, together with a general description of the operation of the computer programs. (author)

Modular High Temperature Gas-Cooled Reactor Safety Basis and Approach

Energy Technology Data Exchange (ETDEWEB)

David Petti; Jim Kinsey; Dave Alberstein

2014-01-01

Various international efforts are underway to assess the safety of advanced nuclear reactor designs. For example, the International Atomic Energy Agency has recently held its first Consultancy Meeting on a new cooperative research program on high temperature gas-cooled reactor (HTGR) safety. Furthermore, the Generation IV International Forum Reactor Safety Working Group has recently developed a methodology, called the Integrated Safety Assessment Methodology, for use in Generation IV advanced reactor technology development, design, and design review. A risk and safety assessment white paper is under development with respect to the Very High Temperature Reactor to pilot the Integrated Safety Assessment Methodology and to demonstrate its validity and feasibility. To support such efforts, this information paper on the modular HTGR safety basis and approach has been prepared. The paper provides a summary level introduction to HTGR history, public safety objectives, inherent and passive safety features, radionuclide release barriers, functional safety approach, and risk-informed safety approach. The information in this paper is intended to further the understanding of the modular HTGR safety approach. The paper gives those involved in the assessment of advanced reactor designs an opportunity to assess an advanced design that has already received extensive review by regulatory authorities and to judge the utility of recently proposed new methods for advanced reactor safety assessment such as the Integrated Safety Assessment Methodology.

Technical committee meeting on Liquid Metal Fast Reactor (LMFR) developments. 33rd annual meeting of the International Working Group on Fast Reactors (IWG-FR). Working material

International Nuclear Information System (INIS)

2000-01-01

Over the past 33 years, the IAEA has actively encouraged and advocated international cooperation in fast reactor technology. The present publication contains information on the status of fast reactor development and on worldwide activities in this advanced nuclear power technology during 1999/2000, as reported at the 33. annual meeting of the International Working Group on Fast Reactors. It is intended to provide information regarding the current status of LMFR development in IAEA Member States

Prospects for nuclear energy in the 21st century and the role of the International Atomic Energy Agency

International Nuclear Information System (INIS)

Juhn, P.E.; Kupitz, J.; Cleveland, J.; Lyon, R.B.

1998-01-01

In the second half of the 20th century nuclear power has evolved to an industry that supplies 17% of the world's electricity. A great deal has been achieved and many lessons have been learned. The past decade, however, has seen stagnation or slow growth in nuclear power plant construction except in East Asia. The turn of the century can be a turning point for nuclear power due to increasing world energy consumption, the need to reduce greenhouse gas emissions, and improvements in operation and economics of nuclear power plants. Advanced reactor designs are being developed so that the technology will be ready for the next century. Issues that have dampened support for nuclear power are being addressed through the emerging global nuclear safety culture, attention to nuclear waste disposal and a strengthened safeguards system. The International Atomic Energy Agency promotes technical information exchange and co-operation, provides a source of balanced, objective information, and publishes reports on the current status of reactor development. The Agency will continue to play a major role as the nuclear industry faces the challenges and opportunities of the 21st century. (author)

Atom for peace, code for war. The technology policy of the atomic power solution in Finland between 1955-1970

International Nuclear Information System (INIS)

Sarkikoski, T.

2011-01-01

This dissertation investigates the atomic power solution in Finland between 1955 - 1970. During these years a national arrangement for atomic energy technology evolved. The foundations of the Finnish atomic energy policy; the creation of basic legislation and the first governmental bodies, were laid between 1955 - 1965. In the late 1960's, the necessary technological and political decisions were made in order to purchase the first commercial nuclear reactor. A historical narration of this process is seen in the international context of 'atoms for peace' policies and Cold War history in general. The geopolitical position of Finland made it necessary to become involved in the balanced participation in international scientific-technical exchange and assistive nuclear programs. The Paris Peace Treaty of 1947 categorically denied Finland acquisition of nuclear weapons. Accordingly, from the 'Geneva year' of 1955, the emphasis was placed on peaceful purposes for atomic energy as well as on the education of national professionals in Finland. An initiative for the governmental atomic energy commission came from academia but the ultimate motive behind it was an anticipated structural change in the supply of national energy. Economically exploitable hydro power resources were expected to be built within ten years and atomic power was seen as a promising and complementing new energy technology. While importing fuels like coal was out of the question, because of scarce foreign currency, domestic uranium mineral deposits were considered as a potential source of nuclear fuel. Nevertheless, even then nuclear energy was regarded as just one of the possible future energy options. In the mid-1960 s a bandwagon effect of light water reactor orders was witnessed in the United States and soon elsewhere in the world. In Finland, two separate invitations for bids for nuclear reactors were initiated. This study explores at length both their preceding grounds and later phases. An

Internal corium catcher of a nuclear reactor

International Nuclear Information System (INIS)

Anatolii S Vlasov; Vladimir N Mineev; Aleksandr S Sidorov; Yuri A Zeigarnik

2005-01-01

Full text of publication follows: A corium catcher is one of the main devices of a nuclear reactor that provides corium melt and fission products retention within a containment during severe accidents. Several studies and design developments have shown that corium retention within a reactor vessel can be attained with a moderate capacity of the latter (up to 600 - 650 MW el.). With a higher reactor capacity external corium catchers are applied both at Russian (VVER-1000) and European (EPR) reactors. In the external catcher of a VVER-1000 reactor, most technological problems are solved due to using sacrificial material. They are as follows: (a) endo-thermal interaction of corium and sacrificial material reduces a level of the temperatures in the final melt pool; (b) solution in the melt of a great amount of the sacrificial material reduces the specific heat release density and the heat flux density at the boundaries of a melt; (c) due to changing of the oxide-component density an inverse stratification of the metallic and oxide components of the corium takes place, thus excluding heat-flux focusing in the zone of the metallic layer and making it possible to supply water on the free surface of the corium without a danger of incipience of the vapor explosion; (d) final oxidation of zirconium occurs without hydrogen generation. The above principles have been realized in the external catcher of the VVER- 1000 reactor at Tyanvan NPS that is presently under construction in China. Successfully solving of the problems concerning to the external catcher makes it possible to return on the new conceptual and technological basis to the idea of retention of the corium melt inside the vessel of a nuclear reactor of large capacity, that is, to provide the reactor vessel to play a role of an internal catcher. For this purpose, a reactor vessel is elongated by approximately two meters. In the lower part of the vessel, on elliptical bottom, pieces of sacrificial material are arranged

Aging management strategy for reactor internals of Korean nuclear power plants

International Nuclear Information System (INIS)

Hwang, Seong Sik; Kim, Soung Woo; Lee, Sam Lai; Hong, Seung Mo; Kim, Hong Pyo; Kim, Dong Jin; Lim, Yun Soo; Kim, Joung Soo; Jung, Man Kyo; Park, Jang Yul

2010-01-01

This report describes various factors on the IASCC of reactor internals in terms of fluence, stress, water chemistries and materials. Materials of each components of Korean nuclear power plants have been surveyed. A technical report for a management of reactor internals issued by EPRI was reviewed for a selection of most susceptible area among many components. Baffle former bolts are considered to be the most susceptible parts due to high irradiation level(fluence) and high tensile stress. Neutron fluence of Kori-1 and Kori-2 was calculated based on fuel exchange history, fuel performance and plant operation history. This report will be used for more advanced inspection and maintenance guidelines development by supplying inspection intervals and components (most susceptible regions) for the long term operation plants

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.)

24. International Conference on Atomic Collisions in Solids ICACS-24

International Nuclear Information System (INIS)

2010-01-01

This Book contains the abstracts of invited and contributed talks submitted for presentation at the 24 th International Conference on Atomic Collisions in Solids - ICACS-24. Out of nearly 200 submitted abstracts the International Programme Committee selected 46 oral and 89 poster contributions. Furthermore, 15 plenary invited lectures and the honorary Lindhard lecture are included in the scientific program. An additional tutorial day with 4 tutorial lectures is organised on Sunday prior to the Conference.

Reactor accidents and the environment

International Nuclear Information System (INIS)

Beattie, J.R.; Griffiths, R.F.; Kaiser, G.D.; Kinchin, G.H.

1978-01-01

This is a condensed version of a paper, entitled 'The Environmental Impact of Radioactive Releases from Accidents in Nuclear Power Reactors', by the authors, presented to the Nuclear Energy Panel of the International Atomic Energy Agency/United Nations Environmental Programme. Headings include - Effects of ionising radiation on man; number of deaths expected from leukaemia and other cancers; risk estimates for incidence of benign nodules and thyroid cancer; maximum permissible levels and emergency levels of radiation and radioactivity; ICRP recommended dose limits for members of the general public; atmospheric dispersion and modelling; ICRP emergency reference levels for 1 131 , Cs 137 , Ru 106 and Sr 90 ; environmental consequences of accidental releases from nuclear power reactors; environmental impact of accidents to Magnox gas-cooled reactors; environmental impact of accidents to advanced gas-cooled reactors; environmental impact of accidents to fast reactors; and nature of risks. consequences are examined in terms of early and late biological effects on man, and contamination of land areas. Serious accidents are of low probability of occurrence, and the risk of accidents to nuclear power reactors is estimated to be very small. 43 references. (U.K.)

Reactor technology

International Nuclear Information System (INIS)

Erdoes, P.

1977-01-01

This is one of a series of articles discussing aspects of nuclear engineering ranging from a survey of various reactor types for static and mobile use to mention of atomic thermo-electric batteries of atomic thermo-electric batteries for cardiac pacemakers. Various statistics are presented on power generation in Europe and U.S.A. and economics are discussed in some detail. Molten salt reactors and research machines are also described. (G.M.E.)

Pressurized water reactor simulator. Workshop material

International Nuclear Information System (INIS)

2003-01-01

The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and educational material and sponsors courses and workshops. The workshops are in two parts: techniques and tools for reactor simulator development; and the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA Training Course Series No. 12, 'Reactor Simulator Development' (2001). Course material for workshops using a WWER- 1000 reactor department simulator from the Moscow Engineering and Physics Institute, the Russian Federation is presented in the IAEA Training Course Series No. 21 'WWER-1000 Reactor Simulator' (2002). Course material for workshops using a boiling water reactor simulator developed for the IAEA by Cassiopeia Technologies Incorporated of Canada (CTI) is presented in the IAEA publication: Training Course Series No.23 'Boiling Water Reactor Simulator' (2003). This report consists of course material for workshops using a pressurized water reactor simulator

Jacking mechanism for upper internals structure of a liquid metal nuclear reactor

International Nuclear Information System (INIS)

Gillett, J.E.; Wineman, A.L.

1984-01-01

A jacking mechanism is described for raising the upper internals structure of a liquid metal nuclear reactor which jacking mechanism uses a system of gears and drive shafts to transmit force from a single motor to four mechanically synchronized ball jacks to raise and lower support columns which support the upper internals structure. The support columns have a pin structure which rides up and down in a slot in a housing fixed to the reactor head. The pin has two locking plates which can be rotated around the pin to bring bolt holes through the locking plates into alignment with a set of bolt holes in the housing, there being a set of such housing bolt holes corresponding to both a raised and a lowered position of the support column. When the locking plate is so aligned, a surface of the locking plate mates with a surface in the housing such that the support column is then supported by the locking plate and not by the ball jacks. Since the locking plates are to be installed and bolted to the housing during periods of reactor operation, the ball jacks need not be sized to react the large forces which occur or potentially could occur on the upper internals structure of the reactor during operation. The locking plates react these loads. The ball jacks, used only during refueling, can be smaller, which enable conventionally available equipment to fulfill the precision requirements for the task within available space

Hope over fear. The establishment of the International Atomic Energy Agency

International Nuclear Information System (INIS)

Forland, A.

1995-01-01

President Eisenhower's ''Atoms for Peace'' speech in 1953 started a process that led to the establishment of the International Atomic Energy Agency (IAEA). This study analyses the negotiations of the Statute of the Agency. It focuses on how the Agency's scope of action regarding safeguards was negotiated. The Statute is seen as a reflection of the dual purpose of the new Agency to promote peaceful uses of nuclear energy, and to control that peaceful nuclear activities receiving IAEA assistance are not diverted to military purposes. On one hand, the countries receiving assistance from the Agency accepted a degree of international control that represented a breakthrough in international relations. On the other hand, many countries strongly resented controls. Consequently, compromises had to be made in the course of the negotiations in order to reach a consensus. Thus, although the IAEA was established as a competent and technical body, the underlying compromises meant that its scope of actions was restricted. 6 refs

Hope over fear. The establishment of the International Atomic Energy Agency

Energy Technology Data Exchange (ETDEWEB)

Forland, A.

1995-12-01

President Eisenhower`s ``Atoms for Peace`` speech in 1953 started a process that led to the establishment of the International Atomic Energy Agency (IAEA). This study analyses the negotiations of the Statute of the Agency. It focuses on how the Agency`s scope of action regarding safeguards was negotiated. The Statute is seen as a reflection of the dual purpose of the new Agency to promote peaceful uses of nuclear energy, and to control that peaceful nuclear activities receiving IAEA assistance are not diverted to military purposes. On one hand, the countries receiving assistance from the Agency accepted a degree of international control that represented a breakthrough in international relations. On the other hand, many countries strongly resented controls. Consequently, compromises had to be made in the course of the negotiations in order to reach a consensus. Thus, although the IAEA was established as a competent and technical body, the underlying compromises meant that its scope of actions was restricted. 6 refs.

Reactor safety research and development in Chalk River Laboratories

Energy Technology Data Exchange (ETDEWEB)

Nitheanandan, T. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

2014-07-01

Atomic Energy of Canada Limited's Chalk River Laboratories provides three different services to stakeholders and customers. The first service provided by the laboratory is the implementation of Research and Development (R&D) programs to provide the underlying technological basis of safe nuclear power reactor designs. A significant portion of the Canadian R&D capability in reactor safety resides at Atomic Energy of Canada Limited's Chalk River Laboratories, and this capability was instrumental in providing the science and technology required to aid in the safety design of CANDU power reactors. The second role of the laboratory has been in supporting nuclear facility licensees to ensure the continued safe operation of nuclear facilities, and to develop safety cases to justify continued operation. The licensing of plant life extension is a key industry objective, requiring extensive research on degradation mechanisms, such that safety cases are based on the original safety design data and valid and realistic assumptions regarding the effect of ageing and management of plant life. Recently, Chalk River Laboratories has been engaged in a third role in research to provide the technical basis and improved understanding for decision making by regulatory bodies. The state-of-the-art test facilities in Chalk River Laboratories have been contributing to the R&D needs of all three roles, not only in Canada but also in the international community, thorough Canada's participation in cooperative programs lead by International Atomic Energy Agency and the OECD's Nuclear Energy Agency. (author)

Development status of PIUS/ISER - a inherently safe reactor for the international use

International Nuclear Information System (INIS)

Wakabayashi, Hiroaki

1987-01-01

It is just in early 1980s that LWR-based nuclear power has become a substantial power source. Though the safety level of nuclear power is always claimed to be sufficiently high by the industry, it rests on the idea of defense in depth, the calculation by probabilistic risk assessment (PRA) or probabilistic safety assessment (PSA). The TMI-2 and Chernobyl-4 accidents occurred in the industrially most advanced countries. In this paper, an alternative way to safe nuclear power is sought in so-called inherently safe reactors (ISR) including the LWR type PIUS/ISER. With proper consideration into the design of nuclear reactor plants, those can be made basically safe through the use of passive safe mechanism for their design. In short, an ISR is a nuclear power reactor which has passive and intrinsic core cooling capability and automatic shutdown capability. As the nuclear power reactors which are currently claimed to be inherently safe, there are the process inherent and ultimately safe reactor (PIUS) of ASEA-ATOM Sweden and the inherently safe and economical reactor (ISER) of the University of Tokyo, Japan, of LWR type. The current status of the development, the reliability, and some technical problems of ISER/PIUS and the attitude of various countries toward ISER/PIUS are described. (Kako, I.)

A calculation of internal kinetic energy and polarizability of compressed argon from the statistical atom model

NARCIS (Netherlands)

Seldam, C.A. ten; Groot, S.R. de

1952-01-01

From Jensen's and Gombás' modification of the statistical Thomas-Fermi atom model, a theory for compressed atoms is developed by changing the boundary conditions. Internal kinetic energy and polarizability of argon are calculated as functions of pressure. At 1000 atm. an internal kinetic energy of

Optimization of reactor pressure vessel internals segmentation in Korea

Energy Technology Data Exchange (ETDEWEB)

Lee, Byung-Sik [Dankook Univ., Chungnam (Korea, Republic of). Dept. of Nuclear Engineering

2017-11-15

One of the most challenging tasks during plant decommissioning is the removal of highly radioactive internal components from the reactor pressure vessel (RPV). For RPV internals dismantling, it is essential that all activities are thoroughly planned and discussed in the early stage of the decommissioning project. One of the key activities in the detailed planning is to prepare the segmentation and packaging plan that describes the sequential steps required to segment, separate, and package each individual component of RPV, based on an activation analysis and component characterization study.

International bulletin on atomic and molecular data for fusion. No. 54-55

International Nuclear Information System (INIS)

Stephens, J.A.

1998-12-01

This bulletin is published by the International Atomic Energy Agency to provide atomic and molecular data relevant to fusion research and technology. In the first part the indexed papers are listed separately for (i) structure and spectra (energy levels, wavelengths, transition probabilities, oscillator strengths, polarizabilities, electric moments, interatomic potentials), (ii) atomic and molecular collisions (photon collisions, electron collisions, heavy-particle collisions), and (iii) surface interactions (sputtering, chemical reactions, trapping and detrapping, adsorption, desorption, reflection, and secondary electron emission). There are also chapters with beam-matter interactions and data on interactions of atomic particles with fields. In the second Part contains the bibliographic data, essentially for the above listed topics

Statement to International Conference on Fast Reactors and Related Fuel Cycles: Challenges and Opportunities, 7 December 2009, Kyoto, Japan

International Nuclear Information System (INIS)

Amano, Yukiya

2009-01-01

Full text: Distinguished Guests, Ladies and Gentlemen, It is my honour to address participants at this opening session of the International Conference on Fast Reactors and Related Fuel Cycles: Challenges and Opportunities, organized by the IAEA and hosted by the Japan Atomic Energy Agency. Fast reactor technology has the potential to ensure that energy resources which would last hundreds of years with the technology we are using today will actually last several thousand years. In other words it can withstand enormous increases in demand. This innovative technology also reduces the risk to the environment and helps to limit the burden that will be placed on future generations in the form of waste products. The coming year will be an exciting one for the development of fast-spectrum nuclear reactors. We expect to reach many important milestones: - the first criticality of the China Experimental Fast Reactor; - the restart of the Monju prototype fast reactor in Japan; and - the new insights we will gain through the end-of-life studies at the Phenix reactor in France. In the near future, new fast reactors will be commissioned: the 500MW(e) Prototype Fast Breeder Reactor in India, the first in a series of five of the same type, and the BN-800 reactor in the Russian Federation. Moreover, France, Japan, India, China and the Republic of Korea are preparing advanced prototypes, demonstration or commercial reactors for the 2020-2030 period. Nuclear power is set to be an increasingly important part of the global energy mix in the coming decades as demand for energy grows. Scores of countries in both the developed and developing world have told the IAEA that they are interested in introducing nuclear power. The 30 countries which already have nuclear power reactors are set to build more. This trend is likely to be accompanied by accelerated deployment of fast reactors. Continued advances in research and technology development are necessary to ensure improved economics and

Basic principles of accounting and control of nuclear materials in the BOR-60 experimental fast reactor

International Nuclear Information System (INIS)

Gryazev, V.M.; Gadzhiev, G.I.; Alekseev, I.N.

1979-01-01

Under a contract with the International Atomic Energy Agency, the V.I. Lenin Atomic Reactor Research Institute is currently carrying out a study of ways of organizing a nuclear materials accounting and control system for the BOR-60 fast reactor. Some results of this study are presented in the paper. The special physical and technological features of fast reactors create additional difficulties in safeguards systems and give rise to a number of new possibilities for the illicit removal of nuclear materials. These questions are discussed with reference to the BOR-60 reactor but the conclusions are probably applicable to all fast reactors. The proposed accounting and control system is based on non-destructive measurements of the amount of fissile materials in the operating fuel assemblies and screened bundles of the reactor, on the independent control of the principal facility parameters (a list of which is given) and on an automated information collection and evaluation system. Visual means of inspection can be very effective in fast reactor safeguards systems, especially for controlling storage, but they are not used with the BOR-60 reactor. (author)

Status of computer codes available in AEOI for reactor physics analysis

International Nuclear Information System (INIS)

Karbassiafshar, M.

1986-01-01

Many of the nuclear computer codes available in Atomic Energy Organization of Iran AEOI can be used for physics analysis of an operating reactor or design purposes. Grasp of the various methods involved and practical experience with these codes would be the starting point for interesting design studies or analysis of operating conditions of presently existing and future reactors. A review of the objectives and flowchart of commonly practiced procedures in reactor physics analysis of LWRs and related computer codes was made, extrapolating to the nationally and internationally available resources. Finally, effective utilization of the existing facilities is discussed and called upon

International Atomic Energy Agency. Highlights of activities. September 1993

International Nuclear Information System (INIS)

Gillen, V.A.

1993-09-01

This document describes the most important activities of the International Atomic Energy Agency during the period September 1992 - September 1993, in particular in the following areas: (i) nuclear power; (ii) nuclear fuel cycle; (iii) radioactive waste management; (iv) comparative assessment of energy sources; (v) IAEA laboratory activities; (vi) nuclear applications in the food industry and in agriculture; (vii) human health applications of nuclear techniques, especially in the treatment and prevention of diseases and in the analysis of health problems related to the environment; (viii) industry and earth sciences; (ix) physical and chemical sciences; (x) radiation protection; (xi) safety of nuclear installations; (xii) safeguards and non-proliferation activities; (xiii) activities in the area of public and technical information such as the International Nuclear Information System (INIS) and other IAEA computerized databases and reference systems, the publication Nuclear Fusion, a monthly scientific journal of articles on thermonuclear fusion research and development, and the organization of meetings on atomic energy; and (xiv) a description of the Agency's technical assistance activities, including financial data

Atoms at work

International Nuclear Information System (INIS)

1982-07-01

This illustrated booklet discusses the following: atoms; fission of uranium; nuclear power plants; reactor types; plutonium (formation, properties, uses); radioactive waste (fuel cycle, reprocessing, waste management); nuclear fusion; fusion reactors; radiation; radioisotopes and their uses. (U.K.)

Post-remedial-action survey report for Kinetic Experiment Water Boiler Reactor Facility, Santa Susana Field Laboratories, Rockwell International, Ventura County, California

International Nuclear Information System (INIS)

Wynveen, R.A.; Smith, W.H.; Sholeen, C.M.; Flynn, K.F.; Justus, A.L.

1981-10-01

Rockwell International's Santa Susana Laboratories in Ventura County, California, have been the site of numerous federally-funded contracted projects involving the use of radioactive materials. Among these was the Kinetics Experiment Water Boiler (KEWB) Reactor which was operated under the auspices of the US Atomic Energy Commission (AEC). The KEWB Reactor was last operated in 1966. The facility was subsequently declared excess and decontamination and decommissioning operations were conducted during the first half of calendar year 1975. The facility was completely dismantled and the site graded to blend with the surrounding terrain. During October 1981, a post-remedial-action (certification) survey of the KEWB site was conducted on the behalf of the US Department of Energy by the Radiological Survey Group (RSG) of the Occupational Health and Safety Division's Health Physics Section (OHS/HP) of Argonne National Laboratory (ANL). The survey confirmed that the site was free from contamination and could be released for unrestricted use

Gas-cooled reactor safety and accident analysis

International Nuclear Information System (INIS)

1985-12-01

The Specialists' Meeting on Gas-Cooled Reactor Safety and Accident Analysis was convened by the International Atomic Energy Agency in Oak Ridge on the invitation of the Department of Energy in Washington, USA. The meeting was hosted by the Oak Ridge National Laboratory. The purpose of the meeting was to provide an opportunity to compare and discuss results of safety and accident analysis of gas-cooled reactors under development, construction or in operation, to review their lay-out, design, and their operational performance, and to identify areas in which additional research and development are needed. The meeting emphasized the high safety margins of gas-cooled reactors and gave particular attention to the inherent safety features of small reactor units. The meeting was subdivided into four technical sessions: Safety and Related Experience with Operating Gas-Cooled Reactors (4 papers); Risk and Safety Analysis (11 papers); Accident Analysis (9 papers); Miscellaneous Related Topics (5 papers). A separate abstract was prepared for each of these papers

Fuel performance at high burnup for water reactors

International Nuclear Information System (INIS)

1991-02-01

The present meeting was scheduled by the International Atomic Energy Agency, upon proposal of the Members of the International Working Group on Water Reactor Fuel Performance and Technology. The purpose of this meeting was to review the ''state-of-the-art'' in the area of Fuel Performance at High Burnup for Water Reactors. Previous IAEA meetings on this topic were held in Mol in 1981 and 1984 and on related topics in Stockholm and Lyon in 1987. Fifty-five participants from 16 countries and two international organizations attended the meeting and 28 papers were presented and discussed. The papers were presented in five sub-sessions and during the meeting, working groups composed of the session chairmen and paper authors prepared the summary of each session with conclusions and recommendations for future work. A separate abstract was prepared for each of these papers. Refs, figs and tabs

Bismuth as a general internal standard for lead in atomic absorption spectrometry

International Nuclear Information System (INIS)

Bechlin, Marcos A.; Fortunato, Felipe M.; Ferreira, Edilene C.; Neto, José A. Gomes; Nóbrega, Joaquim A.; Donati, George L.; Jones, Bradley T.

2014-01-01

Highlights: • Single internal standard is commonly proposed for definite application in AAS. • Internal standard for general use in AAS techniques is original. • Bi showed efficiency as internal standard for Pb determinations by FAAS and GFAAS. • Assorted samples were analyzed and accurate results were found. - Abstract: Bismuth was evaluated as internal standard for Pb determination by line source flame atomic absorption spectrometry (LS FAAS), high-resolution continuum source flame atomic absorption spectrometry (HR-CS FAAS) and line source graphite furnace atomic absorption spectrometry (LS GFAAS). Analysis of samples containing different matrices indicated close relationship between Pb and Bi absorbances. Correlation coefficients of calibration curves built up by plotting A Pb /A Bi versus Pb concentration were higher than 0.9953 (FAAS) and higher than 0.9993 (GFAAS). Recoveries of Pb improved from 52–118% (without IS) to 97–109% (IS, LS FAAS); 74–231% (without IS) to 96–109% (IS, HR-CS FAAS); and 36–125% (without IS) to 96–110% (IS, LS GFAAS). The relative standard deviations (n = 12) were reduced from 0.6–9.2% (without IS) to 0.3–4.3% (IS, LS FAAS); 0.7–7.7% (without IS) to 0.1–4.0% (IS, HR-CS FAAS); and 2.1–13% (without IS) to 0.4–5.9% (IS, LS GFAAS)

Study on dynamic characteristics of reduced analytical model for PWR reactor internal structures

International Nuclear Information System (INIS)

Yoo, Bong; Lee, Jae Han; Kim, Jong Bum; Koo, Kyeong Hoe

1993-01-01

The objective of this study is to establish the procedure of the reduced analytical modeling technique for the PWR reactor internal(RI) structures and to carry out the sensitivity study of the dynamic characteristics of the structures by varying the structural parameters such as the stiffness, the mass and the damping. Modeling techniques for the PWR reactor internal structures and computer programs used for the dynamic analysis of the reactor internal structures are briefly investigated. Among the many components of RI structures, the dynamic characteristics for CSB was performed. The sensitivity analysis of the dynamic characteristics for the reduced analytical model considering the variations of the stiffnesses for the lower and upper flanges of the CSB and for the RV Snubber were performed to improve the dynamic characteristics of the RI structures against the external loadings given. In order to enhance the structural design margin of the RI components, the nonlinear time history analyses were attempted for the RI reduced models to compare the structural responses between the reference model and the modified one. (Author)

Atomic vapor laser isotope separation

International Nuclear Information System (INIS)

Stern, R.C.; Paisner, J.A.

1985-01-01

Atomic vapor laser isotope separation (AVLIS) is a general and powerful technique. A major present application to the enrichment of uranium for light-water power reactor fuel has been under development for over 10 years. In June 1985 the Department of Energy announced the selection of AVLIS as the technology to meet the nation's future need for the internationally competitive production of uranium separative work. The economic basis for this decision is considered, with an indicated of the constraints placed on the process figures of merit and the process laser system. We then trace an atom through a generic AVLIS separator and give examples of the physical steps encountered, the models used to describe the process physics, the fundamental parameters involved, and the role of diagnostic laser measurements

Westinghouse experience in using mechanical cutting for reactor vessel internals segmentation

International Nuclear Information System (INIS)

Boucau, Joseph; Fallstroem, Stefan; Segerud, Per; Kreitman, Paul J.

2010-01-01

Some commercial nuclear power plants have been permanently shut down to date and decommissioned using dismantling methods. Other operating plants have decided to undergo an upgrade process that includes replacement of reactor internals. In both cases, there is a need to perform a segmentation of the reactor vessel internals with proven methods for long term waste disposal. Westinghouse has developed several concepts to dismantle reactor internals based on safe and reliable techniques. Mechanical cutting has been used by Westinghouse since 1999 for both PWRs and BWRs and its process has been continuously improved over the years. Detailed planning is essential to a successful project, and typically a 'Segmentation and Packaging Plan' is prepared to document the effort. The usual method is to start at the end of the process, by evaluating the waste disposal requirements imposed by the waste disposal agency, what type and size of containers are available for the different disposal options, and working backwards to select the best cutting tools and finally the cut geometry required. These plans are made utilizing advanced 3-D CAD software to model the process. Another area where the modelling has proven invaluable is in determining the logistics of component placement and movement in the reactor cavity, which is typically very congested when all the internals are out of the reactor vessel in various stages of segmentation. The main objective of the segmentation and packaging plan is to determine the strategy for separating the highly activated components from the less activated material, so that they can be disposed of in the most cost effective manner. Usually, highly activated components cannot be shipped off-site, so they must be packaged such that they can be dry stored with the spent fuel in an Independent Spent Fuel Storage Installation (ISFSI). Less activated components can be shipped to an off-site disposal site depending on space availability. Several of the

Advanced nuclear reactor safety issues and research needs

International Nuclear Information System (INIS)

2002-01-01

On 18-20 February 2002, the OECD Nuclear Energy Agency (NEA) organised, with the co-sponsorship of the International Atomic Energy Agency (IAEA) and in collaboration with the European Commission (EC), a Workshop on Advanced Nuclear Reactor Safety Issues and Research Needs. Currently, advanced nuclear reactor projects range from the development of evolutionary and advanced light water reactor (LWR) designs to initial work to develop even further advanced designs which go beyond LWR technology (e.g. high-temperature gas-cooled reactors and liquid metal-cooled reactors). These advanced designs include a greater use of advanced technology and safety features than those employed in currently operating plants or approved designs. The objectives of the workshop were to: - facilitate early identification and resolution of safety issues by developing a consensus among participating countries on the identification of safety issues, the scope of research needed to address these issues and a potential approach to their resolution; - promote the preservation of knowledge and expertise on advanced reactor technology; - provide input to the Generation IV International Forum Technology Road-map. In addition, the workshop tried to link advancement of knowledge and understanding of advanced designs to the regulatory process, with emphasis on building public confidence. It also helped to document current views on advanced reactor safety and technology, thereby contributing to preserving knowledge and expertise before it is lost. (author)

Development of telerobotic manipulators for reactor dismantling work

International Nuclear Information System (INIS)

Shinohara, Yoshikuni; Usui, Hozumi; Fujii, Yoshio

1991-01-01

This paper describes the amphibious electrical manipulators JARM-10, JART-25, JART-100 and JARM-25 which were developed in the program of reactor decommissioning technology development carried out by the Japan Atomic Energy Research Institute. They are multi-functional telerobotic light-duty (10 and 25 daN) and heavy-duty (100 daN) Manipulators which can be used in hostile environments in reactor dismantling work such as high radiation, underwater work and electrical noise. Each manipulator can be operated in either a bilateral master-slave, a teach-and-playback or a programmed control mode. By combining these modes appropriately, it is possible to perform complex tasks of remote handling. The usefulness of the telerobotic systems for dismantling nuclear reactors has been demonstrated by successful application of the JARM-25 for remote underwater dismantlement of highly radioactive reactor internals of complex form of an experimental nuclear power reactor. (author)

IGORR 2: Proceedings of the 2. meeting of the International Group On Research Reactors

Energy Technology Data Exchange (ETDEWEB)

NONE

1992-07-01

The International group on Research Reactors was formed to facilitate the sharing of knowledge and experience among those institutions and individuals who are actively working to design, build, and promote new research reactors or to make significant upgrades to existing facilities. Sessions during this second meeting were devoted to research reactor reports (GRENOBLE reactor, FRM-II, HIFAR, PIK, reactors at JAERI, MAPLE, ANS, NIST, MURR, TRIGA, BR-2, SIRIUS 2); other neutron sources; and two workshops were dealing with research and development results and needs and reports on progress in needed of R and D areas identified at IGORR 1.

IGORR 2: Proceedings of the 2. meeting of the International Group On Research Reactors

International Nuclear Information System (INIS)

1992-01-01

The International group on Research Reactors was formed to facilitate the sharing of knowledge and experience among those institutions and individuals who are actively working to design, build, and promote new research reactors or to make significant upgrades to existing facilities. Sessions during this second meeting were devoted to research reactor reports (GRENOBLE reactor, FRM-II, HIFAR, PIK, reactors at JAERI, MAPLE, ANS, NIST, MURR, TRIGA, BR-2, SIRIUS 2); other neutron sources; and two workshops were dealing with research and development results and needs and reports on progress in needed of R and D areas identified at IGORR 1

Inspection and repair of reactor pressure vessel (RPV) internals

International Nuclear Information System (INIS)

Bohmann, W.; Poetz, F.; Nicolai, M.

1996-01-01

The past 10 years of operation of light water reactors were characterized by intensive inspection- and repair work on vital components. For boiling water reactors (BWR) it was typical to totally replace the piping system and for pressurized water reactors (PWR) it was the step to complete steam generator (SG) replacement - besides the development of increasingly diligent inspection and repair methods for SG tubes. It can be expected that in the 10 years to come the development of inspection- and repair methods will be aimed mainly at the core internals of BWR's as well as PWR's. Our prediction is that before the end of this decade a first complete replacement of these components will be performed. Already to date a broad range of techniques are available which enable the utilities to carry out inspections and repair of components of core internals in a relatively short time and acceptable expenses. Using examples such as Fuel Alignment Pin Inspection and Replacement, Baffle Former Bolt Inspection and Replacement, Core Barrel Former Bolt Inspection which are typical for PWR's we will in the following describe the existing methods, their development and - last but not least - their successful utilization. What is going to happen in the future? Ageing of the operating plants will continue, thus requesting the plant operators as well as the service companies to work on advanced technologies to fulfill the needs of the industry. (author)

Report to the International Atomic Energy Agency

International Nuclear Information System (INIS)

Woznicka, U.

1986-07-01

The results of research done under the International Atomic Energy Agency Contract no 4121/RB: 'Measurement of the thermal neutron macroscopic absorption cross-section on small samples' at the Institut of Nuclear Physics during the period from July 1, 1985 to June 30, 1986 are presented. The research was based on the Plexiglass thermal neutron diffusion parameters and on the method of preparation of the rock samples for the measurements according to the INP method. Three rock samples delivered by the IAEA: Ottawa Sand, Royer Dolomite and Dunite Sand have been measured. (author)

A sustainability analysis of the Brazilian multipurpose reactor project

International Nuclear Information System (INIS)

Obadia, I.J.; Perrotta, J.A.

2010-01-01

The project of a new research reactor in Brazil for radioisotope production, support of the nuclear energy program and scientific research has received a positive sign of the government and is starting to be developed by the Brazilian Commission of Nuclear Energy. International Atomic Energy Agency points out that the implementation of a new research reactor is a major undertaking for a country, requiring an analysis to identify to which extent the conditions of the national nuclear program are proper and adequate to lead to a sustainable research reactor life cycle. This paper introduces the Brazilian Multipurpose Reactor Project (RMB) and describes the sustainability analysis performed, which has shown that the national nuclear infrastructure presents a very favourable condition to the implementation of the RMB project as well as to provide a sustainable life cycle for this new research reactor. (author)

IAEA activities in nuclear reactor simulation for educational purposes

International Nuclear Information System (INIS)

Lyon, R.B.

2001-01-01

The International Atomic Energy Agency (IAEA) has established a programme in nuclear reactor simulation computer programs to assist its Member States in education and training. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the supply or development of simulation programs and training material, sponsors training courses and workshops, and distributes documentation and computer programs. Two simulation programs are presented at this workshop: the Classroom-based Advanced Reactor Demonstrators package, and the Advanced Reactor Simulator. Both packages simulate the behaviour of BWR, PWR and HWR reactor types. For each package, the modeling approach and assumptions are broadly described, together with a general description of the operation of the computer programs. (author)

Reactor Core Internals Replacement of Ikata Units 1 and 2

International Nuclear Information System (INIS)

Ikeda, K.; Ishikawa, T.; Miyoshi, T.; Takagi, T.

2012-01-01

This paper presents an overview of the reactor core internals replacement project carried out at the Ikata Nuclear Power Station in Japan, which was the first of its kind among PWRs in the world. Failure of baffle former bolts was first reported in 1989 at Bugey 2 in France. Since then, similar incidents have been reported in Belgium and in the U.S., but not in Japan. However, the possibility of these bolts failing in Japanese plants cannot be denied in the future as operating hours increase. Ageing degradation mechanisms for the reactor core internals include irradiation-assisted stress corrosion cracking of baffle former bolts and mechanical wear of control rod guide cards. Two different approaches can be taken to address these ageing issues: to inspect and repair whenever a problem is found; and to replace the entire core internals with those of a new design having advanced features to prevent ageing degradation problems. The choice of our company was the latter. This paper explains the reasons for the choice and summarizes the replacement project activities at Ikata Units 1 and 2 as well as the improvements incorporated in the new design. (author)

International bulletin on atomic and molecular data for fusion. Nos. 50-51

International Nuclear Information System (INIS)

Botero, J.; Stephens, J.A.

1996-10-01

This bulletin is published by the International Atomic Energy Agency to provide atomic and molecular data relevant to fusion research and technology. In part 1 the indexed papers are listed separately for (i) structure and spectra (energy levels, wavelengths, transition probabilities, oscillator strengths, polarizabilities, electric moments, interatomic potentials); (ii) atomic and molecular collisions (photon collisions, electron collisions, heavy-particle collisions); and (iii) surface interactions (sputtering, chemical reactions, trapping and detrapping, adsorption, desorption, reflection, and secondary electron emission). Part 2 contains the bibliographic data, essentially for the above listed topics

WWER-1000 reactor simulator. Workshop material

International Nuclear Information System (INIS)

2003-01-01

The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and educational material and sponsors courses and workshops. The workshops are in two parts: techniques and tools for reactor simulator development; and the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA publication: Training Course Series 12, 'Reactor Simulator Development' (2001). Course material for workshops using a pressurized water reactor (PWR) Simulator developed for the IAEA by Cassiopeia Technologies Inc. of Canada is presented in the IAEA publication: Training Course Series No. 22 'Pressurized Water Reactor Simulator' (2003) and Training Course Series No. 23 'Boiling Water Reactor Simulator' (2003). This report consists of course material for workshops using the WWER-1000 Reactor Department Simulator from the Moscow Engineering and Physics Institute, Russian Federation. N. V. Tikhonov and S. B. Vygovsky of the Moscow Engineering and Physics Institute prepared this report for the IAEA

Boiling water reactor simulator. Workshop material

International Nuclear Information System (INIS)

2003-01-01

The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and workshop material and sponsors workshops. The workshops are in two parts: techniques and tools for reactor simulator development; and the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA publication: Training Course Series No. 12, 'Reactor Simulator Development' (2001). Course material for workshops using a WWER- 1000 simulator from the Moscow Engineering and Physics Institute, Russian Federation is presented in the IAEA publication: Training Course Series No. 21 'WWER-1000 Reactor Simulator' (2002). Course material for workshops using a pressurized water reactor (PWR) simulator developed by Cassiopeia Technologies Incorporated, Canada, is presented in the IAEA publication: Training Course Series No. 22 'Pressurized Water Reactor Simulator' (2003). This report consists of course material for workshops using a boiling water reactor (BWR) simulator. Cassiopeia Technologies Incorporated, developed the simulator and prepared this report for the IAEA

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

International Nuclear Information System (INIS)

Iracane, Daniel; Bignan, Gilles; Lindbaeck, Jan-Erik; Blomgren, Jan

2010-01-01

The development of sustainable nuclear energy requires R and D on fuel and material behaviour under irradiation with a high level of performance in order to meet the needs and challenges for the benefit of industry, research and public bodies. These stakes require a sustainable and secured access to an up-to-date high performance Material Testing Reactor. Following a broad survey within the European Research Area, the international community agreed that the need for Material Test Reactors in support of nuclear power plant safety and operation will continue in the context of sustainable nuclear energy. The Jules Horowitz Reactor project (JHR) copes with this context. JHR is designed as a user facility addressing the needs of the international community. This means: - flexibility with irradiation loops able to reproduce a large variation in operation conditions of different power reactor technologies, - high flux capacity to address Generations II, III, and IV needs. JHR is designed, built and operated as an international user facility because: - Given the maturity and globalization of the industry, domestic tools have no more the required level of economic and technical efficiency. Meanwhile, countries with nuclear energy need an access to high performance irradiation experimental capabilities to support technical skill and guarantee the competitiveness and safety of nuclear energy. - Many research items related to safety or public policy (waste management, etc.) require international cooperation to share costs and benefits of resulting consensus. JHR design is optimised for offering high performance material and fuel irradiation capability for the coming decades. This project is driven and funded by an international consortium gathering vendors, utilities and public stakeholders. This consortium has been set up in March 2007 when the construction began. The construction is in progress and the start of operation is scheduled for 2014. The JHR is a research

International bulletin on atomic and molecular data for fusion. No. 11

International Nuclear Information System (INIS)

Katsonis, K.; Rumble, J. Jr.

1980-01-01

This bulletin deals with atomic and molecular data for fusion. A bibliography for the most recent data presented in the document is provided. Work in progress is briefly reported. The bulletin contains a list of references the publications on controlled fusion and plasma physics for 1979. It contains an index to the contributed papers presented at the 11th International Conference on the Physics of Electronics and Atomic Collision (ICPEAC) held in Kyoto (Japan) in summer 1979

HEU core conversion of Russian production reactors: a major threat to the international RERTR regime

International Nuclear Information System (INIS)

Kuperman, Alan J.; Leventhal, Paul L.

1998-01-01

This paper calls the attention for the major threat to the International Reduced Enrichment for Research and Test Reactors (RERTR) program, represented by the HEU core conversion of russian production reactors. This program aims to reduce and eventually eliminate international civilian commerce in nuclear weapons-usable, highly enriched uranium , and thereby significantly lower risks of the material being stolen or diverted by terrorist or states for producing nuclear weapons

International Reactor Physics Experiment Evaluation (IRPhE) Project. IRPhE Handbook - 2017 edition

International Nuclear Information System (INIS)

2017-01-01

The International Reactor Physics Evaluation (IRPhE) Project was initiated as a pilot in 1999 by the Nuclear Energy Agency (NEA) Nuclear Science Committee (NSC). The project was endorsed as an official activity of the NSC in June 2003. While the NEA co-ordinates and administers the IRPhE Project at the international level, each participating country is responsible for the administration, technical direction and priorities of the project within their respective countries. The information and data included in this handbook are available to NEA member countries, to all contributing countries and to others on a case-by-case basis. The IRPhE Project is patterned after the International Criticality Safety Benchmark Evaluation Project (ICSBEP). It closely co-ordinates with the ICSBEP to avoid duplication of efforts and publication of conflicting information. Some benchmark data are applicable to both nuclear criticality safety and reactor physics technology. Some have already been evaluated and published by the ICSBEP, but have been extended to include other types of measurements in addition to the critical configuration. Through this effort, the IRPhE Project will be able to 1) consolidate and preserve the existing worldwide information base; 2) retrieve lost data; 3) identify areas where more data are needed; 4) draw upon the resources of the international reactor physics community to help fill knowledge gaps; 5) identify discrepancies between calculations and experiments due to deficiencies in reported experimental data, cross-section data, cross-section processing codes and neutronics codes; 6) eliminate a large amount of redundant research and processing of reactor physics experiment data, and 7) improve future experimental planning, execution and reporting. This handbook contains reactor physics benchmark specifications that have been derived from experiments performed at nuclear facilities around the world. The benchmark specifications are intended for use by

Environment, 1986-1997. International Atomic Energy Agency publications

International Nuclear Information System (INIS)

1998-04-01

This catalogue lists all sales publications of the International Atomic Energy Agency dealing with Environment and issued during the period of 1986-1997. Some earlier titles which form part of an established series or are still considered of importance have been included. Most publications are in English. Proceedings of conferences, symposia and panels of experts may contain papers in languages other than English, but all of these papers have abstracts in English

Development of a research nuclear reactor simulator using LABVIEW®

Energy Technology Data Exchange (ETDEWEB)

Lage, Aldo Marcio Fonseca; Mesquita, Amir Zacarias; Pinto, Antonio Juscelino; Souza, Luiz Claudio Andrade [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

2015-07-01

The International Atomic Energy Agency recommends the use of safety and friendly interfaces for monitoring and controlling the operational parameters of the nuclear reactors. The most important variable in the nuclear reactors control is the power released by fission of the fuel in the core which is directly proportional to neutron flux. It was developed a digital system to simulate the neutron evolution flux and monitoring their interaction on the other operational parameters. The control objective is to bring the reactor power from its source level (mW) to a few W. It is intended for education of basic reactor neutronic principles such as the multiplication factor, criticality, reactivity, period, delayed neutron and control by rods. The 250 kW IPR-R1 TRIGA research reactor at Nuclear Technology Development Center - CDTN (Belo Horizonte/Brazil) was used as reference. TRIGA reactors, developed by General Atomics (GA), are the most widely used research reactor in the world. They are cooled by light water under natural convection and are characterized by being inherently safety. The simulation system was developed using the LabVIEW® (Laboratory Virtual Instruments Engineering Workbench) software, considering the modern concept of virtual instruments (VI's). The main purpose of the system is to provide to analyze the behavior, and the tendency of some processes that occur in the reactor using a user-friendly operator interface. The TRIGA simulator system will allow the study of parameters, which affect the reactor operation, without the necessity of using the facility.(author)

Development of a research nuclear reactor simulator using LABVIEW®

International Nuclear Information System (INIS)

Lage, Aldo Marcio Fonseca; Mesquita, Amir Zacarias; Pinto, Antonio Juscelino; Souza, Luiz Claudio Andrade

2015-01-01

The International Atomic Energy Agency recommends the use of safety and friendly interfaces for monitoring and controlling the operational parameters of the nuclear reactors. The most important variable in the nuclear reactors control is the power released by fission of the fuel in the core which is directly proportional to neutron flux. It was developed a digital system to simulate the neutron evolution flux and monitoring their interaction on the other operational parameters. The control objective is to bring the reactor power from its source level (mW) to a few W. It is intended for education of basic reactor neutronic principles such as the multiplication factor, criticality, reactivity, period, delayed neutron and control by rods. The 250 kW IPR-R1 TRIGA research reactor at Nuclear Technology Development Center - CDTN (Belo Horizonte/Brazil) was used as reference. TRIGA reactors, developed by General Atomics (GA), are the most widely used research reactor in the world. They are cooled by light water under natural convection and are characterized by being inherently safety. The simulation system was developed using the LabVIEW® (Laboratory Virtual Instruments Engineering Workbench) software, considering the modern concept of virtual instruments (VI's). The main purpose of the system is to provide to analyze the behavior, and the tendency of some processes that occur in the reactor using a user-friendly operator interface. The TRIGA simulator system will allow the study of parameters, which affect the reactor operation, without the necessity of using the facility.(author)

International human cooperation in Japan Atomic Energy Research Institute

International Nuclear Information System (INIS)

Shiba, Koreyuki; Kaieda, Keisuke; Makuuchi, Keizo; Takada, Kazuo; Nomura, Masayuki

1997-01-01

Rearing of talented persons in the area of nuclear energy is one of the important works in Japan Atomic Energy Research Institute. In this report, the present situations and future schedules of international human cooperation in this area wsere summarized. First, the recent activities of International Nuclear Technology Center were outlined in respect of international human cooperation. A study and training course which was started in cooperation with JICA and IAEA from the middle of eighties and the international nuclear safety seminar aiming at advancing the nuclear safety level of the world are now being put into practice. In addition, a study and training for rearing talented persons was started from 1996 to improve the nuclear safety level of the neighbouring countries. The activities of the nuclear research interchange system by Science and Technology Agency established in 1985 and Bilateral Co-operation Agreement from 1984 were explained and also various difficulties in the international cooperation were pointed out. (M.N.)

Pressure vessel codes: Their application to nuclear reactor systems

International Nuclear Information System (INIS)

1966-01-01

A survey has been made by the International Atomic Energy Agency of how the problems of applying national pressure vessel codes to nuclear reactor systems have been treated in those Member States that had pressurized reactors in operation or under construction at the beginning of 1963. Fifteen answers received to an official inquiry form the basis of this report, which also takes into account some recently published material. Although the answers to the inquiry in some cases data back to 1963 and also reflect the difficulty of describing local situations in answer to standard questions, it is hoped that the report will be of interest to reactor engineers. 21 refs, 1 fig., 2 tabs

Nuclear measurements, techniques and instrumentation industrial applications plasma physics and nuclear fusion. 1980-1994. International Atomic Energy Agency publications

International Nuclear Information System (INIS)

1995-04-01

This catalogue lists all sales publications of the International Atomic Energy Agency dealing with Nuclear Measurements, Techniques and Instrumentation, with Industrial Applications (of Nuclear Physics and Engineering), and with Plasma Physics and Nuclear Fusion, issued during the period 1980-1994. Most publications are in English. Proceedings of conferences, symposia, and panels of experts may contain some papers in other languages (French, Russian, or Spanish), but all papers have abstracts in English. Price quotes are in Austrian Schillings, do not include local taxes, and are subject to change without notice. Contents cover the three main categories of (i) Nuclear Measurements, Techniques and Instrumentation (Physics, Chemistry, Dosimetry Techniques, Nuclear Analytical Techniques, Research Reactors and Particle Accelerator Applications, Nuclear Data); (ii) Industrial Applications (Radiation Processing, Radiometry, Tracers); and (iii) Plasma Physics and Nuclear Fusion

Nuclear measurements, techniques and instrumentation industrial applications plasma physics and nuclear fusion, 1980-1993. International Atomic Energy Agency publications

International Nuclear Information System (INIS)

1994-01-01

This catalogue lists all sales publications of the International Atomic Energy Agency dealing with Nuclear Measurements, Techniques and Instrumentation, with Industrial Applications (of Nuclear Physics and Engineering), and with Plasma Physics and Nuclear Fusion, issued during the period 1980-1993. Most publications are in English. Proceedings of conferences, symposia, and panels of experts may contain some papers in other languages (French, Russian, or Spanish), but all papers have abstracts in English. Price quotes are in Austrian Schillings, do not include local taxes, and are subject to change without notice. Contents cover the three main categories of (I) Nuclear Measurements, Techniques and Instrumentation (Physics, Chemistry, Dosimetry Techniques, Nuclear Analytical Techniques, Research Reactors and Particle Accelerator Applications, Nuclear Data); (ii) Industrial Applications (Radiation Processing, Radiometry, Tracers); and (iii) Plasma Physics and Nuclear Fusion

Evaluation and optimization of General Atomics' GT-MHR reactor cavity cooling system using an axiomatic design approach

International Nuclear Information System (INIS)

Thielman, Jeff; Ge, Ping; Wu, Qiao; Parme, Laurence

2005-01-01

The development of the Generation IV (Gen-IV) nuclear reactors has presented social, technical, and economical challenges to nuclear engineering design and research. To develop a robust, reliable nuclear reactor system with minimal environmental impact and cost, modularity has been gradually accepted as a key concept in designing high-quality nuclear reactor systems. While the establishment and reliability of a nuclear power plant is largely facilitated by the installment of standardized base units, the realization of modularity at the sub-system/sub-unit level in a base unit is still highly heuristic, and lacks consistent, quantifiable measures. In this work, an axiomatic design approach is developed to evaluate and optimize the reactor cavity cooling system (RCCS) of General Atomics' Gas Turbine-Modular Helium Reactor (GT-MHR) nuclear reactor, for the purpose of constructing a quantitative tool that is applicable to Gen-IV systems. According to Suh's axiomatic design theory, modularity is consistently represented by functional independence through the design process. Both qualitative and quantitative measures are developed here to evaluate the modularity of the current RCCS design. Optimization techniques are also used to improve the modularity at both conceptual and parametric level. The preliminary results of this study have demonstrated that the axiomatic design approach has great potential in enhancing modular design, and generating more robust, safer, and less expensive nuclear reactor sub-units

Improved spin squeezing of an atomic ensemble through internal state control

Science.gov (United States)

Hemmer, Daniel; Montano, Enrique; Deutsch, Ivan; Jessen, Poul

2016-05-01

Squeezing of collective atomic spins is typically generated by quantum backaction from a QND measurement of the relevant spin component. In this scenario the degree of squeezing is determined by the measurement resolution relative to the quantum projection noise (QPN) of a spin coherent state (SCS). Greater squeezing can be achieved through optimization of the 3D geometry of probe and atom cloud, or by placing the atoms in an optical cavity. We explore here a complementary strategy that relies on quantum control of the large internal spin available in alkali atoms such as Cs. Using a combination of rf and uw magnetic fields, we coherently map the internal spins in our ensemble from the SCS (| f = 4, m = 4>) to a ``cat'' state which is an equal superposition of | f = 4, m = 4>and | f = 4, m = -4>. This increases QPN by a factor of 2 f = 8 relative to the SCS, and therefore the amount of backaction and spin-spin entanglement produced by our QND measurement. In a final step, squeezing generated in the cat state basis can be mapped back to the SCS basis, where it corresponds to increased squeezing of the physical spin. Our experiments suggest that up to 8dB of metrologically useful squeezing can be generated in this way, compared to ~ 3 dB in an otherwise identical experiment starting from a SCS.

Keeping the peaceful atom from raising the risk of war. Third of four articles

International Nuclear Information System (INIS)

Faltermayer, E.

1979-01-01

Many opponents of nuclear power contend that by the end of the century worldwide expansion of peaceful atomic energy will lead to the spread of nuclear weaponry among small nations and terrorists. Champions of fission power, though, say terrorists would have a tough time fashioning a bomb from commercial nuclear fuel, but that national governments might use civil atomic power as a shortcut to atomic weapons. It is not easy to divert fuel under the present system of international surveillance, but some proposed new safeguards could make it much harder. The world has about ten years in which to adopt such safeguards. Countries without reactors have one of two technical hurdles to surmount to convert commercial reactor fuel to bombs--converting the uranium in new, unburned reactor fuel to bomb material and extracting bomb material from spent commercial fuel. President Carter in deferring indefinitely the reprocessing of spent nuclear fuel hoped to set an example that other countries would follow. It has not deterred France, Britain, and West Germany from proceeding with construction of their own reprocessing plants, nor lessened their enthusiasm for breeder reactors. Terrorists compared to governments are considered to be a smaller threat of potential diverters of nuclear fuel for bombs. Many experts are convinced that new political arrangements must be supplemented with technical fixes and the most interesting proposals involve the denaturing or spiking of commercial reactor fuel. Proposals made by Princeton's Program on Nuclear Policy Alternatives, EPRI, Britain's Atomic Energy Authority, and Allied Chemical are discussed briefly

Participation of Czechoslovakia in the beginnings of international cooperation on peaceful use of atomic energy (IAEA, UNSCEAR, ICRP)

International Nuclear Information System (INIS)

Tesinska, E.

2006-01-01

The former Czechoslovakia, as a producer or natural uranium and a country heading to construction of nuclear power stations, took an active part in the establishment of international cooperation in peaceful use of atomic energy in the mid-1950s. Its 17-member delegation attended the 1 st International Conference on Peaceful Use of Atomic Energy in Geneva in August 1955. Czechoslovakia was invited to work of the twelve countries on the final draft of the International Atomic Energy Agency Statutes in 1956, it became a member of the IAEA Board of Governors (alternating with Poland), and the first Czechoslovak Governor P. Winkler was elected President of the Board in 1957. Czechoslovakia also was one of the fifteen countries designated to the UN Scientific Committee on the Effects of Atomic Radiation in 1955. Besides, Czechoslovak specialists participated actively in the work of committees of the International Commission on Radiological Protection for many years. Involvement in the international cooperation on peaceful use of atomic energy represented for Czechoslovakia obligations as well as stimuli to the national development of nuclear fields, including radiobiology. (author)

Application of Reactor Antineutrinos: Neutrinos for Peace

Science.gov (United States)

Suekane, F.

2013-02-01

In nuclear reactors, 239Pu are produced along with burn-up of nuclear fuel. 239Pu is subject of safeguard controls since it is an explosive component of nuclear weapon. International Atomic Energy Agency (IAEA) is watching undeclared operation of reactors to prevent illegal production and removal of 239Pu. In operating reactors, a huge numbers of anti electron neutrinos (ν) are produced. Neutrino flux is approximately proportional to the operating power of reactor in short term and long term decrease of the neutrino flux per thermal power is proportional to the amount of 239Pu produced. Thus rector ν's carry direct and real time information useful for the safeguard purposes. Since ν can not be hidden, it could be an ideal medium to monitor the reactor operation. IAEA seeks for novel technologies which enhance their ability and reactor neutrino monitoring is listed as one of such candidates. Currently neutrino physicists are performing R&D of small reactor neutrino detectors to use specifically for the safeguard use in response to the IAEA interest. In this proceedings of the neutrino2012 conference, possibilities of such reactor neutrinos application and current world-wide R&D status are described.

Multiple microprocessor based nuclear reactor power monitor

International Nuclear Information System (INIS)

Lewis, P.S.; Ethridge, C.D.

1979-01-01

The reactor power monitor is a portable multiple-microprocessor controlled data acquisition device being built for the International Atomic Energy Association. Its function is to measure and record the hourly integrated operating thermal power level of a nuclear reactor for the purpose of detecting unannounced plutonium production. The monitor consists of a 3 He proportional neutron detector, a write-only cassette tape drive and control electronics based on two INTEL 8748 microprocessors. The reactor power monitor operates from house power supplied by the plant operator, but has eight hours of battery backup to cover power interruptions. Both the hourly power levels and any line power interruptions are recorded on tape and in memory. Intermediate dumps from the memory to a data terminal or strip chart recorder can be performed without interrupting data collection

Decay heat removal and heat transfer under normal and accident conditions in gas cooled reactors

International Nuclear Information System (INIS)

1994-08-01

The meeting was convened by the International Atomic Energy Agency on the recommendation of the IAEA's International Working Group on Gas Cooled Reactors. It was attended by participants from China, France, Germany, Japan, Poland, the Russian Federation, Switzerland, the United Kingdom and the United States of America. The meeting was chaired by Prof. Dr. K. Kugeler and Prof. Dr. E. Hicken, Directors of the Institute for Safety Research Technology of the KFA Research Center, and covered the following: Design and licensing requirements for gas cooled reactors; concepts for decay heat removal in modern gas cooled reactors; analytical methods for predictions of thermal response, accuracy of predictions; experimental data for validation of predictive methods - operational experience from gas cooled reactors and experimental data from test facilities. Refs, figs and tabs

The obscure factor analysis on the vibration reliability of the internals of nuclear power plant reactor and anti-vibration measures

International Nuclear Information System (INIS)

Fu Geyan; Zhu Qirong

1998-11-01

It is pointed out that the main reason making nuclear power plants reactors leak is the vibration of internals of reactors. The factors which lead the vibration all have randomness and obscureness. The obscure reliability theory is introduced to the vibration system of internals of nuclear power reactor. Based on a quantity of designing and moving data, the obscure factors effecting the vibration reliability of the internals of nuclear power plant reactor are analyzed and the anti-vibration reliability criteria and the evaluating model are given. And the anti-vibration reliability measures are advanced from different quarters of the machine design and building, the thermohydraulics design, the control of reactivity, etc.. They may benefit the theory and practice for building and perfecting the vibration obscure reliability model of the reactor internals

Rough estimate demand of atomic energy-related budget for fiscal year 1996

International Nuclear Information System (INIS)

Kitagishi, Tatsuro

1996-01-01

The rough estimate demand of the budget for fiscal year 1996 of eight atomic energy-related ministries and agencies was determined at about 494,879 million yen, which is 2.4% growth as compared with that for the previous year. Concretely, the general account is 204,594 million yen, 2.2% growth, and the special account is 290,285 million yen, 2.6% growth. The budget is 357,060 million yen and 3.7% growth for Science and Technology Agency, 130, 787 million yen and 2% decrease for Ministry of International Trade and Industry, and 7,032 million yen and 29.2% increase for other six ministries and agencies. Emphasis is placed on the research of upgrading LWRs including the disassembling of reactors, the performance test for fuel, the improvement of reactor technology and the verifying test of practical reactor decommissioning facilities, and the research and development of advanced nuclear fuel cycle technology. Also the technical development of waste treatment and disposal including high level radioactive waste is carried out with 40.3 billion yen. Atomic Energy Commission exerts efforts for the development of atomic energy policy for the peaceful utilization, the establishment of coordinative LWR power generation system, the development of nuclear fuel recycling and the strengthening of the basic research on atomic energy. (K.I.)

Research Update: Atmospheric pressure spatial atomic layer deposition of ZnO thin films: Reactors, doping, and devices

Energy Technology Data Exchange (ETDEWEB)

Hoye, Robert L. Z., E-mail: rlzh2@cam.ac.uk, E-mail: jld35@cam.ac.uk; MacManus-Driscoll, Judith L., E-mail: rlzh2@cam.ac.uk, E-mail: jld35@cam.ac.uk [Department of Materials Science and Metallurgy, University of Cambridge, 27 Charles Babbage Road, Cambridge CB3 0FS (United Kingdom); Muñoz-Rojas, David [LMGP, University Grenoble-Alpes, CNRS, F-3800 Grenoble (France); Nelson, Shelby F. [Kodak Research Laboratories, Eastman Kodak Company, Rochester, New York 14650 (United States); Illiberi, Andrea; Poodt, Paul [Holst Centre/TNO Thin Film Technology, Eindhoven, 5656 AE (Netherlands); Roozeboom, Fred [Holst Centre/TNO Thin Film Technology, Eindhoven, 5656 AE (Netherlands); Department of Applied Physics, Eindhoven University of Technology, P.O. Box 513, Eindhoven, 5600 MB (Netherlands)

2015-04-01

Atmospheric pressure spatial atomic layer deposition (AP-SALD) has recently emerged as an appealing technique for rapidly producing high quality oxides. Here, we focus on the use of AP-SALD to deposit functional ZnO thin films, particularly on the reactors used, the film properties, and the dopants that have been studied. We highlight how these films are advantageous for the performance of solar cells, organometal halide perovskite light emitting diodes, and thin-film transistors. Future AP-SALD technology will enable the commercial processing of thin films over large areas on a sheet-to-sheet and roll-to-roll basis, with new reactor designs emerging for flexible plastic and paper electronics.

China: EDF's feedback experience of reactor operating is essential to win international markets

International Nuclear Information System (INIS)

Maillart, H.

2016-01-01

The main assets of EDF on the Chinese nuclear power market is first, its very important feedback experience of reactor operations (EDF cumulates one year of reactor operations every week due to its fleet of 58 reactors), secondly the cooperation with China allowed China to enter nuclear energy in 1983 with the construction of the Daya Bay plant and now to develop its own technology: the CPR-1000 reactor. China is the world leader in terms of nuclear market dynamism with 30 reactors in operation, 24 reactors being built and 40 others planned. A new stage in the Franco-China cooperation would be to share relevant good practices in the managing of both French and Chinese fleets of reactors. EDF has upgraded its commercial international offer, it now proposes to cover all the stages of the nuclear power plant from site selection to plant deconstruction via construction, operation, maintenance and waste management which constitutes a commitment over a 100 year period. (A.C.)

Fast breeder reactor-block antiseismic design and verification

International Nuclear Information System (INIS)

Martelli, A.; Forni, M.

1988-02-01

The Specialists' Meeting on ''Fast Breeder Reactor-Block Antiseismic Design and Verification'' was organized by the ENEA Fast Reactor Department in co-operation with the International Working Group (IWGFR) of the International Atomic Energy Agency (IAEA), according to the recommendations of the 19th IAEA/IWGFR Meeting. It was held in Bologna, at the Headquarters of the ENEA Fast Reactor Department, on October 12-15, 1987, in the framework of the Celebrations for the Ninth Centenary of the Bologna University. The proceedings of the meeting consists of three parts. Part 1 contains the introduction and general comments, the agenda of the meeting, session summaries, conclusions and recommendations and the list of participants. Part 2 contains 8 status reports of Member States participating in the Working Group. Contributed papers were published in Part 3 and were further subdivided into 5 sessions as follows: whole reactor-block analysis (4 papers); whole reactor-block analysis (sloshing and buckling, seismic isolation effects) (8 papers); detailed core analysis (6 papers); shutdown systems and core structural and functional verifications (6 papers); component and piping analysis (7 papers). A separate abstract was prepared for each of the 8 status reports and 31 contributed papers. Refs, figs and tabs

International bulletin on atomic and molecular data for fusion. No. 42-45

International Nuclear Information System (INIS)

Botero, J.

1991-01-01

The bulletin is published by the International Atomic Energy Agency to provide atomic and molecular data relevant to fusion research and technology. In Part I the indexed papers are listed separately for (i) structure and spectra (energy levels, wavelengths; transition probabilities, oscillator strengths; polarizabilities, electric moments; interatomic potentials); (ii) atomic and molecular collisions (photon collisions; electro collisions; heavy-particle collisions; homonuclear sequences), and (iii) surface interactions (sputtering; trapping, detrapping; adsorption, desorption; surface damage; blistering, flaking; chemical reactions). Part II contains the bibliographic data for the above listed topics and for plasma composition and impurities; plasma heating, cooling and fuelling; fusion research of general interest; high energy laser- and beam-matter interaction; interaction of atomic particles with fields. A list of evaluated data bases on atomic and molecular collisions and on particle-surface interactions is also given

CAC-RA1 1958-1998. The first years of the Constituyentes Atomic Center (CAC). History of the first Argentine nuclear reactor (RA-1)

International Nuclear Information System (INIS)

Forlerer, Elena; Palacios, Tulio A.

1998-01-01

After giving the milestones of the development of the Constituyentes Atomic Center since 1957, the history of the construction of the first nuclear reactor (RA-1) in Argentina, including the local fabrication of its fuel elements, is surveyed. The RA-1 reached criticality on January 17, 1958. The booklet commemorates the 40th year of the reactor operation

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

Advancing the CANDU reactor: From generation to generation

International Nuclear Information System (INIS)

Hopwood, Jerry; Duffey, Romney B.; Yu, Steven; Torgerson, Dave F.

2006-01-01

Emphasizing safety, reliability and economics, the CANDU reactor development strategy is one of continuous improvement, offering value and assured support to customers worldwide. The Advanced CANDU Reactor (ACR-1000) generation, designed by Atomic Energy of Canada Limited (AECL), meets the new economic expectation for low-cost power generation with high capacity factors. The ACR is designed to meet customer needs for reduced capital cost, shorter construction schedule, high plant capacity factor, low operating cost, increased operating life, simple component replacement, enhanced safety features, and low environmental impact. The ACR-1000 design evolved from the internationally successful medium-sized pressure tube reactor (PTR) CANDU 6 and incorporates operational feedback from eight utilities that operate 31 CANDU units. This technical paper provides a brief description of the main features of the ACR-1000, and its major role in the development path of the generations of the pressure tube reactor concept. The motivation, philosophy and design approach being taken for future generation of CANDU pressure tube reactors are described

Combined development of international nuclear fusion test reactors

International Nuclear Information System (INIS)

Anon.

1987-01-01

Ambassadors of the four most important partners (Common Market, Japan, USA and USSR) in the IAEA sponsored INTOR project, met on the 15 and 16 March 1987 in Vienna under the auspices of the IAEA. A press release was issued acknowledging the considerable technical progress made in magnetic nuclear fusion research. Future design concepts, assistance in research and development work and other activities towards the provision of an international experimental thermonuclear reactor were discussed. (G.T.H.)

Proceedings of the international seminar on atomic processes in plasmas

Energy Technology Data Exchange (ETDEWEB)

Kato, Takako; Murakami, Izumi [eds.

2000-01-01

The International Seminar on Atomic Processes in Plasmas (ISAPP), a satellite meeting to the ICPEAC was held July 28-29 at the National Institute for Fusion Science in Toki, Gifu, Japan. About 110 scientists attended the ISAPP meeting and discussed atomic processes and atomic data required for fusion research. This Proceedings book includes the papers of the talks, posters and panel discussion given at the meeting. The invited talks described the super configuration array method for complex spectra, near-LTE atomic kinetics, R-matrix calculations, the binary-encounter dipole model for electron-impact ionization of molecules, other calculations of molecular processes, the ADAS project and the NIFS atomic data-base, and a survey of the role of molecular processes in divertor plasmas. On the experimental side crossed-beam ion-ion collision-experiments for charge transfer, and storage-ring and EBIT measurements of ionization, excitation and dielectronic recombination cross-sections were presented, and atomic processes important for x-ray laser experiments and x-ray spectroscopy of astrophysical plasmas were described. The new method of plasma polarization spectroscopy was outlined. There was also a spectroscopic study of particle transport in JT-60U, new results for detached plasmas, and a sketch of the first hot plasma experiments with the Large Helical Device recently completed at NIFS. The 63 of the presented papers are indexed individually. (J.P.N.)

Ukrainian Nuclear Society International Conference 'Modernization of the NPP with VVER reactor' (abstracts)

International Nuclear Information System (INIS)

Bar'yakhtar, V.G.

1999-01-01

Abstracts of the papers presented at International conference of the Ukrainian Nuclear Society 'Modernization of the NPP with VVER reactor'. The following problems are considered: improving the NPP's safety and reliability; reactor modernization, the lifetime prolongation; increasing of the reactor operating characteristics; methods of capacity factor increasing: refueling control, maintenance control; technical and economical aspects of NPP modernization; modernization of the automated control system of the fuel process at the NPP's; technical features and methods for the continued radiation and technology control at the NPP's; training, increasing the staff qualification and NPP modernization

Jacking mechanism for upper internals structure of a liquid metal nuclear reactor

International Nuclear Information System (INIS)

Gillett, J.E.; Wineman, A.L.

1983-01-01

A jacking mechanism for raising the upper internals structure of a liquid metal nuclear reactor which jacking mechanism uses a system of gears and drive shafts to transmit force from a single motor to four mechanically synchronized ball jacks to raise and lower support columns which support the upper internals structure. The support columns each have a pin which rides in a slot in a housing fixed to the reactor head. The pin has two locking plates which can be rotated around the pin to bring the locking plates into engagement with the housing in a raised or a lowered position of the support column such that the support column is then supported by the locking plate and not by the ball screw jacks. (author)

Crosscutting Requirements in the International Project on Innovative Reactors and Fuel Cycles (INPRO)

International Nuclear Information System (INIS)

Steur, Ronald; Lyubenov Yaven, Yanko; Gueorguiev, Boris; Mahadeva, Rao; Shen, Wenquan

2002-01-01

There are two categories of requirements: (i) user requirements that need to be met by the designers and manufacturers of innovative reactors and fuel cycles, and (ii) a wide spectrum of requirements that need to be met by countries, willing to successfully deploy innovative nuclear reactors for energy production. This part of the International Project on Innovative Reactors and Fuel Cycles will mainly deal with the second category of requirements. Both categories of requirements will vary depending on the institutional development, infrastructure availability and social attitude in any given country. Out of the need for sustainable development requirements will also more specific in the future. Over a 50-year time frame both categories of requirements will evolve with social and economic development as nuclear technology develops further. For example, the deployment of innovative reactors in countries with marginal or non-existing nuclear infrastructures would be possible only if the reactors are built, owned and operated by an international nuclear utility or if they are inherently safe and can be delivered as a 'black box - nuclear battery'. A number of issues will need to be addressed and conditions and requirements developed if this is going to become a reality. One general requirement for wider utilization of innovative nuclear power will be the public and environmental considerations, which will play a role in the decision making processes. Five main clusters of topics will be handled: - Infra-structural aspects, typology and consequences for nuclear development. - Industrial requirements for the different innovative concepts. - Institutional developments and requirements for future deployment of nuclear energy. (National as well as international) - Socio-political aspects, a.o. public acceptance and role of governments. - Sustainability: requirements following the need for sustainability Analysis will be made of the evolution of national and international

Handling of spent fuel from research reactors in Japan

International Nuclear Information System (INIS)

Kanda, K.

1997-01-01

In Japan eleven research reactors are in operation. After the 19th International Meeting on Reduced Enrichment for Research Reactors and Test Reactors (RERTR) on October 6-10, 1996, Seoul, Korea, the Five Agency Committee on Highly Enriched Uranium, which consists of Science and Technology Agency, the Ministry of Education, Science and Culture, the Ministry of Foreign Affairs, Japan Atomic Energy Research Institute (JAERI) and Kyoto University Research Reactor Institute (KURRI) met on November 7,1996, to discuss the handling of spent fuel from research reactors in Japan. Advantages and disadvantages to return spent fuel to the USA in comparison to Europe were discussed. So far, a number of spent fuel elements in JAERI and KURRI are to be returned to the US. The first shipment to the US is planned for 60 HEU elements from JMTR in 1997. The shipment from KURRI is planned to start in 1999. (author)

Scram device for atomic reactors

International Nuclear Information System (INIS)

Noyes, R.C.; Zaman, S.O.; Stuteville, D.W.

1978-01-01

A sensor chamber having one cavity containing coolant separated by a diaphragm from another cavity containing a fixed mass of inert gas is located within a safety assembly of a liquid metal-cooled nuclear reactor. The liquid cavity is in fluid communication with the coolant outside the chamber through a flow limiting orifice. An actuating bellows in fluid communication with the gas cavity is in contact with coolant outside the chamber and is connected to a push rod, which serves as a trigger for a poison bundle release mechanism. During slow changes in reactor coolant pressure experienced under normal operation, the diaphragm moves to equalize the gas cavity and liquid cavity pressures with the coolant pressure outside the chamber. The actuating bellows does not move because it is biased so that a threshold pressure difference is required before it will expand. Under a more rapid drop in coolant pressure, such as is associated with a loss of forced flow, the threshold is overcome and the actuating bellows will also move, thereby triggering the release mechanism to shut down the reactor. In an alternate embodiment, the actuating bellows is connected to the liquid cavity rather than to the gas cavity. (Auth.)

Review of inspection of Torness reactor internals using remote techniques

International Nuclear Information System (INIS)

Dynan, J.

1993-01-01

Inspection of reactor internals is increasingly being achieved by deployment of cameras into areas of sometimes high radiation fields using manipulators. The manipulators have been developed over a number of years and from a variety of sources and those provided at Torness can maybe be seen to be the state of the art development of such machines for AGRS. Torness and Heysham 2 reactors were specifically designed for inspection using remote techniques and special facilities have been provided for this purpose. This paper is written by an operator and not a manipulator specialist and is intended to show what the operator needs versus what the designer gave him. (author)

From the 'Austrian Foundation for Atomic Energy Research' to the 'Seibersdorf Laboratories'

International Nuclear Information System (INIS)

Rößner, M.

2013-01-01

The aim of this thesis is the description of the process of institutionalization of nuclear research in Austria in the context of the „Atoms for Peace“ program in the 1950s. This aspect of the history of Austrian nuclear research has been relatively unexplored. The focus of this work is the presentation of measures that have been put in Austria to participate in the 'Atoms for Peace' program. Moreover, the expectations and goals of the Austrian nuclear program are analyzed. For this purpose, foundation, structure and research activities of the “Austrian Atomic Energy Commission”, the “Austrian Society for the Study of Atomic Energy” and the “Nuclear Reactor Centre Seibersdorf“ in the period from 1954 until about 1970 are examined. To establish a relationship with the international historical research of the history of nuclear research, the Austrian situation is compared to the Swiss and the German. It turns out that the Austrian expectations and objectives in nuclear research at the beginning of the period of observation roughly corresponded with the international trends. It also follows that by the establishment of the SGAE and the “Nuclear Reactor Centre Seibersdorf“ the development course of Austrian nuclear research occupied a special position compared to foreign institutions.(author) [de

International Atomic Energy Agency use of facility calorimeters for safeguards purposes

International Nuclear Information System (INIS)

McRae, L.P.; Delegard, C.H.; Hamilton, R.A.; Westsik, G.A.; Moriarty, T.F.; Lemaire, R.J.

1996-01-01

The International Atomic Energy Agency is performing nuclear materials safeguards on an inventory of pure and scrap plutonium oxide powder materials held in Vault 3 of the Plutonium Finishing Plant, operated by the Westinghouse Hanford Company for the US Department of Energy at the Hanford Site in Washington State. The International Atomic Energy Agency uses qualitative and quantitative techniques to verify the presence and quantity of the nuclear materials under safeguards. The Agency uses weighing, sampling, and destructive analyses to obtain the most accurate verification measurements of containers of plutonium powders. In contrast, the plant operator generally uses non-destructive plutonium assay based on gamma spectrometry and calorimetry for its most accurate plutonium powder container measurements. Recent results have shown that the operator''s calorimeter system achieves measurement variabilities comparable with, or better than, the destructive analyses, particularly for scrap. The results are achieved more quickly and economically, with less waste and lower radiation exposure and contamination hazard, by calorimetry than by classical destructive analyses. Techniques, including authentication methods, are being jointly developed to permit use of the operator''s calorimeter system for international safeguards purposes. The authentication is to ensure the independence of, and to substantiate the validity of, calorimeter measurements for international safeguards. The authentication methods considered and being developed are discussed

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

International Nuclear Information System (INIS)

Lim, Dong-Won; Lee, Hyeong-Yeon; Eoh, Jae-Hyuk; Son, Seok-Kwon; Kim, Jong-Bum; Jeong, Ji-Young

2016-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-05-15

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

Dual-cell reduction and group effect in an internal microelectrolysis reactor

International Nuclear Information System (INIS)

Ying, Diwen; Peng, Juan; Li, Kan; Wang, Yalin; Pan, Siwen; Jia, JinPing

2013-01-01

Highlights: ► 2,4-DCP was removed simultaneously in cathode and anode cells in IME reactor. ► Mechanism of dual-cell reduction gave an insight of cathode and anode cells. ► Significant V oc increase with one/two electrodes couples being installed in series. ► Group effect reveals possible high redox potential in IME reactor. -- Abstract: To address some of the fundamental questions regarding the functions of cathodes and anodes (e.g., iron and granular active carbon) and what happens when numerous particles (microscopic galvanic cells) are combined in the widely used and efficient wastewater treatment of an internal microelectrolysis (IME) reactor, we employed a specifically designed dual-cell IME reactor with single-couple electrodes and a narrow cell IME reactor with multi-couple electrodes. The simultaneous removal of 6.36 mg L −1 and 2.93 mg L −1 of 2,4-dichlorophenol within 30 min from the graphite cathode and the iron anode cells, respectively, was observed in the dual-cell reactor. The innovative concept behind this phenomenon is that the anode, which is generally believed to be oxidative, could probably be reductive in an IME reactor. Thus, it is important to understand the unique performance of IME reactors. The group effect, which provides a 45% increase of open-circuit potential with just two additional electrode couples in aqueous solution, was tested and verified in a multi-couple electrode reactor. It suggests a complex potential distribution in the IME reactor and that compounds even with high redox potential could possibly be reduced, which was generally believed to be difficult to accomplish

Annual Report to Congress of the Atomic Energy Commission for 1970

Energy Technology Data Exchange (ETDEWEB)

Seaborg, Glenn T.

1971-01-29

The document represents the 1970 Annual Report of the Atomic Energy Commission (AEC) to Congress. The report opens with ''An Introduction to the Atomic Energy Programs during 1970'' followed by 14 Chapters, 8 appendices and an index. Chapters are as follows: (1) The Industrial Base; (2) Environmental and Safety Aspects; (3) Licensing and Regulating the Atom; (4) Source, Special, and Byproduct Nuclear Materials; (5) National Defense Programs; (6) Reactor Development and Technology; (7) Space Nuclear Systems; (8) Isotopic Systems Development; (9) Peaceful Nuclear Explosives; (10) International Affairs and Cooperation; (11) Nuclear Educational Activities; (12) Biomedical and Physical Research; (13) Administrative and Management Matters; and, (14) License Reviews and Adjudicatory Proceedings.

WWER reactor fuel performance, modelling and experimental support. Proceedings

International Nuclear Information System (INIS)

Stefanova, S.; Chantoin, P.; Kolev, I.

1994-01-01

This publication is a compilation of 36 papers presented at the International Seminar on WWER Reactor Fuel Performance, Modelling and Experimental Support, organised by the Institute for Nuclear Research and Nuclear Energy (BG), in cooperation with the International Atomic Energy Agency. The Seminar was attended by 76 participants from 16 countries, including representatives of all major Russian plants and institutions responsible for WWER reactor fuel manufacturing, design and research. The reports are grouped in four chapters: 1) WWER Fuel Performance and Economics: Status and Improvement Prospects: 2) WWER Fuel Behaviour Modelling and Experimental Support; 3) Licensing of WWER Fuel and Fuel Analysis Codes; 4) Spent Fuel of WWER Plants. The reports from the corresponding four panel discussion sessions are also included. All individual papers are recorded in INIS as separate items

WWER reactor fuel performance, modelling and experimental support. Proceedings

Energy Technology Data Exchange (ETDEWEB)

Stefanova, S; Chantoin, P; Kolev, I [eds.

1994-12-31

This publication is a compilation of 36 papers presented at the International Seminar on WWER Reactor Fuel Performance, Modelling and Experimental Support, organised by the Institute for Nuclear Research and Nuclear Energy (BG), in cooperation with the International Atomic Energy Agency. The Seminar was attended by 76 participants from 16 countries, including representatives of all major Russian plants and institutions responsible for WWER reactor fuel manufacturing, design and research. The reports are grouped in four chapters: (1) WWER Fuel Performance and Economics: Status and Improvement Prospects: (2) WWER Fuel Behaviour Modelling and Experimental Support; (3) Licensing of WWER Fuel and Fuel Analysis Codes; (4) Spent Fuel of WWER Plants. The reports from the corresponding four panel discussion sessions are also included. All individual papers are recorded in INIS as separate items.

Design of a new wet storage rack for spent fuels from IEA-R1 reactor

International Nuclear Information System (INIS)

Rodrigues, Antonio C.I.; Madi Filho, Tufic; Siqueira, Paulo T.D.; Ricci Filho, Walter

2015-01-01

The IEA-R1 research reactor operates in a regimen of 64h weekly, at the power of 4.5 MW. In these conditions, the racks of the spent fuel elements have less than half of its initial capacity. Thus, maintaining these operating conditions, the storage will have capacity for about six years. Since the estimated useful life of the IEA-R1 is about another 20 years, it will be necessary to increase the storage capacity of spent fuel. Dr. Henrik Grahn, expert of the International Atomic Energy Agency on wet storage, visiting the IEA-R1 Reactor (September/2012) made some recommendations: among them, the design and installation of racks made with borated stainless steel and internally coated with an aluminum film, so that corrosion of the fuel elements would not occur. After an extensive literature review of material options given for this type of application we got to Boral® manufactured by 3M due to numerous advantages. This paper presents studies on the analysis of criticality using the computer code MCNP 5, demonstrating the possibility of doubling the storage capacity of current racks to attend the demand of the IEA-R1 reactor while attending the safety requirements the International Atomic Energy Agency. (author)

Design of a new wet storage rack for spent fuels from IEA-R1 reactor

Energy Technology Data Exchange (ETDEWEB)

Rodrigues, Antonio C.I.; Madi Filho, Tufic; Siqueira, Paulo T.D.; Ricci Filho, Walter, E-mail: acirodri@ipen.br, E-mail: tmfilho@ipen.br, E-mail: ptsiquei@ipen.br, E-mail: wricci@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2015-07-01

The IEA-R1 research reactor operates in a regimen of 64h weekly, at the power of 4.5 MW. In these conditions, the racks of the spent fuel elements have less than half of its initial capacity. Thus, maintaining these operating conditions, the storage will have capacity for about six years. Since the estimated useful life of the IEA-R1 is about another 20 years, it will be necessary to increase the storage capacity of spent fuel. Dr. Henrik Grahn, expert of the International Atomic Energy Agency on wet storage, visiting the IEA-R1 Reactor (September/2012) made some recommendations: among them, the design and installation of racks made with borated stainless steel and internally coated with an aluminum film, so that corrosion of the fuel elements would not occur. After an extensive literature review of material options given for this type of application we got to Boral® manufactured by 3M due to numerous advantages. This paper presents studies on the analysis of criticality using the computer code MCNP 5, demonstrating the possibility of doubling the storage capacity of current racks to attend the demand of the IEA-R1 reactor while attending the safety requirements the International Atomic Energy Agency. (author)

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

Energy Technology Data Exchange (ETDEWEB)

Rosenbalm, K.F. [comp.

1995-12-31

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

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

International Nuclear Information System (INIS)

Rosenbalm, K.F.

1995-01-01

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

Annual report 1982-83 [of the Department of Atomic Energy, India

International Nuclear Information System (INIS)

1983-01-01

The annual report of the Department of Atomic Energy (DAE) of the Government of India for the financial year 1982-83 surveys the work of its various establishments. The major thrust of the DAE's programme is directed towards peaceful uses of atomic ener%y - primarily for electric power generation and applications of radiation and radioisotopes in medicine, agriculture and industry. The Bhabha Atomic Research Centre at Bombay is the major R and D establishment of the DAE and its activities in the fields of nuclear physics, chemistry and materials science, radiochemistry, nuclear fuels, reactor engineering, radiation protection, radioactive waste management and applications of radiation and radioactive isotopes are described in detail. The R and D activities of the Reactor Research Centre at Kanpakkam, the Tata Institute of Fundamental Research and the Tata A1emorial Centre, both at Bombay, and the Saha Institute of Nuclear Physics at Calcutta are described in brief. The performance of the Tarapur Atomic Power Station, the Rajasthan Atomic Power Station, the Nuclear Fuel Complex at Hyderabad, the Atomic Minerals Division, Uranium Corporation of India Ltd at Jaduguda, various heavy water plants and other industrial units of DAE is reported. Progress of nuclear power projects at Narora and Kakrapar, R-5 Project at Bombay and FBTR Project at Kalpakkam is described. India's participation in the activities of the International Atomic Energy Agency is also mentioned. (M.G.B.)

Upgraded prototype-reactor internal pump for ABWR

International Nuclear Information System (INIS)