WorldWideScience

Sample records for technology research reactor

  1. Technological Transfer from Research Nuclear Reactors to New Generation Nuclear Power Reactors

    Science.gov (United States)

    Radulescu, Laura; Pavelescu, Margarit

    2010-01-01

    The goal of this paper is the analysis of the technological transfer role in the nuclear field, with particular emphasis on nuclear reactors domain. The presentation is sustained by historical arguments. In this frame, it is very important to start with the achievements of the first nuclear systems, for instant those with natural uranium as fuel and heavy water as moderator, following in time through the history until the New Generation Nuclear Power Reactors. Starting with 1940, the accelerated development of the industry has implied the increase of the global demand for energy. In this respect, the nuclear energy could play an important role, being essentially an unlimited source of energy. However, the nuclear option faces the challenges of increasingly demanding safety requirements, economic competitiveness and public acceptance. Worldwide, a significant amount of experience has been accumulated during development, licensing, construction, and operation of nuclear power reactors. The experience gained is a strong basis for further improvements. Actually, the nuclear programs of many countries are addressing the development of advanced reactors, which are intended to have better economics, higher reliability, improved safety, and proliferation-resistant characteristics in order to overcome the current concerns about nuclear power. Advanced reactors, now under development, may help to meet the demand for energy power of both developed and developing countries as well as for district heating, desalination and for process heat. The paper gives historical examples that illustrate the steps pursued from first research nuclear reactors to present advanced power reactors. Emphasis was laid upon the fact that the progress is due to the great discoveries of the nuclear scientists using the technological transfer.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-02-17

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

  3. Recent expansion of research for light water reactor technology

    Energy Technology Data Exchange (ETDEWEB)

    Takahashi, Kiichi (Japan Electric Power Information Center, Tokyo (Japan)); Nemoto, Kazuyasu; Aoki, Norichika; Kusanagi, Hideo

    1990-09-01

    It is needless to say that for simultaneously coping with the increase of energy consumption and the prevention of the worsening of environment in the world, and for maintaining the standard of living in Japan where energy resources are scarce, the development of atomic energy is necessary. Though the technology of LWRs has been already established, the efforts of aiming at the further high safety and reliability of LWRs must be exerted. In this report, the recent technical development is described, centering around the research and technical development promoted by the Central Research Institute of Electric Power Industry. The energy consumption in the world recorded the yearly growth of about 3%, and in 1987, it was 9.65 billion tons in terms of coal (7000 kcal/kg). The problems of earth environment will relax by promoting atomic energy. As for the recent development of LWR technology, the research on existing LWRs, the research on the LWRs of next generation, the research on the new technology for locating nuclear facilities and the research on radiation are carried out. As the research aiming at the LWRs of next generation, the design and evaluation of statically safe LWRs, the evaluation of fuel behavior at high burnup and the development of new location technology are carried out. (K.I.).

  4. 75 FR 62892 - Massachusetts Institute of Technology Research Reactor Environmental Assessment and Finding of No...

    Science.gov (United States)

    2010-10-13

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Massachusetts Institute of Technology Research Reactor Environmental Assessment and Finding of No Significant Impact Correction In notice document 2010-24809 beginning on page 61220 in the issue of...

  5. Department of Reactor Technology

    DEFF Research Database (Denmark)

    Risø National Laboratory, Roskilde

    The general development of the Department of Reactor Technology at Risø during 1981 is presented, and the activities within the major subject fields are described in some detail. Lists of staff, publications, and computer programs are included.......The general development of the Department of Reactor Technology at Risø during 1981 is presented, and the activities within the major subject fields are described in some detail. Lists of staff, publications, and computer programs are included....

  6. Georgia Institute of Technology research on the Gas Core Actinide Transmutation Reactor (GCATR)

    Science.gov (United States)

    Clement, J. D.; Rust, J. H.; Schneider, A.; Hohl, F.

    1976-01-01

    The program reviewed is a study of the feasibility, design, and optimization of the GCATR. The program is designed to take advantage of initial results and to continue work carried out on the Gas Core Breeder Reactor. The program complements NASA's program of developing UF6 fueled cavity reactors for power, nuclear pumped lasers, and other advanced technology applications. The program comprises: (1) General Studies--Parametric survey calculations performed to examine the effects of reactor spectrum and flux level on the actinide transmutation for GCATR conditions. The sensitivity of the results to neutron cross sections are to be assessed. Specifically, the parametric calculations of the actinide transmutation are to include the mass, isotope composition, fission and capture rates, reactivity effects, and neutron activity of recycled actinides. (2) GCATR Design Studies--This task is a major thrust of the proposed research program. Several subtasks are considered: optimization criteria studies of the blanket and fuel reprocessing, the actinide insertion and recirculation system, and the system integration. A brief review of the background of the GCATR and ongoing research is presented.

  7. Nuclear research reactors in Brazil

    Energy Technology Data Exchange (ETDEWEB)

    Cota, Anna Paula Leite; Mesquita, Amir Zacarias, E-mail: aplc@cdtn.b, E-mail: amir@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-07-01

    The rising concerns about global warming and energy security have spurred a revival of interest in nuclear energy, giving birth to a 'nuclear power renaissance' in several countries in the world. Particularly in Brazil, in the recent years, the nuclear power renaissance can be seen in the actions that comprise its nuclear program, summarily the increase of the investments in nuclear research institutes and the government target to design and build the Brazilian Multipurpose research Reactor (BMR). In the last 50 years, Brazilian research reactors have been used for training, for producing radioisotopes to meet demands in industry and nuclear medicine, for miscellaneous irradiation services and for academic research. Moreover, the research reactors are used as laboratories to develop technologies in power reactors, which are evaluated today at around 450 worldwide. In this application, those reactors become more viable in relation to power reactors by the lowest cost, by the operation at low temperatures and, furthermore, by lower demand for nuclear fuel. In Brazil, four research reactors were installed: the IEA-R1 and the MB-01 reactors, both at the Instituto de Pesquisas Energeticas Nucleares (IPEN, Sao Paulo); the Argonauta, at the Instituto de Engenharia Nuclear (IEN, Rio de Janeiro) and the IPR-R1 TRIGA reactor, at the Centro de Desenvolvimento da Tecnologia Nuclear (CDTN, Belo Horizonte). The present paper intends to enumerate the characteristics of these reactors, their utilization and current academic research. Therefore, through this paper, we intend to collaborate on the BMR project. (author)

  8. Nuclear research reactors in Brazil

    Energy Technology Data Exchange (ETDEWEB)

    Cota, Anna Paula Leite; Mesquita, Amir Zacarias, E-mail: aplc@cdtn.b, E-mail: amir@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-07-01

    The rising concerns about global warming and energy security have spurred a revival of interest in nuclear energy, giving birth to a 'nuclear power renaissance' in several countries in the world. Particularly in Brazil, in the recent years, the nuclear power renaissance can be seen in the actions that comprise its nuclear program, summarily the increase of the investments in nuclear research institutes and the government target to design and build the Brazilian Multipurpose research Reactor (BMR). In the last 50 years, Brazilian research reactors have been used for training, for producing radioisotopes to meet demands in industry and nuclear medicine, for miscellaneous irradiation services and for academic research. Moreover, the research reactors are used as laboratories to develop technologies in power reactors, which are evaluated today at around 450 worldwide. In this application, those reactors become more viable in relation to power reactors by the lowest cost, by the operation at low temperatures and, furthermore, by lower demand for nuclear fuel. In Brazil, four research reactors were installed: the IEA-R1 and the MB-01 reactors, both at the Instituto de Pesquisas Energeticas Nucleares (IPEN, Sao Paulo); the Argonauta, at the Instituto de Engenharia Nuclear (IEN, Rio de Janeiro) and the IPR-R1 TRIGA reactor, at the Centro de Desenvolvimento da Tecnologia Nuclear (CDTN, Belo Horizonte). The present paper intends to enumerate the characteristics of these reactors, their utilization and current academic research. Therefore, through this paper, we intend to collaborate on the BMR project. (author)

  9. Nuclear Reactors and Technology

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-01-01

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

  10. Multi purpose research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Raina, V.K. [Research Reactor Design and Projects Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)]. E-mail: vkrain@magnum.barc.ernet.in; Sasidharan, K. [Research Reactor Design and Projects Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Sengupta, Samiran [Research Reactor Design and Projects Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Singh, Tej [Research Reactor Services Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)

    2006-04-15

    At present Dhruva and Cirus reactors provide the majority of research reactor based facilities to cater to the various needs of a vast pool of researchers in the field of material sciences, physics, chemistry, bio sciences, research and development work for nuclear power plants and production of radio isotopes. With a view to further consolidate and expand the scope of research and development in nuclear and allied sciences, a new 20 MWt multi purpose research reactor is being designed. This paper describes some of the design features and safety aspects of this reactor.

  11. Experimental research subject and renovation of chemical processing facility (CPF) for advanced fast reactor fuel reprocessing technology development

    Energy Technology Data Exchange (ETDEWEB)

    Koyama, Tomozo; Shinozaki, Tadahiro; Nomura, Kazunori; Koma, Yoshikazu; Miyachi, Shigehiko; Ichige, Yoshiaki; Kobayashi, Tsuguyuki; Nemoto, Shin-ichi [Japan Nuclear Cycle Development Inst., Tokai Works, Tokai, Ibaraki (Japan)

    2002-12-01

    In order to enhance economical efficiency, environmental impact and nuclear nonproliferation resistance, the Advanced Reprocessing Technology, such as simplification and optimization of process, and applicability evaluation of the innovative technology that was not adopted up to now, has been developed for the reprocessing of the irradiated fuel taken out from a fast reactor. Renovation of the hot cell interior equipments, establishment and updating of glove boxes, installation of various analytical equipments, etc. in the Chemical Processing Facility (CPF) was done to utilize the CPF more positivity which is the center of the experimental field, where actual fuel can be used, for research and development towards establishment of the Advanced Reprocessing Technology development. The hot trials using the irradiated fuel pins of the experimental fast reactor 'JOYO' for studies on improved aqueous reprocessing technology, MA separation technology, dry process technology, etc. are scheduled to be carried out with these new equipments. (author)

  12. Reactor Materials Research

    Energy Technology Data Exchange (ETDEWEB)

    Van Walle, E

    2001-04-01

    The activities of the Reactor Materials Research Department of the Belgian Nuclear Research Centre SCK-CEN in 2000 are summarised. The programmes within the department are focussed on studies concerning (1) fusion, in particular mechanical testing; (2) Irradiation Assisted Stress Corrosion Cracking (IASCC); (3) nuclear fuel; and (4) Reactor Pressure Vessel Steel (RPVS)

  13. INVAP's Research Reactor Designs

    Directory of Open Access Journals (Sweden)

    Eduardo Villarino

    2011-01-01

    Full Text Available INVAP, an Argentine company founded more than three decades ago, is today recognized as one of the leaders within the research reactor industry. INVAP has participated in several projects covering a wide range of facilities, designed in accordance with the requirements of our different clients. For complying with these requirements, INVAP developed special skills and capabilities to deal with different fuel assemblies, different core cooling systems, and different reactor layouts. This paper summarizes the general features and utilization of several INVAP research reactor designs, from subcritical and critical assemblies to high-power reactors.

  14. Advances in light water reactor technologies

    CERN Document Server

    Saito, Takehiko; Ishiwatari, Yuki; Oka, Yoshiaki

    2010-01-01

    ""Advances in Light Water Reactor Technologies"" focuses on the design and analysis of advanced nuclear power reactors. This volume provides readers with thorough descriptions of the general characteristics of various advanced light water reactors currently being developed worldwide. Safety, design, development and maintenance of these reactors is the main focus, with key technologies like full MOX core design, next-generation digital I&C systems and seismic design and evaluation described at length. This book is ideal for researchers and engineers working in nuclear power that are interested

  15. Technology, safety, and costs of decommissioning reference nuclear research and test reactors. Main report

    Energy Technology Data Exchange (ETDEWEB)

    Konzek, G.J.; Ludwick, J.D.; Kennedy, W.E. Jr.; Smith, R.I.

    1982-03-01

    Safety and Cost Information is developed for the conceptual decommissioning of two representative licensed nuclear research and test reactors. Three decommissioning alternatives are studied to obtain comparisons between costs (in 1981 dollars), occupational radiation doses, potential radiation dose to the public, and other safety impacts. The alternatives considered are: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and ENTOMB (entombment). The study results are presented in two volumes. Volume 1 (Main Report) contains the results in summary form.

  16. Refurbishment of existing research reactors for BNCT

    Energy Technology Data Exchange (ETDEWEB)

    Jatuff, F.E.; Gessaghi, V. [INVAP S.E., de Bariloche (Argentina)

    1997-12-01

    Some research reactors have been selected for the development of boron neutron capture therapy (BNCT) in the United States like the Massachusetts Institute of Technology research reactor, the University of Missouri research reactor 2 or the Brookhaven Medical Research Reactor, among others. These reactors have received excellent analyses and designs to accommodate extremely optimized beam shaping assemblies (BSAs) for the proper tuning of neutron spectra and absorption of undesired particles such as photons and fast neutrons. Due to the importance of BNCT in these facilities, the physicists and engineers have used many degrees of freedom for the optimization process.

  17. INVAP's Research Reactor Designs

    OpenAIRE

    Eduardo Villarino; Alicia Doval

    2011-01-01

    INVAP, an Argentine company founded more than three decades ago, is today recognized as one of the leaders within the research reactor industry. INVAP has participated in several projects covering a wide range of facilities, designed in accordance with the requirements of our different clients. For complying with these requirements, INVAP developed special skills and capabilities to deal with different fuel assemblies, different core cooling systems, and different reactor layouts. This paper ...

  18. Technology, safety, and costs of decommissioning reference nuclear research and test reactors. Appendices

    Energy Technology Data Exchange (ETDEWEB)

    Konzek, G.J.; Ludwick, J.D.; Kennedy, W.E. Jr.; Smith, R.I.

    1982-03-01

    Safety and Cost Information is developed for the conceptual decommissioning of two representative licensed nuclear research and test reactors. Three decommissioning alternatives are studied to obtain comparisons between costs (in 1981 dollars), occupational radiation doses, potential radiation dose to the public, and other safety impacts. The alternatives considered are: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and EMTOMB (entombment). The study results are presented in two volumes. Volume 2 (Appendices) contains the detailed data that support the results given in Volume 1, including unit-component data.

  19. Research on Precaution and Detection Technology for Flow Blockage of Plate-type Fuel Element in Research Reactors

    Institute of Scientific and Technical Information of China (English)

    DING; Li; QIAO; Ya-xin; ZHANG; Nian-peng; LUO; Bei-bei; HUA; Xiao; JIA; Shu-jie; YAN; Hui-yang

    2013-01-01

    The main aim of this study is to offer the technical support for safety operation and management of research reactors using plate-type fuel assemblies in China,which is performed from analysis of precaution measures for flow blockage and detection methods of accidents.Study shows that most accidents were induced by in-core foreign objects and the swelling of fuel

  20. Reactor Materials Research

    Energy Technology Data Exchange (ETDEWEB)

    Van Walle, E

    2002-04-01

    The activities of SCK-CEN's Reactor Materials Research Department for 2001 are summarised. The objectives of the department are: (1) to evaluate the integrity and behaviour of structural materials used in nuclear power industry; (2) to conduct research to unravel and understand the parameters that determine the material behaviour under or after irradiation; (3) to contribute to the interpretation, the modelling of the material behaviour and to develop and assess strategies for optimum life management of nuclear power plant components. The programmes within the department are focussed on studies concerning (1) Irradiation Assisted Stress Corrosion Cracking (IASCC); (2) nuclear fuel; and (3) Reactor Pressure Vessel Steel.

  1. Research Progress of Irradiation Embrittlement Behavior and Prediction Technology of Reactor Pressure Vessel Steel

    Institute of Scientific and Technical Information of China (English)

    YANG; Wen; TONG; Zhen-feng; NING; Guang-sheng; ZHANG; Chang-yi; BAI; Bing

    2015-01-01

    The reactor pressure vessel(RPV)is the core of the most important equipment in pressurized water reactor,and is the key equipment that cannot be replaced in nuclear power plant.The service life of RPV determines the use of nuclear power plant,and directly affects the safety and economy of nuclear power plant.Because of high temperature,high pressure and high-energy

  2. Nuclear Reactors and Technology; (USA)

    Energy Technology Data Exchange (ETDEWEB)

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

    1991-01-01

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

  3. Status of NDE research in the US-contributions of NDE to reactor safety and implementation of NDE technology

    Energy Technology Data Exchange (ETDEWEB)

    Ammirato, F. [EPRI, Charlotte, NC (United States)

    1999-08-01

    Power plant designers, plant owners, and regulators have developed inservice inspection (ISI) programs as part of their comprehensive approach to ensuring nuclear safety. This paper examines the role of ISI in reactor safety through several examples drawn from recent industry initiatives to address implementation of effective examination technology for nuclear power plant piping, and BWR and PWR reactor pressure vessels. These examples also illustrate the importance of well designed performance demonstration activities to support application of effective ISI. Finally, the efforts required to implement effective ISI technology for field inspection is addressed. (orig./DGE)

  4. Advanced Reactor Technology -- Regulatory Technology Development Plan (RTDP)

    Energy Technology Data Exchange (ETDEWEB)

    Moe, Wayne Leland [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-05-01

    This DOE-NE Advanced Small Modular Reactor (AdvSMR) regulatory technology development plan (RTDP) will link critical DOE nuclear reactor technology development programs to important regulatory and policy-related issues likely to impact a “critical path” for establishing a viable commercial AdvSMR presence in the domestic energy market. Accordingly, the regulatory considerations that are set forth in the AdvSMR RTDP will not be limited to any one particular type or subset of advanced reactor technology(s) but rather broadly consider potential regulatory approaches and the licensing implications that accompany all DOE-sponsored research and technology development activity that deal with commercial non-light water reactors. However, it is also important to remember that certain “minimum” levels of design and safety approach knowledge concerning these technology(s) must be defined and available to an extent that supports appropriate pre-licensing regulatory analysis within the RTDP. Final resolution to advanced reactor licensing issues is most often predicated on the detailed design information and specific safety approach as documented in a facility license application and submitted for licensing review. Because the AdvSMR RTDP is focused on identifying and assessing the potential regulatory implications of DOE-sponsored reactor technology research very early in the pre-license application development phase, the information necessary to support a comprehensive regulatory analysis of a new reactor technology, and the resolution of resulting issues, will generally not be available. As such, the regulatory considerations documented in the RTDP should be considered an initial “first step” in the licensing process which will continue until a license is issued to build and operate the said nuclear facility. Because a facility license application relies heavily on the data and information generated by technology development studies, the anticipated regulatory

  5. Advanced Reactor Technologies - Regulatory Technology Development Plan (RTDP)

    Energy Technology Data Exchange (ETDEWEB)

    Moe, Wayne L.

    2017-08-23

    This DOE-NE Advanced Small Modular Reactor (AdvSMR) regulatory technology development plan (RTDP) will link critical DOE nuclear reactor technology development programs to important regulatory and policy-related issues likely to impact a “critical path” for establishing a viable commercial AdvSMR presence in the domestic energy market. Accordingly, the regulatory considerations that are set forth in the AdvSMR RTDP will not be limited to any one particular type or subset of advanced reactor technology(s) but rather broadly consider potential regulatory approaches and the licensing implications that accompany all DOE-sponsored research and technology development activity that deal with commercial non-light water reactors. However, it is also important to remember that certain “minimum” levels of design and safety approach knowledge concerning these technology(s) must be defined and available to an extent that supports appropriate pre-licensing regulatory analysis within the RTDP. Final resolution to advanced reactor licensing issues is most often predicated on the detailed design information and specific safety approach as documented in a facility license application and submitted for licensing review. Because the AdvSMR RTDP is focused on identifying and assessing the potential regulatory implications of DOE-sponsored reactor technology research very early in the pre-license application development phase, the information necessary to support a comprehensive regulatory analysis of a new reactor technology, and the resolution of resulting issues, will generally not be available. As such, the regulatory considerations documented in the RTDP should be considered an initial “first step” in the licensing process which will continue until a license is issued to build and operate the said nuclear facility. Because a facility license application relies heavily on the data and information generated by technology development studies, the anticipated regulatory

  6. Assessment of Sensor Technologies for Advanced Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Korsah, Kofi [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Ramuhalli, Pradeep [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Vlim, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Kisner, Roger A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Britton, Jr, Charles L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wootan, D. W. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Anheier, Jr, N. C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Diaz, A. A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Hirt, E. H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Chien, H. T. [Argonne National Lab. (ANL), Argonne, IL (United States); Sheen, S. [Argonne National Lab. (ANL), Argonne, IL (United States); Bakhtiari, Sasan [Argonne National Lab. (ANL), Argonne, IL (United States); Gopalsami, S. [Argonne National Lab. (ANL), Argonne, IL (United States); Heifetz, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Tam, S. W. [Argonne National Lab. (ANL), Argonne, IL (United States); Park, Y. [Argonne National Lab. (ANL), Argonne, IL (United States); Upadhyaya, B. R. [Univ. of Tennessee, Knoxville, TN (United States); Stanford, A. [Univ. of Tennessee, Knoxville, TN (United States)

    2016-10-01

    Sensors and measurement technologies provide information on processes, support operations and provide indications of component health. They are therefore crucial to plant operations and to commercialization of advanced reactors (AdvRx). This report, developed by a three-laboratory team consisting of Argonne National Laboratory (ANL), Oak Ridge National Laboratory (ORNL) and Pacific Northwest National Laboratory (PNNL), provides an assessment of sensor technologies and a determination of measurement needs for AdvRx. It provides the technical basis for identifying and prioritizing research targets within the instrumentation and control (I&C) Technology Area under the Department of Energy’s (DOE’s) Advanced Reactor Technology (ART) program and contributes to the design and implementation of AdvRx concepts.

  7. Ship propulsion reactors technology; La technologie des reacteurs de propulsion navale

    Energy Technology Data Exchange (ETDEWEB)

    Fribourg, Ch. [Technicatome, Centre d' Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France)

    2002-07-01

    This paper takes the state of the art on ship propulsion reactors technology. The french research programs with the corresponding technological stakes, the reactors specifications and advantages are detailed. (A.L.B.)

  8. Green nanoparticle production using micro reactor technology

    Science.gov (United States)

    Kück, A.; Steinfeldt, M.; Prenzel, K.; Swiderek, P.; Gleich, A. v.; Thöming, J.

    2011-07-01

    The importance and potential of nanoparticles in daily life as well as in various industrial processes is becoming more predominant. Specifically, silver nanoparticles are increasingly applied, e.g. in clothes and wipes, due to their antibacterial properties. For applications in liquid phase it is advantageous to produce the nanoparticles directly in suspension. This article describes a green production of silver nanoparticles using micro reactor technology considering principles of green chemistry. The aim is to reveal the potential and constraints of this approach and to show, how economic and environmental costs vary depending on process conditions. For this purpose our research compares the proposed process with water-based batch synthesis and demonstrates improvements in terms of product quality. Because of the lower energy consumption and lower demand of cleaning agents, micro reactor is the best ecological choice.

  9. Research reactor de-fueling and fuel shipment

    Energy Technology Data Exchange (ETDEWEB)

    Ice, R.D.; Jawdeh, E.; Strydom, J.

    1998-08-01

    Planning for the Georgia Institute of Technology Research Reactor operations during the 1996 Summer Olympic Games began in early 1995. Before any details could be outlined, several preliminary administrative decisions had to be agreed upon by state, city, and university officials. The two major administrative decisions involving the reactor were (1) the security level and requirements and (2) the fuel status of the reactor. The Georgia Tech Research Reactor (GTRR) was a heavy-water moderated and cooled reactor, fueled with high-enriched uranium. The reactor was first licensed in 1964 with an engineered lifetime of thirty years. The reactor was intended for use in research applications and as a teaching facility for nuclear engineering students and reactor operators. Approximately one year prior to the olympics, the Georgia Tech administration decided that the GTRR fuel would be removed. In addition, a heightened, beyond regulatory requirements, security system was to be implemented. This report describes the scheduling, operations, and procedures.

  10. Development of essential system technologies for advanced reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Y. Y.; Hwang, Y. D.; Cho, B. H. and others

    1999-03-01

    Basic design of SMART adopts the new advanced technologies which were not applied in the existing 1000MWe PWR. However, the R and D experience on these advanced essential technologies is lacking in domestic nuclear industry. Recently, a research on these advanced technologies has been performed as a part of the mid-and-long term nuclear R and D program, but the research was limited only for the small scale fundamental study. The research on these essential technologies such as helically coiled tube steam generator, self pressurizer, core cooling by natural circulation required for the development of integral reactor SMART have not been conducted in full scale. This project, therefore, was performed for the development of analysis models and methodologies, system analysis and thermal hydraulic experiments on the essential technologies to be applied to the 300MWe capacity of integral reactor SMART and the advanced passive reactor expected to be developed in near future with the emphasis on experimental investigation. (author)

  11. Advanced nuclear reactor types and technologies

    Energy Technology Data Exchange (ETDEWEB)

    Ignatiev, V. [ed.; Feinberg, O.; Morozov, A. [Russian Research Centre `Kurchatov Institute`, Moscow (Russian Federation); Devell, L. [Studsvik Eco and Safety AB, Nykoeping (Sweden)

    1995-07-01

    The document is a comprehensive world-wide catalogue of concepts and designs of advanced fission reactor types and fuel cycle technologies. Two parts have been prepared: Part 1 Reactors for Power Production and Part 2 Heating and Other Reactor Applications. Part 3, which will cover advanced waste management technology, reprocessing and disposal for different nuclear fission options is planned for compilation during 1995. The catalogue was prepared according to a special format which briefly presents the project title, technical approach, development status, application of the technology, reactor type, power output, and organization which developed these designs. Part 1 and 2 cover water cooled reactors, liquid metal fast reactors, gas-cooled reactors and molten salt reactors. Subcritical accelerator-driven systems are also considered. Various reactor applications as power production, heat generation, ship propulsion, space power sources and transmutation of such waste are included. Each project is described within a few pages with the main features of an actual design using a table with main technical data and figure as well as references for additional information. Each chapter starts with an introduction which briefly describes main trends and approaches in this field. Explanations of terms and abbreviations are provided in a glossary.

  12. The current status of nuclear research reactor in Thailand

    Energy Technology Data Exchange (ETDEWEB)

    Sittichai, C.; Kanyukt, R.; Pongpat, P. [Office of Atomic Energy for Peace, Bangkok (Thailand)

    1998-10-01

    Since 1962, the Thai Research Reactor has been serving for various kinds of activities i.e. the production of radioisotopes for medical uses and research and development on nuclear science and technology, for more than three decades. The existing reactor site should be abandoned and relocated to the new suitable site, according to Thai cabinet`s resolution on the 27 December 1989. The decommissioning project for the present reactor as well as the establishment of new nuclear research center were planned. This paper discussed the OAEP concept for the decommissioning programme and the general description of the new research reactor and some related information were also reported. (author)

  13. Technical Research for Dedicated Isotope Production Reactor of South Africa

    Institute of Scientific and Technical Information of China (English)

    ZOU; Yao; LIU; Xing-min; CHEN; Hui-qiang; SUN; Zhen; WU; Yuan-yuan

    2012-01-01

    <正>Research reactor plays an important part in nuclear science and technology, application and power development. Currently, many countries in Middle East and Africa are ready to develop their own nuclear industry. South Africa sent its User Requirements Specification (URS) for a dedicated isotope production reactor to several institutes or companies, among of which Department of Reactor Engineering Research and Design (DRERD) in China Institute of Atomic Energy (CIAE) is a competitive candidate.

  14. Production of Fission Product 99Mo using High-Enriched Uranium Plates in Polish Nuclear Research Reactor MARIA: Technology and Neutronic Analysis

    Directory of Open Access Journals (Sweden)

    Jaroszewicz Janusz

    2014-07-01

    Full Text Available The main objective of 235U irradiation is to obtain the 99mTc isotope, which is widely used in the domain of medical diagnostics. The decisive factor determining its availability, despite its short lifetime, is a reaction of radioactive decay of 99Mo into 99mTc. One of the possible sources of molybdenum can be achieved in course of the 235U fission reaction. The paper presents activities and the calculation results obtained upon the feasibility study on irradiation of 235U targets for production of 99Mo in the MARIA research reactor. Neutronic calculations and analyses were performed to estimate the fission products activity for uranium plates irradiated in the reactor. Results of dummy targets irradiation as well as irradiation uranium plates have been presented. The new technology obtaining 99Mo is based on irradiation of high-enriched uranium plates in standard reactor fuel channel and calculation of the current fission power generation. Measurements of temperatures and the coolant flow in the molybdenum installation carried out in reactor SAREMA system give online information about the current fission power generated in uranium targets. The corrective factors were taken into account as the heat generation from gamma radiation from neighbouring fuel elements as well as heat exchange between channels and the reactor pool. The factors were determined by calibration measurements conducted with aluminium mock-up of uranium plates. Calculations of fuel channel by means of REBUS code with fine mesh structure and libraries calculated by means of WIMS-ANL code were performed.

  15. EMERGING TECHNOLOGY BULLETIN: SPOUTED BED REACTOR

    Science.gov (United States)

    The Spouted Bed Reactor (SBR) technology utilizes the unique attributes of the "spouting " fluidization regime, which can provide heat transfer rates comparable to traditional fluid beds, while providing robust circulation of highly heterogeneous solids, concurrent with very agg...

  16. Advanced research reactor fuel development

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Chang Kyu; Pak, H. D.; Kim, K. H. [and others

    2000-05-01

    The fabrication technology of the U{sub 3}Si fuel dispersed in aluminum for the localization of HANARO driver fuel has been launches. The increase of production yield of LEU metal, the establishment of measurement method of homogeneity, and electron beam welding process were performed. Irradiation test under normal operation condition, had been carried out and any clues of the fuel assembly breakdown was not detected. The 2nd test fuel assembly has been irradiated at HANARO reactor since 17th June 1999. The quality assurance system has been re-established and the eddy current test technique has been developed. The irradiation test for U{sub 3}Si{sub 2} dispersed fuels at HANARO reactor has been carried out in order to compare the in-pile performance of between the two types of U{sub 3}Si{sub 2} fuels, prepared by both the atomization and comminution processes. KAERI has also conducted all safety-related works such as the design and the fabrication of irradiation rig, the analysis of irradiation behavior, thermal hydraulic characteristics, stress analysis for irradiation rig, and thermal analysis fuel plate, for the mini-plate prepared by international research cooperation being irradiated safely at HANARO. Pressure drop test, vibration test and endurance test were performed. The characterization on powders of U-(5.4 {approx} 10 wt%) Mo alloy depending on Mo content prepared by rotating disk centrifugal atomization process was carried out in order to investigate the phase stability of the atomized U-Mo alloy system. The {gamma}-U phase stability and the thermal compatibility of atomized U-16at.%Mo and U-14at.%Mo-2at.%X(: Ru, Os) dispersion fuel meats at an elevated temperature have been investigated. The volume increases of U-Mo compatibility specimens were almost the same as or smaller than those of U{sub 3}Si{sub 2}. However the atomized alloy fuel exhibited a better irradiation performance than the comminuted alloy. The RERTR-3 irradiation test of nano

  17. Sodium fast reactor safety and licensing research plan. Volume II.

    Energy Technology Data Exchange (ETDEWEB)

    Ludewig, H. (Brokhaven National Laboratory, Upton, NY); Powers, D. A.; Hewson, John C.; LaChance, Jeffrey L.; Wright, A. (Argonne National Laboratory, Argonne, IL); Phillips, J.; Zeyen, R. (Institute for Energy Petten, Saint-Paul-lez-Durance, France); Clement, B. (IRSN/DPAM.SEMIC Bt 702, Saint-Paul-lez-Durance, France); Garner, Frank (Radiation Effects Consulting, Richland, WA); Walters, Leon (Advanced Reactor Concepts, Los Alamos, NM); Wright, Steve; Ott, Larry J. (Oak Ridge National Laboratory, Oak Ridge, TN); Suo-Anttila, Ahti Jorma; Denning, Richard (Ohio State University, Columbus, OH); Ohshima, Hiroyuki (Japan Atomic Energy Agency, Ibaraki, Japan); Ohno, S. (Japan Atomic Energy Agency, Ibaraki, Japan); Miyhara, S. (Japan Atomic Energy Agency, Ibaraki, Japan); Yacout, Abdellatif (Argonne National Laboratory, Argonne, IL); Farmer, M. (Argonne National Laboratory, Argonne, IL); Wade, D. (Argonne National Laboratory, Argonne, IL); Grandy, C. (Argonne National Laboratory, Argonne, IL); Schmidt, R.; Cahalen, J. (Argonne National Laboratory, Argonne, IL); Olivier, Tara Jean; Budnitz, R. (Lawrence Berkeley National Laboratory, Berkeley, CA); Tobita, Yoshiharu (Japan Atomic Energy Agency, Ibaraki, Japan); Serre, Frederic (Centre d' %C3%94etudes nucl%C3%94eaires de Cadarache, Cea, France); Natesan, Ken (Argonne National Laboratory, Argonne, IL); Carbajo, Juan J. (Oak Ridge National Laboratory, Oak Ridge, TN); Jeong, Hae-Yong (Korea Atomic Energy Research Institute, Daejeon, Korea); Wigeland, Roald (Idaho National Laboratory, Idaho Falls, ID); Corradini, Michael (University of Wisconsin-Madison, Madison, WI); Thomas, Justin (Argonne National Laboratory, Argonne, IL); Wei, Tom (Argonne National Laboratory, Argonne, IL); Sofu, Tanju (Argonne National Laboratory, Argonne, IL); Flanagan, George F. (Oak Ridge National Laboratory, Oak Ridge, TN); Bari, R. (Brokhaven National Laboratory, Upton, NY); Porter D. (Idaho National Laboratory, Idaho Falls, ID); Lambert, J. (Argonne National Laboratory, Argonne, IL); Hayes, S. (Idaho National Laboratory, Idaho Falls, ID); Sackett, J. (Idaho National Laboratory, Idaho Falls, ID); Denman, Matthew R.

    2012-05-01

    Expert panels comprised of subject matter experts identified at the U.S. National Laboratories (SNL, ANL, INL, ORNL, LBL, and BNL), universities (University of Wisconsin and Ohio State University), international agencies (IRSN, CEA, JAEA, KAERI, and JRC-IE) and private consultation companies (Radiation Effects Consulting) were assembled to perform a gap analysis for sodium fast reactor licensing. Expert-opinion elicitation was performed to qualitatively assess the current state of sodium fast reactor technologies. Five independent gap analyses were performed resulting in the following topical reports: (1) Accident Initiators and Sequences (i.e., Initiators/Sequences Technology Gap Analysis), (2) Sodium Technology Phenomena (i.e., Advanced Burner Reactor Sodium Technology Gap Analysis), (3) Fuels and Materials (i.e., Sodium Fast Reactor Fuels and Materials: Research Needs), (4) Source Term Characterization (i.e., Advanced Sodium Fast Reactor Accident Source Terms: Research Needs), and (5) Computer Codes and Models (i.e., Sodium Fast Reactor Gaps Analysis of Computer Codes and Models for Accident Analysis and Reactor Safety). Volume II of the Sodium Research Plan consolidates the five gap analysis reports produced by each expert panel, wherein the importance of the identified phenomena and necessities of further experimental research and code development were addressed. The findings from these five reports comprised the basis for the analysis in Sodium Fast Reactor Research Plan Volume I.

  18. Nuclear research reactors activities in INVAP

    Energy Technology Data Exchange (ETDEWEB)

    Ordonez, Juan Pablo [INVAP, Bariloche (Argentina)

    2013-07-01

    This presentation describes the different activities in the research reactor field that are being carried out by INVAP. INVAP is presently involved in the design of three new research reactors in three different countries. The RA-10 is a multipurpose reactor, in Argentina, planned as a replacement for the RA-3 reactor. INVAP was contracted by CNEA for carrying out the preliminary engineering for this reactor, and has recently been contracted by CNEA for the detailed engineering. CNEA groups are strongly involved in the design of this reactor. The RMB is a multipurpose reactor, planned by CNEN from Brazil. CNEN, through REDETEC, has contracted INVAP to carry out the preliminary engineering for this reactor. As the user requirements for RA-10 and RMB are very similar, an agreement was signed between Argentina and Brasil governments to cooperate in these two projects. The agreement included that both reactors would use the OPAL reactor in Australia, design and built by INVAP, as a reference reactor. INVAP has also designed the LPRR reactor for KACST in Saudi Arabia. The LPRR is a 30 kw reactor for educational purposes. KACST initially contracted INVAP for the engineering for this reactor and has recently signed the contract with INVAP for building the reactor. General details of these three reactors will be presented.

  19. Reactor Safety Research: Semiannual report, January-June 1986: Reactor Safety Research Program

    Energy Technology Data Exchange (ETDEWEB)

    1987-05-01

    Sandia National Laboratories is conducting, under USNRC sponsorship, phenomenological research related to the safety of commercial nuclear power reactors. The research includes experiments to simulate the phenomenology of accident conditions and the development of analytical models, verified by experiment, which can be used to predict reactor and safety systems performance behavior under abnormal conditions. The objective of this work is to provide NRC requisite data bases and analytical methods to (1) identify and define safety issues, (2) understand the progression of risk-significant accident sequences, and (3) conduct safety assessments. The collective NRC-sponsored effort at Sandia National Laboratories is directed at enhancing the technology base supporting licensing decisions.

  20. New design targets and new automated technology for the production of radionuclides with high specificity radioactivity in nuclear research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Gerasimov, A.S.; Kiselev, G.V. [State Russian Center of the Russian Federation, Moscow (Russian Federation). Institute of Theoretical and Experimental Physics

    1997-10-01

    Current demands of industry require the application of radionuclides with high specific radioactivity under low consumption of neutrons. To provide this aim staff of ITEP Reactor Department investigated the different type AEs of start targets for the production of the main radionuclides; Co-60, Ir-192 and others. In first turn the targets of Co and Ir without the block-effect of neutron flux (with low absorption of neutrons) were investigated. The following principal results were received for example for Ir-192: block effect is equal 0.086 for diameter of Ir target mm and is equal 0.615 for diameter Ir target 0.5mm. It means average neutron flux for Ir target diameter 0.5mm and therefore the production of Ir-192 will be at 10 times more than for diameter 6.0mm. To provide the automated technology of the manufacture of radioactive sources with radionuclides with high specific radioactivity it was proposed that the compound targets for the irradiation of ones and for the management with the irradiated targets. Different types of compound targets were analyzed. (authors)

  1. Development of inherent core technologies for advanced reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Keung Koo; Noh, J.M.; Hwang, D.H. [and others

    1999-03-01

    Recently, the developed countries made their effort on developing the advanced reactor which will result in significantly enhanced safety and economy. However, they will protect the advanced reactor and its design technology with patent and proprietary right. Therefore, it is very important to develop our own key core concepts and inherent core design technologies which can form a foundation of indigenous technologies for development of the domestic advanced reactor in order to keep the superiority in the nuclear plant building market among the developing countries. In order to provide the basic technology for the core design of advanced reactor, this project is for developing the inherent core design concepts with enhanced safety and economy, and associated methodologies and technologies for core analyses. The feasibility study of constructing domestic critical facilities are performed by surveying the status and utilization of foreign facilities and by investigating the demand for domestic facilities. The research results developed in this project, such as core analysis methodologies for hexagonal core, conceptual core design based on hexagonal fuel assemblies and soluble boron core design and control strategies, will provide a technical foundation in developing core design of domestic advanced reactor. Furthermore, they will strengthen the competitiveness of Korean nuclear technology. We also expect that some of the design concepts developed in this project to improve the reactor safety and economy can be applicable to the design of advanced reactor. This will significantly reduce the public anxiety on the nuclear power plant, and will contribute to the economy of construction and operation for the future domestic reactors. Even though the critical facility will not be constructed right now, the investigation of the status and utilization of foreign critical facility will contribute to the future critical facility construction. (author). 150 refs., 34 tabs., 103

  2. Current Abstracts Nuclear Reactors and Technology

    Energy Technology Data Exchange (ETDEWEB)

    Bales, J.D.; Hicks, S.C. [eds.

    1993-01-01

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

  3. 10 MW research reactor simulation using fuel plate type

    Energy Technology Data Exchange (ETDEWEB)

    Mustafa, M. El Sayed, E-mail: memmm67@yahoo.com [Reactors Department, Nuclear Researches Center, Inshas (Egypt); Shaat, M. [Reactors Department, Nuclear Researches Center, Inshas (Egypt); Kady, M. El [Mechanical Power Engineering Department, Faculty of Engineering, Al Azhar University, Cairo (Egypt)

    2016-04-15

    A computer code was established named ET-RR-1-10 to investigate the thermal hydraulic behavior of the ETRR1 (first Egyptian research reactor) research reactor when its power upgraded to 10 MW using the new fuel plate elements type. The work done include both normal and flow reduction conditions. The code modeled primary loop, secondary lop, and reactor kinetics. All code models used finite difference technique. The code results were tested against the available corresponding experimental data taken from a similar research reactor MITR (Massachusetts Institute of Technology research reactor) for the sake of code validation. The results showed good agreement, and the code can be used for thermal hydraulic calculations.

  4. Seismic base isolation technologies for Korea advanced liquid metal reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, B.; Lee, J.-H.; Koo, G.-H.; Lee, H.-Y.; Kim, J.-B. [Korea Atomic Energy Research Inst., Taejon (Korea, Republic of)

    2000-06-01

    This paper describes the status and prospects of the seismic base isolation technologies for Korea Advanced Liquid Metal Reactor (KALIMER). The research and development program on the seismic base isolation for KALIMER began in 1993 by KAERI under the national long-term R and D program. The objective of this program is to enhance the seismic safety, to accomplish the economic design, and to standardize the plant design through the establishment of technologies on seismic base isolation for liquid metal reactors. In this paper, tests and analyses performed in the program are presented. (orig.)

  5. Is light water reactor technology sustainable?

    Energy Technology Data Exchange (ETDEWEB)

    Rothwell, G. [Stanford Univ., Dept. of Economics, CA (United States); Van der Zwaan, B. [Vrije Univ., Amsterdam, Inst. for Environmental Studies (Netherlands)

    2001-07-01

    This paper proposes criteria for determining ''intermediate sustainability'' over a 500-year horizon. We apply these criteria to Light Water Reactor (LWR) technology and the LWR industry. We conclude that LWR technology does not violate intermediate sustainability criteria for (1) environmental externalities, (2) worker and public health and safety, or (3) accidental radioactive release. However, it does not meet criteria to (1) efficiently use depleted uranium and (2) avoid uranium enrichment technologies that can lead to nuclear weapons proliferation. Finally, current and future global demand for LWR technology might be below the minimum needed to sustain the current global LWR industry. (author)

  6. Usage of burnable poison on research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Villarino, Eduardo Anibal [INVAP S.E., San Carlos de Bariloche (Argentina)

    2002-07-01

    The fuel assemblies with burnable poison are widely used on power reactors, but there are not commonly used on research reactors. This paper shows a neutronic analysis of the advantages and disadvantages of the burnable poison usage on research reactors. This paper analyses both burnable poison design used on research reactors: Boron on the lateral wall and Cadmium wires. Both designs include a parametric study on the design parameters like the amount and geometry of the burnable poison. This paper presents the design flexibility using burnable poisons, it does not find an optimal or final design, which it will strongly depend on the core characteristics and fuel management strategy. (author)

  7. Proceedings of the Twenty-First Water Reactor Safety Information Meeting: Volume 1, Plenary session; Advanced reactor research; advanced control system technology; advanced instrumentation and control hardware; human factors research; probabilistic risk assessment topics; thermal hydraulics; thermal hydraulic research for advanced passive LWRs

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-04-01

    This three-volume report contains 90 papers out of the 102 that were presented at the Twenty-First Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 25--27, 1993. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Germany, Japan, Russia, Switzerland, Taiwan, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. Individual papers have been cataloged separately. This document, Volume 1 covers the following topics: Advanced Reactor Research; Advanced Instrumentation and Control Hardware; Advanced Control System Technology; Human Factors Research; Probabilistic Risk Assessment Topics; Thermal Hydraulics; and Thermal Hydraulic Research for Advanced Passive Light Water Reactors.

  8. Role of research reactors for nuclear power program in Indonesia

    Energy Technology Data Exchange (ETDEWEB)

    Soentono, S.; Arbie, B. [National Atomic Energy Agency, Batan (Indonesia)

    1994-12-31

    The main objectives of nuclear development program in Indonesia are to master nuclear science and technology, as well as to utilise peaceful uses of nuclear know-how, aiming at stepwisely socioeconomic development. A Triga Mark II, previously of 250 kW, reactor in Bandung has been in operation since 1965 and its design power has been increased to 1000 kW in 1972. Using core grid of the Triga 250 kW, BATAN designed and constructed the Kartini Reactor in Yogyakarta which started its operation in 1979. Both of these Triga reactors have served a wide spectrum of utilisation, such as training of manpower in nuclear engineering as well as radiochemistry, isotope production and beam research activities in solid state physics. In order to support the nuclear power development program in general and to suffice the reactor experiments further, simultaneously meeting the ever increasing demand for radioisotope, the third reactor, a multipurpose reactor of 30 MW called GA. Siwabessy (RSG-GAS) has been in operation since 1987 at Serpong near Jakarta. Each of these reactors has strong cooperation with Universities, namely the Bandung Institute of Technology at Bandung, the Gadjah Mada University at Yogyakarta, and the Indonesia University at Jakarta and has facilitated the man power development required. The role of these reactors, especially the multipurpose GA. Siwabessy reactor, as essential tools in nuclear power program are described including the experience gained during preproject, construction and commissioning, as well as through their operation, maintenance and utilisation.

  9. The experimental and technological developments reactor; Le reacteur d'etudes et de developpements technologiques

    Energy Technology Data Exchange (ETDEWEB)

    Carbonnier, J.L. [CEA Cadarache, Dept. d' Etudes des Reacteurs (DEN/DER), 13 - Saint-Paul-lez-Durance (France)

    2003-07-01

    THis presentation concerns the REDT, gas coolant reactor for experimental and technological developments. The specifications and the research programs concerning this reactor are detailed;: materials, safety aspects, core physic, the corresponding fuel cycle, the reactor cycle and the program management. (A.L.B.)

  10. Technology, safety, and costs of decommissioning reference nuclear research and test reactors: sensitivity of decommissioning radiation exposure and costs to selected parameters

    Energy Technology Data Exchange (ETDEWEB)

    Konzek, G.J.

    1983-07-01

    Additional analyses of decommissioning at the reference research and test (R and T) reactors and analyses of five recent reactor decommissionings are made that examine some parameters not covered in the initial study report (NUREG/CR-1756). The parameters examined for decommissioning are: (1) the effect on costs and radiation exposure of plant size and/or type; (2) the effects on costs of increasing disposal charges and of unavailability of waste disposal capacity at licensed waste disposal facilities; and (3) the costs of and the available alternatives for the disposal of nuclear R and T reactor fuel assemblies.

  11. Renewing Liquid Fueled Molten Salt Reactor Research and Development

    Science.gov (United States)

    Towell, Rusty; NEXT Lab Team

    2016-09-01

    Globally there is a desperate need for affordable, safe, and clean energy on demand. More than anything else, this would raise the living conditions of those in poverty around the world. An advanced reactor that utilizes liquid fuel and molten salts is capable of meeting these needs. Although, this technology was demonstrated in the Molten Salt Reactor Experiment (MSRE) at ORNL in the 60's, little progress has been made since the program was cancelled over 40 years ago. A new research effort has been initiated to advance the technical readiness level of key reactor components. This presentation will explain the motivation and initial steps for this new research initiative.

  12. Research Reactor Design for Export to Myanmar

    Energy Technology Data Exchange (ETDEWEB)

    Win Naing, Lay Lay Myint and Myung-Hyun Kim [Kyunghee Univ. Yongin (Korea, Republic of)

    2006-07-01

    Myanmar is striving to acquire the innovative technology in all field areas including maritime, aerospace and nuclear engineering. There is a high intention to construct a new research reactor for peaceful purposes. The Ministry of Science and Technology (MOST) and Ministry of Education (MOE) are the important government organizations for Myanmar's education and they control most of institutes, universities and colleges. The Department of Atomic Energy (DAE), one of the departments under MOST, leads research projects such as for radiation protection as well as radiation application and coordinates government departments and institutions regarding nuclear energy and its applications. Myanmar's Scientific and Technological Research Department (MSTRD) under MOST guides researches in metallurgy, polymer, pharmacy and biotechnology and so on, and acts as an official body for Myanmar industrial standard. The Department of Higher Education (DHE) under MOE controls art and science universities and colleges including research centers such as Asia Research Center (ARC), Universities Research Center (URC), Microbiology Research Center and so on and does to expand research areas and to utilize advanced technology in science. The wide use of radiation and radioisotopes is developed in Myanmar especially for the field areas such as Medical Science and Agricultural Science. Co{sup 60}, I{sup 131} and Tc{sup 99} are the major use of radioisotopes in diagnosis and therapy. In Agricultural Science, H{sup 3}, C{sup 14}, C{sup 60} etc are used to provide biological effects of radiations on plants, radio-isotopic study of soil physics and tracer studies.

  13. Conceptual Study for development of a low power research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, C.; Kim, H. S.; Park, J. H.; Chae, H. T.; Lee, B. C. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-07-01

    Even though the nuclear society is again facing with difficult situations after Fukusima accident, some countries still continues to consider nuclear power as one option of national energy sources and to introduce nuclear energy. As a research reactor has been regarded as a step-stone to establish infrastructures for the nuclear power development program, some countries that have plan to introduce the nuclear power energy are considering to construct a research reactor. Particularly, a low power research reactor whose main purpose is basic researches on the nuclear technology and education/training would be of interest to developing countries when taking the economy and level of science and technology into consideration. And many low power research reactors at operation are obsolescent and their numbers are decreasing. Hence, some concepts on a low power research reactor are being studied for the future needs. This paper presents the conceptual study on the basic requirements and the preliminary design features of a low power research reactor.

  14. Performance of a multipurpose research electrochemical reactor

    Energy Technology Data Exchange (ETDEWEB)

    Henquin, E.R. [Programa de Electroquimica Aplicada e Ingenieria Electroquimica (PRELINE), Facultad de Ingenieria Quimica, Universidad Nacional del Litoral, Santiago del Estero 2829, S3000AOM Santa Fe (Argentina); Bisang, J.M., E-mail: jbisang@fiq.unl.edu.ar [Programa de Electroquimica Aplicada e Ingenieria Electroquimica (PRELINE), Facultad de Ingenieria Quimica, Universidad Nacional del Litoral, Santiago del Estero 2829, S3000AOM Santa Fe (Argentina)

    2011-07-01

    Highlights: > For this reactor configuration the current distribution is uniform. > For this reactor configuration with bipolar connection the leakage current is small. > The mass-transfer conditions are closely uniform along the electrode. > The fluidodynamic behaviour can be represented by the dispersion model. > This reactor represents a suitable device for laboratory trials. - Abstract: This paper reports on a multipurpose research electrochemical reactor with an innovative design feature, which is based on a filter press arrangement with inclined segmented electrodes and under a modular assembly. Under bipolar connection, the fraction of leakage current is lower than 4%, depending on the bipolar Wagner number, and the current distribution is closely uniform. When a turbulence promoter is used, the local mass-transfer coefficient shows a variation of {+-}10% with respect to its mean value. The fluidodynamics of the reactor responds to the dispersion model with a Peclet number higher than 10. It is concluded that this reactor is convenient for laboratory research.

  15. BNCT Technology Development on HANARO Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chun, Ki Jung; Park, Kyung Bae; Whang, Seung Ryul; Kim, Myong Seop

    2007-06-15

    So as to establish the biological effects of BNCT in the HANARO Reactor, biological damages in cells and animals with treatment of boron/neutron were investigated. And 124I-BPA animal PET image, analysis technology of the boron contents in the mouse tissues by ICP-AES was established. A Standard clinical protocol, a toxicity evaluation report and an efficacy investigation report of BNCT has been developed. Based on these data, the primary permission of clinical application was acquired through IRB of our hospital. Three cases of pre-clinical experiment for boron distribution and two cases of medium-sized animal simulation experiment using cat with verifying for 2 months after BNCT was performed and so the clinical demonstration with a patient was prepared. Also neutron flux, fast neutron flux and gamma ray dose of BNCT facility were calculated and these data will be utilized good informations for clinical trials and further BNCT research. For the new synthesis of a boron compound, o-carboranyl ethylamine, o-carboranylenepiperidine, o-carboranyl-THIQ and o-carboranyl-s-triazine derivatives were synthesized. Among them, boron uptake in the cancer cell of the triazine derivative was about 25 times than that of BPA and so these three synthesized methods of new boron compounds were patented.

  16. Reactor containment research and development

    Energy Technology Data Exchange (ETDEWEB)

    Weil, N. A.

    1963-06-15

    An outline is given of containment concepts, sources and release rates of energy, responses of containment structures, effects of projectiles, and leakage rates of radioisotopes, with particular regard to major reactor accidents. (T.F.H.)

  17. Power Control Method for Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Baang, Dane; Suh, Yongsuk; Park, Cheol [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    Considering safety-oriented design concept and other control environment, we developed a simple controller that provides limiting function of power change- rate as well as fine tracking performance. The design result has been well-proven via simulation and actual application to a TRIGA-II type research reactor. The proposed controller is designed to track the PDM(Power Demand) from operator input as long as maintaining the power change rate lower than a certain value for stable reactor operation. A power control method for a TRIGA-II type research reactor has been designed, simulated, and applied to actual reactor. The control performance during commissioning test shows that the proposed controller provides fine control performance for various changes in reference values (PDM), even though there is large measurement noise from neutron detectors. The overshoot at low power level is acceptable in a sense of reactor operation.

  18. Gaseous fuel nuclear reactor research

    Science.gov (United States)

    Schwenk, F. C.; Thom, K.

    1975-01-01

    Gaseous-fuel nuclear reactors are described; their distinguishing feature is the use of fissile fuels in a gaseous or plasma state, thereby breaking the barrier of temperature imposed by solid-fuel elements. This property creates a reactor heat source that may be able to heat the propellant of a rocket engine to 10,000 or 20,000 K. At this temperature level, gas-core reactors would provide the breakthrough in propulsion needed to open the entire solar system to manned and unmanned spacecraft. The possibility of fuel recycling makes possible efficiencies of up to 65% and nuclear safety at reduced cost, as well as high-thrust propulsion capabilities with specific impulse up to 5000 sec.

  19. Dismantling decontamination of research reactor equipment

    Energy Technology Data Exchange (ETDEWEB)

    Voronik, N. I.; Davydov, Yu. P.; Shatilo, N. N. [Institute of Radioecological Problems Belarus Ac. Sci., Minsk-Sosny (Belarus)

    1999-07-01

    The purpose of the work was to check applicability of the existing and new compositions for decontamination and their adjustment to the specific conditions dealing with operation of the research reactor. (author)

  20. 8th International School of Fusion Reactor Technology "Ettore Majorana"

    CERN Document Server

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

    1988-01-01

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

  1. Supply of enriched uranium for research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, H. [NUKEM GmbH, Alzenau (Germany)

    1997-08-01

    Since the RERTR-meeting In Newport/USA in 1990 the author delivered a series of papers in connection with the fuel cycle for research reactors dealing with its front-end. In these papers the author underlined the need for unified specifications for enriched uranium metal suitable for the production of fuel elements and made proposals with regard to the re-use of in Europe reprocessed highly enriched uranium. With regard to the fuel cycle of research reactors the research reactor community was since 1989 more concentrating on the problems of its back-end since the USA stopped the acceptance of spent research reactor fuel on December 31, 1988. Now, since it is apparent that these back-end problem have been solved by AEA`s ability to reprocess and the preparedness of the USA to again accept physically spent research reactor fuel the author is focusing with this paper again on the front-end of the fuel cycle on the question whether there is at all a safe supply of low and high enriched uranium for research reactors in the future.

  2. Strengthening IAEA Safeguards for Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Reid, Bruce D. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Anzelon, George A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Budlong-Sylvester, Kory [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-09-01

    During their December 10-11, 2013, workshop in Grenoble France, which focused on the history and future of safeguarding research reactors, the United States, France and the United Kingdom (UK) agreed to conduct a joint study exploring ways to strengthen the IAEA’s safeguards approach for declared research reactors. This decision was prompted by concerns about: 1) historical cases of non-compliance involving misuse (including the use of non-nuclear materials for production of neutron generators for weapons) and diversion that were discovered, in many cases, long after the violations took place and as part of broader pattern of undeclared activities in half a dozen countries; 2) the fact that, under the Safeguards Criteria, the IAEA inspects some reactors (e.g., those with power levels under 25 MWt) less than once per year; 3) the long-standing precedent of States using heavy water research reactors (HWRR) to produce plutonium for weapons programs; 4) the use of HEU fuel in some research reactors; and 5) various technical characteristics common to some types of research reactors that could provide an opportunity for potential proliferators to misuse the facility or divert material with low probability of detection by the IAEA. In some research reactors, for example, such characteristics include rapid on-line refueling, and a core design with room for such a large number of assemblies or targets that it is difficult to detect diversion or undeclared irradiation. In addition, infrastructure associated with research reactors, such as hot cells, where plutonium could be separated, could pose a safeguards challenge because, in some cases, they are not declared (because they are not located in the facility or because nuclear materials are not foreseen to be processed inside) and may not be accessible to inspectors in States without an Additional Protocol in force.

  3. Recent progress on tritium technology research and development for a fusion reactor in Japan Atomic Energy Agency

    Energy Technology Data Exchange (ETDEWEB)

    Hayashi, T.; Nakamura, H.; Kawamura, Y.; Iwai, Y.; Isobe, K.; Yamada, M.; Kurata, R.; Edao, Y. [Tritium Technology Group, Japan Atomic Energy Agency, Tokai-mura (Japan); Suzuki, T.; Oyaizu, M.; Yamanishi, T. [Tritium Technology Group, Japan Atomic Energy Agency, Rokkasho-mura (Japan)

    2015-03-15

    JAEA (Japan Atomic Energy Agency) manages 2 tritium handling laboratories: Tritium Processing Laboratory (TPL) in Tokai and DEMO-RD building in Rokkasho. TPL has been accumulating a gram level tritium safety handling experiences without any accidental tritium release to the environment for more than 25 years. Recently, our activities have focused on 3 categories, as follows. First, the development of a detritiation system for ITER. This task is the demonstration test of a wet Scrubber Column (SC) as a pilot scale (a few hundreds m{sup 3}/h of processing capacity). Secondly, DEMO-RD tasks are focused on investigating the general issues required for DEMO-RD design, such as structural materials like RAFM (Reduced Activity Ferritic/Martensitic steels) and SiC/SiC, functional materials like tritium breeder and neutron multiplier, and tritium. For the last 4 years, we have spent a lot of time and means to the construction of the DEMO-RD facility and to its licensing, so we have just started the actual research program with tritium and other radioisotopes. This tritium task includes tritium accountancy, tritium basic safety research such as tritium interactions with various materials, which will be used for DEMO-RD and durability. The third category is the recovery work from the Great East Japan earthquake (2011 earthquake). It is worth noting that despite the high magnitude of the earthquake, TPL was able to confine tritium properly without any accidental tritium release.

  4. Nuclear vapor thermal reactor propulsion technology

    Science.gov (United States)

    Maya, Isaac; Diaz, Nils J.; Dugan, Edward T.; Watanabe, Yoichi; McClanahan, James A.; Wen-Hsiung Tu, Carman, Robert L.

    1993-01-01

    The conceptual design of a nuclear rocket based on the vapor core reactor is presented. The Nuclear Vapor Thermal Rocket (NVTR) offers the potential for a specific impulse of 1000 to 1200 s at thrust-to-weight ratios of 1 to 2. The design is based on NERVA geometry and systems with the solid fuel replaced by uranium tetrafluoride (UF4) vapor. The closed-loop core does not rely on hydrodynamic confinement of the fuel. The hydrogen propellant is separated from the UF4 fuel gas by graphite structure. The hydrogen is maintained at high pressure (˜100 atm), and exits the core at 3,100 K to 3,500 K. Zirconium carbide and hafnium carbide coatings are used to protect the hot graphite from the hydrogen. The core is surrounded by beryllium oxide reflector. The nuclear reactor core has been integrated into a 75 klb engine design using an expander cycle and dual turbopumps. The NVTR offers the potential for an incremental technology development pathway to high performance gas core reactors. Since the fuel is readily available, it also offers advantages in the initial cost of development, as it will not require major expenditures for fuel development.

  5. Facility for a Low Power Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chalker, R. G.

    1949-09-14

    Preliminary investigation indicates that a reactor facility with ample research provisions for use by University or other interested groups, featuring safety in design, can be economically constructed in the Los Angeles area. The complete installation, including an underground gas-tight reactor building, with associated storage and experiment assembly building, administration offices, two general laboratory buildings, hot latoratory and lodge, can be constructed for approxinately $1,500,000. This does not include the cost of the reactor itself or of its auxiliary equipment,

  6. Strengthening IAEA Safeguards for Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Reid, Bruce D. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Anzelon, George A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Budlong-Sylvester, Kory [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-09-01

    During their December 10-11, 2013, workshop in Grenoble France, which focused on the history and future of safeguarding research reactors, the United States, France and the United Kingdom (UK) agreed to conduct a joint study exploring ways to strengthen the IAEA’s safeguards approach for declared research reactors. This decision was prompted by concerns about: 1) historical cases of non-compliance involving misuse (including the use of non-nuclear materials for production of neutron generators for weapons) and diversion that were discovered, in many cases, long after the violations took place and as part of broader pattern of undeclared activities in half a dozen countries; 2) the fact that, under the Safeguards Criteria, the IAEA inspects some reactors (e.g., those with power levels under 25 MWt) less than once per year; 3) the long-standing precedent of States using heavy water research reactors (HWRR) to produce plutonium for weapons programs; 4) the use of HEU fuel in some research reactors; and 5) various technical characteristics common to some types of research reactors that could provide an opportunity for potential proliferators to misuse the facility or divert material with low probability of detection by the IAEA. In some research reactors it is difficult to detect diversion or undeclared irradiation. In addition, infrastructure associated with research reactors could pose a safeguards challenge. To strengthen the effectiveness of safeguards at the State level, this paper advocates that the IAEA consider ways to focus additional attention and broaden its safeguards toolbox for research reactors. This increase in focus on the research reactors could begin with the recognition that the research reactor (of any size) could be a common path element on a large number of technically plausible pathways that must be considered when performing acquisition pathway analysis (APA) for developing a State Level Approach (SLA) and Annual Implementation Plan (AIP). To

  7. Joint KAERI/VAEC pre-possibility study on a new research reactor for Vietnam

    Energy Technology Data Exchange (ETDEWEB)

    Park, Cheol; Lee, B. C.; Chae, H. T.; Kim, H.; Lee, C. S.; Choi, C. O.; Jun, B. J. [KAERI, Taejon (Korea, Republic of); Vien, Luong Ba; Dien, Nguyen Nhi [Vietnam Atomic Energy Commission, Hanoi (Viet Nam)

    2004-05-01

    Based on the agreement on the technical cooperation for nuclear technology between Korea and Vietnam, a KAERI/VAEC joint study on the pre-possibility of a new research reactor for Vietnam has been carried out in the research reactor area from Nov. 2003 to May 2004. In this report, the results of the pre-possibility study on a new research reactor are described. The report presents the necessity of a new research reactor in Vietnam, and the desired performance requirements of the new research reactor if necessary. The major design characteristics of some existing research reactors and those under planning were also reviewed and the main characteristics which should be considered in selecting a new multipurpose research reactor for Vietnam were drawn. Some recommendations on the considerations for the next step of the feasibility study such as the project formulation, manpower requirements and international co-operation were also briefly touched upon.

  8. 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)

  9. Research on plasma core reactors

    Energy Technology Data Exchange (ETDEWEB)

    Jarvis, G.A.; Barton, D.M.; Helmick, H.H.; Bernard, W.; White, R.H.

    1977-01-01

    Experiments and theoretical studies are being conducted for NASA on critical assemblies with 1-m-diam by 1-m-long low-density cores surrounded by a thick beryllium reflector. These assemblies make extensive use of existing nuclear propulsion reactor components, facilities, and instrumentation. Due to excessive porosity in the reflector, the initial critical mass was 19 kg U(93.2). Addition of a 17-cm-thick by 89-cm-diam beryllium flux trap in the cavity reduced the critical mass to 7 kg when all the uranium was in the zone just outside the flux trap. A mockup aluminum UF/sub 6/ container was placed inside the flux trap and fueled with uranium-graphite elements. Fission distributions and reactivity worths of fuel and structural materials were measured. Finally, an 85,000-cm/sup 3/ aluminum canister in the central region was fueled with UF/sub 6/ gas and fission density distributions determined. These results will be used to guide the design of a prototype plasma core reactor which will test energy removal by optical radiation.

  10. Corrosion Minimization for Research Reactor Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Eric Shaber; Gerard Hofman

    2005-06-01

    Existing university research reactors are being converted to use low-enriched uranium fue to eliminate the use of highly-enriched uranium. These conversions require increases in fuel loading that will result in the use of elements with more fuel plates, resulting in a net decrease in the water annulus between fuel plates. The proposed decrease in the water annulus raises questions about the requirements and stability of the surface hydroxide on the aluminum fuel cladding and the potential for runaway corrosion resulting in fuel over-temperature incidents. The Nuclear Regulatory Commission (NRC), as regulator for these university reactors, must ensure that proposed fuel modifications will not result in any increased risk or hazard to the reactor operators or the public. This document reviews the characteristics and behavior of aluminum hydroxides, analyzes the drivers for fuel plate corrosion, reviews relevant historical incidents, and provides recommendations on fuel design, surface treatment, and reactor operational practices to avoid corrosion issues.

  11. Decommissioning of the Salaspils Research Reactor

    Directory of Open Access Journals (Sweden)

    Abramenkovs Andris

    2011-01-01

    Full Text Available In May 1995, the Latvian government decided to shut down the Salaspils Research Reactor and to dispense with nuclear energy in the future. The reactor has been out of operation since July 1998. A conceptual study on the decommissioning of the Salaspils Research Reactor was drawn up by Noell-KRC-Energie- und Umwelttechnik GmbH in 1998-1999. On October 26th, 1999, the Latvian government decided to start the direct dismantling to “green-field” in 2001. The upgrading of the decommissioning and dismantling plan was carried out from 2003-2004, resulting in a change of the primary goal of decommissioning. Collecting and conditioning of “historical” radioactive wastes from different storages outside and inside the reactor hall became the primary goal. All radioactive materials (more than 96 tons were conditioned for disposal in concrete containers at the radioactive wastes depository “Radons” at the Baldone site. Protective and radiation measurement equipment of the personnel was upgraded significantly. All non-radioactive equipment and materials outside the reactor buildings were released for clearance and dismantled for reuse or conventional disposal. Contaminated materials from the reactor hall were collected and removed for clearance measurements on a weekly basis.

  12. Reactor pulse repeatability studies at the annular core research reactor

    Energy Technology Data Exchange (ETDEWEB)

    DePriest, K.R. [Applied Nuclear Technologies, Sandia National Laboratories, Mail Stop 1146, Post Office Box 5800, Albuquerque, NM 87185-1146 (United States); Trinh, T.Q. [Nuclear Facility Operations, Sandia National Laboratories, Mail Stop 0614, Post Office Box 5800, Albuquerque, NM 87185-1146 (United States); Luker, S. M. [Applied Nuclear Technologies, Sandia National Laboratories, Mail Stop 1146, Post Office Box 5800, Albuquerque, NM 87185-1146 (United States)

    2011-07-01

    The Annular Core Research Reactor (ACRR) at Sandia National Laboratories is a water-moderated pool-type reactor designed for testing many types of objects in the pulse and steady-state mode of operations. Personnel at Sandia began working to improve the repeatability of pulse operations for experimenters in the facility. The ACRR has a unique UO{sub 2}-BeO fuel that makes the task of producing repeatable pulses difficult with the current operating procedure. The ACRR produces a significant quantity of photoneutrons through the {sup 9}Be({gamma}, n){sup 8}Be reaction in the fuel elements. The photoneutrons are the result of the gammas produced during fission and in fission product decay, so their production is very much dependent on the reactor power history and changes throughout the day/week of experiments in the facility. Because the photoneutrons interfere with the delayed-critical measurements required for accurate pulse reactivity prediction, a new operating procedure was created. The photoneutron effects at delayed critical are minimized when using the modified procedure. In addition, the pulse element removal time is standardized for all pulse operations with the modified procedure, and this produces less variation in reactivity removal times. (authors)

  13. Proposed replacement nuclear research reactor at Lucas Heights Science and Technology Centre, NSW. Statement of evidence to the Parliamentary Standing Committee on Public Works

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-02-01

    This submission demonstrates the manner in which the replacement research reactor project is to be undertaken in accordance with all relevant Commonwealth requirements and standards. Successive submissions to Government have shown that the construction and operation of the replacement reactor will result in a range of significant benefits to Australia in the areas of health care, the national interest, scientific achievement and in industrial applications. ANSTO is confident that the construction and operation of the replacement research reactor will: meet the identified needs for an ongoing neutron source for Australia into the next century in a cost-effective manner; be effectively managed to ensure that the project is delivered to the agreed schedule and budget; involve an effective community consultation process with ongoing community consultation a feature of ANSTO`s approach; will have negligible environmental and public health implications taking account of the environmental management measures and commitments made by ANSTO in the Environmental Impact Statement and the stringent licensing arrangement by ARPANSA 24 refs., 8 tabs., 5 figs.

  14. Reactor and process design in sustainable energy technology

    CERN Document Server

    Shi, Fan

    2014-01-01

    Reactor Process Design in Sustainable Energy Technology compiles and explains current developments in reactor and process design in sustainable energy technologies, including optimization and scale-up methodologies and numerical methods. Sustainable energy technologies that require more efficient means of converting and utilizing energy can help provide for burgeoning global energy demand while reducing anthropogenic carbon dioxide emissions associated with energy production. The book, contributed by an international team of academic and industry experts in the field, brings numerous reactor design cases to readers based on their valuable experience from lab R&D scale to industry levels. It is the first to emphasize reactor engineering in sustainable energy technology discussing design. It provides comprehensive tools and information to help engineers and energy professionals learn, design, and specify chemical reactors and processes confidently. Emphasis on reactor engineering in sustainable energy techn...

  15. Study for improvement of performance of the test and research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Sakurai, Fumio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-03-01

    Current utilization needs for the test and research reactors become more advanced and diversified along with the advance of nuclear science and technology. Besides, the requested safety for the research and test reactors grows strictly every year as well as a case of the power reactors. Under this circumstance, every effort to improve reactor performance including its safety is necessary to be sustained for allowing more effective utilization of the test and research reactors as experimental apparatus for advanced researches. In this study, the following three themes i.e., JMTR high-performance fuel element, evaluation method of fast neutron irradiation dose in the JMTR, evaluation method of performance of siphon break valve as core covering system for water-cooled test and research reactors, were investigated respectively from the views of improvement of core performance as a neutron source, utilization performance as an experimental apparatus, and safety as a reactor plant. (author)

  16. Licensing of the Australian replacement research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Garea, Veronica B. [INVAP S.E., San Carlos de Bariloche (Argentina)

    2002-07-01

    The Australian Nuclear Science and Technology Organisation (ANSTO)'s Replacement Research Reactor has been submitted to a comprehensive licensing process of which peer review has been a fundamental part. Following Australian Regulation, an application for a site licence was the first step supported by an Environmental Impact Statement approved by The Minister for the Environment and Heritage, and a Reference Accident Analysis. After the site licence had been granted and the contract awarded to the Designer and Constructor, INVAP S.E:, a 2500 page Preliminary Safety Analysis Report was submitted by ANSTO to the Australian Radiation Protection and Nuclear Safety Agency (ARPANSA), which conducted its review. ARPANSA requested that the PSAR be also reviewed by an experts mission from the International Atomic Energy Agency. The PSAR was also reviewed by the Argentine Regulatory Body, it was submitted to public examination in Australia and it was reviewed by international experts hired as consultants by several Australian organisations. A public forum was also held in Sydney. The Regulator, the applicant and the Designer-Constructor maintained constant interaction during the whole process, so that questions, comments and observations that arose from the review of the PSAR were fed back to the designers. This process allowed for a robust, safe design enriched by the results of the safety analysis and review process. (author)

  17. Remediation of Site of Decommissioning Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Danilovich, A.S.; Ivanov, O.P.; Lemus, A.V.; Pavlenko, V.I.; Potapov, V.N.; Semenov, S.G.; Shisha, A.D.; Chesnokov, A.V. [National Research Center ' Kurchatov Institute' , 123182, Moscow (Russian Federation)

    2014-07-01

    In the world the most widespread method of soil decontamination consists of removing the contaminated upper layer and sending it for long-term controlled storage. However, implementation of this soil cleanup method for remediation of large contaminated areas would involve high material and financial expenditures, because it produces large amounts of radioactive waste demanding removal to special storage sites. Contaminated soil extraction and cleanup performed right on the spot of remediation activities represents a more advanced and economically acceptable method. Radiological separation of the radioactive soil allows reducing of amount of radwaste. Studies performed during the liquidation of the Chernobyl accident consequences revealed that a considerable fraction of radioactivity is accumulated in minute soil grains. So, the separation of contaminated soil by size fractions makes it possible to extract and concentrate the major share of radioactivity in the fine fraction. Based on these researches water gravity separation technology was proposed by Bochvar Institute. The method extracts the fine fraction from contaminated soil. Studies carried out by Bochvar Institute experts showed that, together with the fine fraction (amounting to 10-20% of the initial soil), this technology can remove up to 85-90% of contaminating radionuclides. The resulting 'dirty' soil fraction could be packaged into containers and removed as radwaste, and decontaminated fractions returned back to their extraction site. Use of radiological and water gravity separations consequently increases the productivity of decontamination facility. Efficiency of this technology applied for contaminated soil cleanup was confirmed in the course of remediation of the contaminated territories near decommissioning research reactor in the Kurchatov Institute. For soil cleaning purposes, a special facility implementing the technology of water gravity separation and radiometric monitoring of soil

  18. Generation IV Reactors Integrated Materials Technology Program Plan: Focus on Very High Temperature Reactor Materials

    Energy Technology Data Exchange (ETDEWEB)

    Corwin, William R [ORNL; Burchell, Timothy D [ORNL; Katoh, Yutai [ORNL; McGreevy, Timothy E [ORNL; Nanstad, Randy K [ORNL; Ren, Weiju [ORNL; Snead, Lance Lewis [ORNL; Wilson, Dane F [ORNL

    2008-08-01

    Since 2002, the Department of Energy's (DOE's) Generation IV Nuclear Energy Systems (Gen IV) Program has addressed the research and development (R&D) necessary to support next-generation nuclear energy systems. The six most promising systems identified for next-generation nuclear energy are described within this roadmap. Two employ a thermal neutron spectrum with coolants and temperatures that enable hydrogen or electricity production with high efficiency (the Supercritical Water Reactor-SCWR and the Very High Temperature Reactor-VHTR). Three employ a fast neutron spectrum to enable more effective management of actinides through recycling of most components in the discharged fuel (the Gas-cooled Fast Reactor-GFR, the Lead-cooled Fast Reactor-LFR, and the Sodium-cooled Fast Reactor-SFR). The Molten Salt Reactor (MSR) employs a circulating liquid fuel mixture that offers considerable flexibility for recycling actinides and may provide an alternative to accelerator-driven systems. At the inception of DOE's Gen IV program, it was decided to significantly pursue five of the six concepts identified in the Gen IV roadmap to determine which of them was most appropriate to meet the needs of future U.S. nuclear power generation. In particular, evaluation of the highly efficient thermal SCWR and VHTR reactors was initiated primarily for energy production, and evaluation of the three fast reactor concepts, SFR, LFR, and GFR, was begun to assess viability for both energy production and their potential contribution to closing the fuel cycle. Within the Gen IV Program itself, only the VHTR class of reactors was selected for continued development. Hence, this document will address the multiple activities under the Gen IV program that contribute to the development of the VHTR. A few major technologies have been recognized by DOE as necessary to enable the deployment of the next generation of advanced nuclear reactors, including the development and qualification of

  19. Research, Development and Demonstration (RD&D) Needs for Light Water Reactor (LWR) Technologies A Report to the Reactor Technology Subcommittee of the Nuclear Energy Advisory Committee (NEAC) Office of Nuclear Energy U.S. Department of Energy

    Energy Technology Data Exchange (ETDEWEB)

    McCarthy, Kathryn A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Adams, Bradley J. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-04-01

    The LWR RD&D Working Group developed a detailed list of RD&D suggestions and recommendations, which are provided in Appendix D. The Working Group then undertook a systematic ranking process, described in Appendix E. The results of the ranking process are not meant to be a strict set of priorities, but rather should provide insight into how the items generally ranked within the Working Group. Future discussions and investigation into these items could provide information that would support a change in these priorities or in their emphasis. The results of this prioritization are provided below. Note that in general, many RD&D ideas are applicable to both new Advanced Light Water Reactor (ALWR) plants and currently operating plants.

  20. Conversion Preliminary Safety Analysis Report for the NIST Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Diamond, D. J. [Brookhaven National Lab. (BNL), Upton, NY (United States); Baek, J. S. [Brookhaven National Lab. (BNL), Upton, NY (United States); Hanson, A. L. [Brookhaven National Lab. (BNL), Upton, NY (United States); Cheng, L-Y [Brookhaven National Lab. (BNL), Upton, NY (United States); Brown, N. [Brookhaven National Lab. (BNL), Upton, NY (United States); Cuadra, A. [Brookhaven National Lab. (BNL), Upton, NY (United States)

    2015-01-30

    The NIST Center for Neutron Research (NCNR) is a reactor-laboratory complex providing the National Institute of Standards and Technology (NIST) and the nation with a world-class facility for the performance of neutron-based research. The heart of this facility is the NIST research reactor (aka NBSR); a heavy water moderated and cooled reactor operating at 20 MW. It is fueled with high-enriched uranium (HEU) fuel elements. A Global Threat Reduction Initiative (GTRI) program is underway to convert the reactor to low-enriched uranium (LEU) fuel. This program includes the qualification of the proposed fuel, uranium and molybdenum alloy foil clad in an aluminum alloy, and the development of the fabrication techniques. This report is a preliminary version of the Safety Analysis Report (SAR) that would be submitted to the U.S. Nuclear Regulatory Commission (NRC) for approval prior to conversion. The report follows the recommended format and content from the NRC codified in NUREG-1537, “Guidelines for Preparing and Reviewing Applications for the Licensing of Non-power Reactors,” Chapter 18, “Highly Enriched to Low-Enriched Uranium Conversions.” The emphasis in any conversion SAR is to explain the differences between the LEU and HEU cores and to show the acceptability of the new design; there is no need to repeat information regarding the current reactor that will not change upon conversion. Hence, as seen in the report, the bulk of the SAR is devoted to Chapter 4, Reactor Description, and Chapter 13, Safety Analysis.

  1. Complete dismantling of the research reactor DIORIT

    Energy Technology Data Exchange (ETDEWEB)

    Beer, Hans-Frieder [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

    2013-08-01

    The research reactor DIORIT at the Paul Scherrer Institute was a natural uranium reactor moderated by D{sub 2}O. It was put into operation in August 1960 and finally shut down in August 1977. The original dismantling plan, developed in 1980, comprised 3 phases and 13 steps. The dismantling started in 1982. It was interrupted for several times due to financial restrictions and during the last dismantling step due to the unexpected occurrence of asbestos. The dismantling could be successfully finished on September 11{sup th}, 2012. (orig.)

  2. Decommissioning Project for the Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chung, U. S.; Park, J. H.; Paik, S. T. (and others)

    2009-02-15

    In 2008, tried to complete the whole decommissioning project of KRR-1 and KRR-2 and preparing work for memorial museum of KRR-1 reactor. Now the project is delayed for 3 months because of finding unexpected soil contamination around facility and treatment of. To do final residual radioactivity assessment applied by MARSSIM procedure. Accumulated decommissioning experiences and technologies will be very usefully to do decommissioning other nuclear related facility. At the decommissioning site of the uranium conversion plant, the decontamination of the dismantled carbon steel waste are being performed and the lagoon 1 sludge waste is being treated this year. The technologies and experiences obtained from the UCP dismantling works are expected to apply to other fuel cycle facilities decommissioning. The lagoon sludge treatment technology is the first applied technology in the actual field and it is expected that this technology could be applied to other country.

  3. Development of mechanical design technology for integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Keun Bae; Choi, Suhn; Kim, Kang Soo; Kim, Tae Wan; Jeong, Kyeong Hoon; Lee, Gyu Mahn

    1999-03-01

    While Korean nuclear reactor strategy seems to remain focused on the large capacity power generation, it is expected that demand of small and medium size reactor will arise for multi-purpose application such as small capacity power generation, co-generation and sea water desalination. With this in mind, an integral reactor SMART is under development. Design concepts, system layout and types of equipment of integral reactor are significantly different from those of loop type reactor. Conceptual design development of mechanical structures of integral reactor SMART is completed through the first stage of the project. Efforts were endeavored for the establishment of design basis and evaluation of applicable codes and standards. Design and functional requirements of major structural components were setup, and three dimensional structural modelling of SMART reactor vessel assembly was prepared. Also, maintenance and repair scheme as well as preliminary fabricability evaluation were carried out. Since small integral reactor technology includes sensitive technologies and know-how's, it is hard to achieve systematic and comprehensive technology transfer from nuclear-advanced countries. Thus, it is necessary to develop the related design technology and to verify the adopted methodologies through test and experiments in order to assure the structural integrity of reactor system. (author)

  4. A novel concept for CRIEC-driven subcritical research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Nieto, M.; Miley, G.H. [Illinois Univ., Fusion Studies Lab., Dept. of Nuclear, Plasma, and Radiological Engineering, Urbana, IL (United States)

    2001-07-01

    A novel scheme is proposed to drive a low-power subcritical fuel assembly by means of a long Cylindrical Radially-convergent Inertial Electrostatic Confinement (CRIEC) used as a neutron source. The concept is inherently safe in the sense that the fuel assembly remains subcritical at all times. Previous work has been done for the possible implementation of CRIEC as a subcritical assembly driver for power reactors. However, it has been found that the present technology and stage of development of IEC-based neutron sources can not meet the neutron flux requirements to drive a system as big as a power reactor. Nevertheless, smaller systems, such as research and training reactors, could be successfully driven with levels of neutron flux that seem more reasonable to be achieved in the near future by IEC devices. The need for custom-made expensive nuclear fission fuel, as in the case of the TRIGA reactors, is eliminated, and the CRIEC presents substantial advantages with respect to the accelerator-driven subcritical reactors in terms of simplicity and cost. In the present paper, a conceptual design for a research/training CRIEC-driven subcritical assembly is presented, emphasizing the description, principle of operation and performance of the CRIEC neutron source, highlighting its advantages and discussing some key issues that require study for the implementation of this concept. (author)

  5. Neutron scattering at Australia's replacement research reactor

    Science.gov (United States)

    Robinson, R. A.; Kennedy, S. J.

    2002-01-01

    On August 25 1999, the Australian government gave final approval to build a research reactor to replace the existing HIFAR reactor at Lucas Heights. The replacement reactor, which will commence operation in 2005, will be multipurpose in function, with capabilities for both neutron-beam research and radioisotope production. Regarding beams, cold and thermal neutron sources are to be installed and the intent is to use supermirror guides, with coatings with critical angles up to 3 times that of natural Ni, to transport cold and thermal neutron beams into a large modern guide hall. The reactor and all the associated infrastructure, with the exception of the neutron beam instruments, is to be built by INVAP, SE and subcontractors in a turnkey contract. The goal is to have at least eight leading-edge neutron-beam instruments ready in 2005, and they will be developed by ANSTO and other contracted organisations, in consultation with the Australian user community and interested overseas parties. A review of the planned scientific capabilities, a description of the facility and a status report on the activities so far is given.

  6. The current status of Kartini research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tri Wulan Tjiptono; Syarip

    1998-10-01

    The Kartini reactor reached the first criticality on January 25, 1979. In the first three years, the reactor power is limited up to 50 kW thermal power and on July 1, 1982 has been increased to 100 kW. It has been used as experiments facility by researcher of Atomic Energy National Agency and students of the Universities. Three beam tubes used as experiments facilities, the first, is used as a neutron source for H{sub 2}O-Natural Uranium Subcritical Assembly, the second, is developed for neutron radiography facility and the third, is used for gamma radiography facility. The other facilities are rotary rack and two pneumatic transfer systems, one for delayed neutron counting system and the other for the new Neutron Activation Analysis (NAA) facility. The rotary rack used for isotope production for NAA purpose (for long time irradiation), the delayed neutron counting system used for analysis the Uranium contents of the ores and the new NAA is provided for short live elements analysis. In the last three years the Reactor Division has a joint use program with the Nuclear Component and Engineering Center in research reactor instrumentation and control development. (author)

  7. Nuclear Energy Enabling Technologies (NEET) Reactor Materials: News for the Reactor Materials Crosscut, May 2016

    Energy Technology Data Exchange (ETDEWEB)

    Maloy, Stuart Andrew [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Materials Science in Radiation and Dynamics Extremes

    2016-09-26

    In this newsletter for Nuclear Energy Enabling Technologies (NEET) Reactor Materials, pages 1-3 cover highlights from the DOE-NE (Nuclear Energy) programs, pages 4-6 cover determining the stress-strain response of ion-irradiated metallic materials via spherical nanoindentation, and pages 7-8 cover theoretical approaches to understanding long-term materials behavior in light water reactors.

  8. Reactor technology assessment and selection utilizing systems engineering approach

    Science.gov (United States)

    Zolkaffly, Muhammed Zulfakar; Han, Ki-In

    2014-02-01

    The first Nuclear power plant (NPP) deployment in a country is a complex process that needs to consider technical, economic and financial aspects along with other aspects like public acceptance. Increased interest in the deployment of new NPPs, both among newcomer countries and those with expanding programs, necessitates the selection of reactor technology among commercially available technologies. This paper reviews the Systems Decision Process (SDP) of Systems Engineering and applies it in selecting the most appropriate reactor technology for the deployment in Malaysia. The integrated qualitative and quantitative analyses employed in the SDP are explored to perform reactor technology assessment and to select the most feasible technology whose design has also to comply with the IAEA standard requirements and other relevant requirements that have been established in this study. A quick Malaysian case study result suggests that the country reside with PWR (pressurized water reactor) technologies with more detailed study to be performed in the future for the selection of the most appropriate reactor technology for Malaysia. The demonstrated technology assessment also proposes an alternative method to systematically and quantitatively select the most appropriate reactor technology.

  9. Kartini Research Reactor prospective studies for neutron scattering application

    Energy Technology Data Exchange (ETDEWEB)

    Widarto [Yogyakarta Nuclear Research Center, BATAN (Indonesia)

    1999-10-01

    The Kartini Research Reactor (KRR) is located in Yogyakarta Nuclear Research Center, Yogyakarta - Indonesia. The reactor is operated for 100 kW thermal power used for research, experiments and training of nuclear technology. There are 4 beam ports and 1 column thermal are available at the reactor. Those beam ports have thermal neutron flux around 10{sup 7} n/cm{sup 2}s each other and used for sub critical assembly, neutron radiography studies and Neutron Activation Analysis (NAA). Design of neutron collimator has been done for piercing radial beam port and the calculation result of collimated neutron flux is around 10{sup 9} n/cm{sup 2}s. This paper describes experiment facilities and parameters of the Kartini research reactor, and further more the prospective studies for neutron scattering application. The purpose of this paper is to optimize in utilization of the beam ports facilities and enhance the manpower specialty. The special characteristic of the beam ports and preliminary studies, pre activities regarding with neutron scattering studies for KKR is presented. (author)

  10. Development of technology for next generation reactor - Development of next generation reactor in Korea -

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong Kyun; Chang, Moon Heuy; Hwang, Yung Dong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)] [and others

    1993-09-01

    The project, development of next generation reactor, aims overall related technology development and obtainment of related license in 2001. The development direction is to determine the reactor type and to build up the design concept in 1994. For development trend analysis of foreign next generation reactor, level-1 PSA, fuel cycle analysis and computer code development are performed on System 80+ and AP 600. Especially for design characteristics analysis and volume upgrade of AP 600, nuclear fuel and reactor core design analysis, coolant circuit design analysis, mechanical structure design analysis and safety analysis etc. are performed. (Author).

  11. The past, present and potential for microfluidic reactor technology in chemical synthesis.

    Science.gov (United States)

    Elvira, Katherine S; Casadevall i Solvas, Xavier; Wootton, Robert C R; deMello, Andrew J

    2013-11-01

    The past two decades have seen far-reaching progress in the development of microfluidic systems for use in the chemical and biological sciences. Here we assess the utility of microfluidic reactor technology as a tool in chemical synthesis in both academic research and industrial applications. We highlight the successes and failures of past research in the field and provide a catalogue of chemistries performed in a microfluidic reactor. We then assess the current roadblocks hindering the widespread use of microfluidic reactors from the perspectives of both synthetic chemistry and industrial application. Finally, we set out seven challenges that we hope will inspire future research in this field.

  12. Convective cooling in a pool-type research reactor

    Science.gov (United States)

    Sipaun, Susan; Usman, Shoaib

    2016-01-01

    A reactor produces heat arising from fission reactions in the nuclear core. In the Missouri University of Science and Technology research reactor (MSTR), this heat is removed by natural convection where the coolant/moderator is demineralised water. Heat energy is transferred from the core into the coolant, and the heated water eventually evaporates from the open pool surface. A secondary cooling system was installed to actively remove excess heat arising from prolonged reactor operations. The nuclear core consists of uranium silicide aluminium dispersion fuel (U3Si2Al) in the form of rectangular plates. Gaps between the plates allow coolant to pass through and carry away heat. A study was carried out to map out heat flow as well as to predict the system's performance via STAR-CCM+ simulation. The core was approximated as porous media with porosity of 0.7027. The reactor is rated 200kW and total heat density is approximately 1.07+E7 Wm-3. An MSTR model consisting of 20% of MSTR's nuclear core in a third of the reactor pool was developed. At 35% pump capacity, the simulation results for the MSTR model showed that water is drawn out of the pool at a rate 1.28 kg s-1 from the 4" pipe, and predicted pool surface temperature not exceeding 30°C.

  13. Convective cooling in a pool-type research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sipaun, Susan, E-mail: susan@nm.gov.my [Malaysian Nuclear Agency, Industrial Technology Division, Blok 29T, Bangi 43200, Selangor (Malaysia); Usman, Shoaib, E-mail: usmans@mst.edu [Missouri University of Science and Technology, Nuclear Engineering, 222 Fulton Hall 301 W.14th St., Rolla 64509 MO (United States)

    2016-01-22

    A reactor produces heat arising from fission reactions in the nuclear core. In the Missouri University of Science and Technology research reactor (MSTR), this heat is removed by natural convection where the coolant/moderator is demineralised water. Heat energy is transferred from the core into the coolant, and the heated water eventually evaporates from the open pool surface. A secondary cooling system was installed to actively remove excess heat arising from prolonged reactor operations. The nuclear core consists of uranium silicide aluminium dispersion fuel (U{sub 3}Si{sub 2}Al) in the form of rectangular plates. Gaps between the plates allow coolant to pass through and carry away heat. A study was carried out to map out heat flow as well as to predict the system’s performance via STAR-CCM+ simulation. The core was approximated as porous media with porosity of 0.7027. The reactor is rated 200kW and total heat density is approximately 1.07+E7 Wm{sup −3}. An MSTR model consisting of 20% of MSTR’s nuclear core in a third of the reactor pool was developed. At 35% pump capacity, the simulation results for the MSTR model showed that water is drawn out of the pool at a rate 1.28 kg s{sup −1} from the 4” pipe, and predicted pool surface temperature not exceeding 30°C.

  14. Membrane reactor technology for ultrapure hydrogen production

    Energy Technology Data Exchange (ETDEWEB)

    Patil, C.S.

    2005-11-17

    The main objectives of this thesis are (1) to compare different reactor types and assess the feasibility of operation; (2) to develop and design a novel reactor concept based on the integration of perm-selective hydrogen and oxygen membranes; and (3) to give an experimental proof of principle of the developed reactor concept. In Chapter 2, available perm-selective hydrogen and oxygen membranes are reviewed. The focus is on the reactor concepts using these membranes and commercial developments that have taken place. In Chapter 3, the feasibility of performing autothermal membrane reforming in a packed bed membrane reactor with perm-selective hydrogen membrane is investigated based on detailed two-dimensional non-isothermal reactor modelling. In Chapter 4, an alternative reactor concept is developed for the autothermal reforming of methane integrating both hydrogen and oxygen perm-selective membranes. In Chapter 5, experimental work on the perm-selective hydrogen membranes that are used in the top section of the proposed reactor concept has been elaborated. These membranes, procured from a commercial supplier, are tested for their perm-selectivity and the permeability of hydrogen at different temperature and hydrogen partial pressures. Using the flux data a lumped flux expression is developed which is subsequently used in the pilot scale reactor design (Chapter 7). In Chapter 6, the kinetic rate measurements for SRM on a highly active Shell CPO catalyst are described. A kinetic rate expression for the steam reforming/ water gas shift top section of the proposed novel reactor concept is developed. The bottom section of this reactor is essentially at thermodynamic equilibrium because of highly active CPO catalyst and high temperatures and hence a detailed kinetic investigation for this section is not undertaken. In Chapter 7, a single membrane prototype of the top section is tested experimentally followed by a scale-up and design to a pilot scale unit with 10 Pd

  15. Research and Development Roadmaps for Liquid Metal Cooled Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, T. K. [Argonne National Lab. (ANL), Argonne, IL (United States); Grandy, C. [Argonne National Lab. (ANL), Argonne, IL (United States); Natesan, K. [Argonne National Lab. (ANL), Argonne, IL (United States); Sienicki, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Hill, R. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2017-04-20

    The United States Department of Energy (DOE) commissioned the development of technology roadmaps for advanced (non-light water reactor) reactor concepts to help focus research and development funding over the next five years. The roadmaps show the research and development needed to support demonstration of an advanced (non-LWR) concept by the early 2030s, consistent with DOE’s Vision and Strategy for the Development and Deployment of Advanced Reactors. The intent is only to convey the technical steps that would be required to achieve such a goal; the means by which DOE will determine whether to invest in specific tasks will be treated separately. The starting point for the roadmaps is the Technical Readiness Assessment performed as part of an Advanced Test and Demonstration Reactor study released in 2016. The roadmaps were developed based upon a review of technical reports and vendor literature summarizing the technical maturity of each concept and the outstanding research and development needs. Critical path tasks for specific systems were highlighted on the basis of time and resources needed to complete the tasks and the importance of the system to the performance of the reactor concept. The roadmaps are generic, i.e. not specific to a particular vendor’s design but vendor design information may have been used as representative of the concept family. In the event that both near-term and more advanced versions of a concept are being developed, either a single roadmap with multiple branches or separate roadmaps for each version were developed. In each case, roadmaps point to a demonstration reactor (engineering or commercial) and show the activities that must be completed in parallel to support that demonstration in the 2030-2035 window. This report provides the roadmaps for two fast reactor concepts, the Sodium-cooled Fast Reactor (SFR) and the Lead-cooled Fast Reactor (LFR). The SFR technology is mature enough for commercial demonstration by the early 2030s

  16. Technology Selection for Offshore Underwater Small Modular Reactors

    Directory of Open Access Journals (Sweden)

    Koroush Shirvan

    2016-12-01

    Full Text Available This work examines the most viable nuclear technology options for future underwater designs that would meet high safety standards as well as good economic potential, for construction in the 2030–2040 timeframe. The top five concepts selected from a survey of 13 nuclear technologies were compared to a small modular pressurized water reactor (PWR designed with a conventional layout. In order of smallest to largest primary system size where the reactor and all safety systems are contained, the top five designs were: (1 a lead–bismuth fast reactor based on the Russian SVBR-100; (2 a novel organic cooled reactor; (3 an innovative superheated water reactor; (4 a boiling water reactor based on Toshiba's LSBWR; and (5 an integral PWR featuring compact steam generators. A similar study on potential attractive power cycles was also performed. A condensing and recompression supercritical CO2 cycle and a compact steam Rankine cycle were designed. It was found that the hull size required by the reactor, safety systems and power cycle can be significantly reduced (50–80% with the top five designs compared to the conventional PWR. Based on the qualitative economic consideration, the organic cooled reactor and boiling water reactor designs are expected to be the most cost effective options.

  17. Technology selection for offshore underwater small modular reactors

    Energy Technology Data Exchange (ETDEWEB)

    Shivan, Koroush; Ballinger, Ronald; Buongiorno, Jacopo; Forsberg, Charles; Kazimi, Mujid; Todreas, Neil [Dept. of Nuclear Science and Engineering, Massachusetts Institute of Technology, Cambridge (United States)

    2016-12-15

    This work examines the most viable nuclear technology options for future underwater designs that would meet high safety standards as well as good economic potential, for construction in the 2030-2040 time frame. The top five concepts selected from a survey of 13 nuclear technologies were compared to a small modular pressurized water reactor (PWR) designed with a conventional layout. In order of smallest to largest primary system size where the reactor and all safety systems are contained, the top five designs were: (1) a lead-bismuth fast reactor based on the Russian SVBR-100; (2) a novel organic cooled reactor; (3) an innovative superheated water reactor; (4) a boiling water reactor based on Toshiba's LSBWR; and (5) an integral PWR featuring compact steam generators. A similar study on potential attractive power cycles was also performed. A condensing and recompression supercritical CO{sub 2} cycle and a compact steam Rankine cycle were designed. It was found that the hull size required by the reactor, safety systems and power cycle can be significantly reduced (50-80%) with the top five designs compared to the conventional PWR. Based on the qualitative economic consideration, the organic cooled reactor and boiling water reactor designs are expected to be the most cost effective options.

  18. Indian fast reactor technology: Current status and future programme

    Indian Academy of Sciences (India)

    S C Chetal; P Chellapandi

    2013-10-01

    The paper brings out the advantages of fast breeder reactor and importance of developing closed nuclear fuel cycle for the large scale energy production, which is followed by its salient safety features. Further, the current status and future strategy of the fast reactor programme since the inception through 40 MWt/13 MWe Fast Breeder Test Reactor (FBTR), is highlighted. The challenges and achievements in science and technology of FBRs focusing on safety are described with the particular reference to 500 MWe capacity Prototype Fast Breeder Reactor (PFBR), being commissioned at Kalpakkam. Roadmap with comprehensive R&D for the large scale deployment of Sodium Cooled Fast Reactor (SFRs) and timely introduction of metallic fuel reactors with emphasis on breeding gain and enhanced safety are being brought out in this paper.

  19. Technology Selection for Offshore Underwater Small Modular Reactors

    OpenAIRE

    Koroush Shirvan; Ronald Ballinger; Jacopo Buongiorno; Charles Forsberg; Mujid Kazimi; Neil Todreas

    2016-01-01

    This work examines the most viable nuclear technology options for future underwater designs that would meet high safety standards as well as good economic potential, for construction in the 2030–2040 timeframe. The top five concepts selected from a survey of 13 nuclear technologies were compared to a small modular pressurized water reactor (PWR) designed with a conventional layout. In order of smallest to largest primary system size where the reactor and all safety systems are contained, the ...

  20. Future development of the research nuclear reactor IRT-2000 in Sofia

    Energy Technology Data Exchange (ETDEWEB)

    Apostolov, T.G. [Institute for Nuclear Research and Nuclear Energy, BAS, Sofia (Bulgaria)

    1999-07-01

    The present paper presents a short description of the research reactor IRT-2000 Sofia, started in 1961 and operated for 28 years. Some items are considered, connected to the improvements made in the contemporary safety requirements and the unrealized project for modernization to 5 MW. Proposals are considered for reconstruction of reactor site to a 'reactor of low power' for education purposes and as a basis for the country's nuclear technology development. (author)

  1. A new high performance research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Abbate, Pablo M. [INVAP S.E., Bariloche, Rio Negro (Argentina)

    2002-07-01

    A contract for the design, construction and commissioning of the Replacement Research Reactor was signed in July 2000 between Australia authorities and INVAP from Argentina. Since then the detailed design has been completed, an application for a construction license was made in May 2001 and the construction authorisation was issued on 4{sup th} April 2002. This paper explains the safety philosophy embedded into the design together with the approach taken for main elements of the design and their relation to the proposed applications of the reactor. Also information is provided on the suit of neutron beam facilities and irradiation facilities being constructed. Finally it is presented an outline of the project management organisation, project planing and schedule. (author)

  2. Modular Pebble-Bed Reactor Project: Laboratory-Directed Research and Development Program FY 2002 Annual Report

    Energy Technology Data Exchange (ETDEWEB)

    Petti, David Andrew; Dolan, Thomas James; Miller, Gregory Kent; Moore, Richard Leroy; Terry, William Knox; Ougouag, Abderrafi Mohammed-El-Ami; Oh, Chang H; Gougar, Hans D

    2002-11-01

    This report documents the results of our research in FY-02 on pebble-bed reactor technology under our Laboratory Directed Research and Development (LDRD) project entitled the Modular Pebble-Bed Reactor. The MPBR is an advanced reactor concept that can meet the energy and environmental needs of future generations under DOE’s Generation IV initiative. Our work is focused in three areas: neutronics, core design and fuel cycle; reactor safety and thermal hydraulics; and fuel performance.

  3. Modular Pebble-Bed Reactor Project: Laboratory-Directed Research and Development Program FY 2002 Annual Report

    Energy Technology Data Exchange (ETDEWEB)

    Petti, David Andrew; Dolan, Thomas James; Miller, Gregory Kent; Moore, Richard Leroy; Terry, William Knox; Ougouag, Abderrafi Mohammed-El-Ami; Oh, Chang H; Gougar, Hans D

    2002-11-01

    This report documents the results of our research in FY-02 on pebble-bed reactor technology under our Laboratory Directed Research and Development (LDRD) project entitled the Modular Pebble-Bed Reactor. The MPBR is an advanced reactor concept that can meet the energy and environmental needs of future generations under DOE’s Generation IV initiative. Our work is focused in three areas: neutronics, core design and fuel cycle; reactor safety and thermal hydraulics; and fuel performance.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-10-01

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

  5. Safe operation and maintenance of research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Munsorn, S. [Reactor Operation Division, Office of Atomic Energy for Peace, Chatuchak, Bangkok (Thailand)

    1999-10-01

    The first Thai Research Reactor (TRR-1) was established in 1961 at the Office of Atomic Energy for Peace (OAEP), Bangkok. The reactor was light water moderated and cooled, using HEU plate-type with U{sub 3}O{sub 8}- Al fuel meat and swimming pool type. The reactor went first critical on October 27, 1962 and had been licensed to operate at 1 MW (thermal). On June 30, 1975 the reactor was shutdown for modification and the core and control system was disassemble and replaced by that of TRIGA Mark III type while the pool cooling system, irradiation facilities and other were kept. Thus the name TRR-1/M1' has been designed due to this modification the fuel has been changed from HEU plate type to Uranium Zirconium Hydride (UZrH) Low Enrichment Uranium (LEU) which include 4 Fuel Follower Control Rods and 1 Air Follower Control Rod. The TRR-1/M1 went critical on November 7, 1977 and the purpose of the operation are training, isotope production and research. Nowadays the TRR-1/M1 has been operated with core loading No.12 which released power of 1,056 MWD. (as of October 1998). The TRR-1/M1 has been operated at the power of 1.2 MW, three days a week with 34 hours per week, Shut-down on Monday for weekly maintenance and Tuesday for special experiment. The everage energy released is about 40.8 MW-hour per week. Every year, the TRR-1/M1 is shut-down about 2 months between February to March for yearly maintenance. (author)

  6. Simulator platform for fast reactor operation and safety technology demonstration

    Energy Technology Data Exchange (ETDEWEB)

    Vilim, R. B.; Park, Y. S.; Grandy, C.; Belch, H.; Dworzanski, P.; Misterka, J. (Nuclear Engineering Division)

    2012-07-30

    A simulator platform for visualization and demonstration of innovative concepts in fast reactor technology is described. The objective is to make more accessible the workings of fast reactor technology innovations and to do so in a human factors environment that uses state-of-the art visualization technologies. In this work the computer codes in use at Argonne National Laboratory (ANL) for the design of fast reactor systems are being integrated to run on this platform. This includes linking reactor systems codes with mechanical structures codes and using advanced graphics to depict the thermo-hydraulic-structure interactions that give rise to an inherently safe response to upsets. It also includes visualization of mechanical systems operation including advanced concepts that make use of robotics for operations, in-service inspection, and maintenance.

  7. The AFR. An approved network of research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hampel, Gabriele [Mainz Univ. (Germany). Arbeitsgemeinschaft fuer Betriebs- und Sicherheitsfragen an Forschungsreaktoren (AFR)

    2012-10-15

    AFR (Arbeitsgemeinschaft fuer Betriebs- und Sicherheitsfragen an Forschungsreaktoren) is the German acronym for 'Association for Research Reactor Operation and Safety Issues' which was founded in 1959. Reactor managers of European research reactors mainly from the German linguistic area meet regularly for their mutual benefit to exchange experience and knowledge in all areas of operating, managing and utilization of research reactors. In the last 2 years joint meetings were held together with the French association of research reactors CER (Club d'Exploitants des Reacteurs). In this contribution the AFR, its members, work and aims as well as the French partner CER are presented. (orig.)

  8. An assessment of space reactor technology needs and recommendations for development

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, A.C. [Sandia National Labs., Albuquerque, NM (United States); Wiley, R.L. [Consultant, Columbia, MD (United States)

    1995-11-01

    In order to provide a strategy for space reactor technology development, the Defense Nuclear Agency (DNA) has authorized a brief review of potential national needs that may be addressed by space reactor systems. a systematic approach was used to explore needs at several levels that are increasingly specific. Level 0 -- general trends and issues; Level 1 -- generic space capabilities to address trends; Level 2 -- requirements to support capabilities; Level 3 -- system types capable of meeting requirements; Level 4 --generic reactor system types; and Level 5 -- specific baseline systems. Using these findings, a strategy was developed to support important space reactor technologies within a limited budget. A preliminary evaluation identified key technical issues and provide a prioritized set of candidate research projects. The evaluation of issues and the recommended research projects are presented in a companion paper.

  9. Research on Power Ramp Testing Method for PWR Fuel Rod at Research Reactor

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    In order to develop high performance fuel assembly for domestic nuclear power plant, it is necessary to master some fundamental test technology. So the research on the power ramp testing methods is proposed. A tentative power ramp test for short PWR fuel rod has been conducted at the heavy water research reactor (HWRR) in China Institute of Atomic Energy (CIAE) in May of 2001. The in-pile test rig was placed into the central channel of the reactor . The test rig consists of pressure pipe assembly, thimble, solid neutron absorbing screen and its driving parts, etc.. The test

  10. Monte Carlo modelling of TRIGA research reactor

    Science.gov (United States)

    El Bakkari, B.; Nacir, B.; El Bardouni, T.; El Younoussi, C.; Merroun, O.; Htet, A.; Boulaich, Y.; Zoubair, M.; Boukhal, H.; Chakir, M.

    2010-10-01

    The Moroccan 2 MW TRIGA MARK II research reactor at Centre des Etudes Nucléaires de la Maâmora (CENM) achieved initial criticality on May 2, 2007. The reactor is designed to effectively implement the various fields of basic nuclear research, manpower training, and production of radioisotopes for their use in agriculture, industry, and medicine. This study deals with the neutronic analysis of the 2-MW TRIGA MARK II research reactor at CENM and validation of the results by comparisons with the experimental, operational, and available final safety analysis report (FSAR) values. The study was prepared in collaboration between the Laboratory of Radiation and Nuclear Systems (ERSN-LMR) from Faculty of Sciences of Tetuan (Morocco) and CENM. The 3-D continuous energy Monte Carlo code MCNP (version 5) was used to develop a versatile and accurate full model of the TRIGA core. The model represents in detailed all components of the core with literally no physical approximation. Continuous energy cross-section data from the more recent nuclear data evaluations (ENDF/B-VI.8, ENDF/B-VII.0, JEFF-3.1, and JENDL-3.3) as well as S( α, β) thermal neutron scattering functions distributed with the MCNP code were used. The cross-section libraries were generated by using the NJOY99 system updated to its more recent patch file "up259". The consistency and accuracy of both the Monte Carlo simulation and neutron transport physics were established by benchmarking the TRIGA experiments. Core excess reactivity, total and integral control rods worth as well as power peaking factors were used in the validation process. Results of calculations are analysed and discussed.

  11. Application Technology Research Unit

    Data.gov (United States)

    Federal Laboratory Consortium — To conduct fundamental and developmental research on new and improved application technologies to protect floricultural, nursery, landscape, turf, horticultural, and...

  12. A Study on the demands of research reactors and considerations for an export

    Energy Technology Data Exchange (ETDEWEB)

    Park, Cheol; Lee, Young Jun

    2008-11-15

    Among around 240 research reactors in operation over the world, around 80% have been operated for more than 20 years and 65% for more than 30 years. Hence the number of operable reactors is expected, between 2010 and 2020, to be reduced to 1/3 of the present situation if the lifetime of a research reactor is assumed to be 40 years. However, considering the recent re-highlighting of nuclear energy as a practical mass energy source and the contributions to the overall areas of science and technology, the demands for constructing a new research reactor and replacing the existing research reactors will be increased in the near future. On the other hand, vendors which participate in providing research reactors are not few, and AREVA in France and INVAP in Argentina are example of them in a positive position. Japan and Russia are regarded as potential competitors, but they do not actively appear in the market so far. Comparing those competitors with Korea, we have weak points regarding experiences on exports and the organizational systems as an integrated vendor. But we may have a competitiveness by grafting our experiences on the development of nuclear power technology and the construction and operation of the HANARO. In this report, the future potential demands for research reactors and the related considerations for exports have been surveyed and described, particularly, centering around the Netherlands, Vietnam and Thailand that are countries which may construct research reactors in the near future. Considerations for exporting a research reactor have been categorized into two groups of technical and nontechnical items. From a technical point of view, the issues on fuel and reactor type, design data and design ability, design codes, and technology property rights have been reviewed. For the non-technical items, an integrated project system, reasonable estimate of demands, social and economic conditions for potential demand countries, MOU status, nuclear non

  13. Generation IV Reactors Integrated Materials Technology Program Plan: Focus on Very High Temperature Reactor Materials

    Energy Technology Data Exchange (ETDEWEB)

    Corwin, William R [ORNL; Burchell, Timothy D [ORNL; Katoh, Yutai [ORNL; McGreevy, Timothy E [ORNL; Nanstad, Randy K [ORNL; Ren, Weiju [ORNL; Snead, Lance Lewis [ORNL; Wilson, Dane F [ORNL

    2008-08-01

    Since 2002, the Department of Energy's (DOE's) Generation IV Nuclear Energy Systems (Gen IV) Program has addressed the research and development (R&D) necessary to support next-generation nuclear energy systems. The six most promising systems identified for next-generation nuclear energy are described within this roadmap. Two employ a thermal neutron spectrum with coolants and temperatures that enable hydrogen or electricity production with high efficiency (the Supercritical Water Reactor-SCWR and the Very High Temperature Reactor-VHTR). Three employ a fast neutron spectrum to enable more effective management of actinides through recycling of most components in the discharged fuel (the Gas-cooled Fast Reactor-GFR, the Lead-cooled Fast Reactor-LFR, and the Sodium-cooled Fast Reactor-SFR). The Molten Salt Reactor (MSR) employs a circulating liquid fuel mixture that offers considerable flexibility for recycling actinides and may provide an alternative to accelerator-driven systems. At the inception of DOE's Gen IV program, it was decided to significantly pursue five of the six concepts identified in the Gen IV roadmap to determine which of them was most appropriate to meet the needs of future U.S. nuclear power generation. In particular, evaluation of the highly efficient thermal SCWR and VHTR reactors was initiated primarily for energy production, and evaluation of the three fast reactor concepts, SFR, LFR, and GFR, was begun to assess viability for both energy production and their potential contribution to closing the fuel cycle. Within the Gen IV Program itself, only the VHTR class of reactors was selected for continued development. Hence, this document will address the multiple activities under the Gen IV program that contribute to the development of the VHTR. A few major technologies have been recognized by DOE as necessary to enable the deployment of the next generation of advanced nuclear reactors, including the development and qualification of

  14. Architecture dependent availability analysis of RPS for Research Reactor Applications

    Energy Technology Data Exchange (ETDEWEB)

    Rahman, Khalilur; Heo, Gyunyoung [Kyung Hee Univ., Yongin (Korea, Republic of); Son, Hanseong [Joongbu Univ., Geumsan (Korea, Republic of); Kim, Youngki; Park, Jaekwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    The research reactors are categorized into two broad categories, Low power research reactors and medium to high power research reactors. According to IAEA TECDOC-1234, Research reactors with 0.250- 2.0 MW power rating or 2.5-10 Χ 10{sup 11} η/cm{sup 2}. s flux are termed low power reactor whereas research reactors ranging from 2-10 MW power rating or 0.1-10 Χ 10{sup 13} η/cm{sup 2}. s are considered as Medium to High power research reactors. Some other standards (IAEA NP-T-5.1) define multipurpose research reactor ranging from power few hundred KW to 10 MW as low power research reactor. The aim of this research, in this article, was to identify a configuration of architecture which gives highest availability with maintaining low cost of manufacturing. In this regard, two configurations of a single channel of RPS are formulated in the current article and their fault trees were developed using AIMS PSA software to get the unavailability. This is a starting point of attempt towards the standardization of I and C architecture for low and medium power research reactors.

  15. Space-reactor electric systems: subsystem technology assessment

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, R.V.; Bost, D.; Determan, W.R.

    1983-03-29

    This report documents the subsystem technology assessment. For the purpose of this report, five subsystems were defined for a space reactor electric system, and the report is organized around these subsystems: reactor; shielding; primary heat transport; power conversion and processing; and heat rejection. The purpose of the assessment was to determine the current technology status and the technology potentials for different types of the five subsystems. The cost and schedule needed to develop these potentials were estimated, and sets of development-compatible subsystems were identified.

  16. Modular fabrication and characterization of complex silicon carbide composite structures Advanced Reactor Technologies (ART) Research Final Report (Feb 2015 – May 2017)

    Energy Technology Data Exchange (ETDEWEB)

    Khalifa, Hesham [General Atomics, San Diego, CA (United States)

    2017-08-03

    Advanced ceramic materials exhibit properties that enable safety and fuel cycle efficiency improvements in advanced nuclear reactors. In order to fully exploit these desirable properties, new processing techniques are required to produce the complex geometries inherent to nuclear fuel assemblies and support structures. Through this project, the state of complex SiC-SiC composite fabrication for nuclear components has advanced significantly. New methods to produce complex SiC-SiC composite structures have been demonstrated in the form factors needed for in-core structural components in advanced high temperature nuclear reactors. Advanced characterization techniques have been employed to demonstrate that these complex SiC-SiC composite structures provide the strength, toughness and hermeticity required for service in harsh reactor conditions. The complex structures produced in this project represent a significant step forward in leveraging the excellent high temperature strength, resistance to neutron induced damage, and low neutron cross section of silicon carbide in nuclear applications.

  17. A Preliminary Analysis of Reactor Performance Test (LOEP) for a Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyeonil; Park, Su-Ki [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    The final phase of commissioning is reactor performance test, which is to prove the integrated performance and safety of the research reactor at full power with fuel loaded such as neutron power calibration, Control Absorber Rod/Second Shutdown Rod drop time, InC function test, Criticality, Rod worth, Core heat removal with natural mechanism, and so forth. The last test will be safety-related one to assure the result of the safety analysis of the research reactor is marginal enough to be sure about the nuclear safety by showing the reactor satisfies the acceptance criteria of the safety functions such as for reactivity control, maintenance of auxiliaries, reactor pool water inventory control, core heat removal, and confinement isolation. After all, the fuel integrity will be ensured by verifying there is no meaningful change in the radiation levels. To confirm the performance of safety equipment, loss of normal electric power (LOEP), possibly categorized as Anticipated Operational Occurrence (AOO), is selected as a key experiment to figure out how safe the research reactor is before turning over the research reactor to the owner. This paper presents a preliminary analysis of the reactor performance test (LOEP) for a research reactor. The results showed how different the transient between conservative estimate and best estimate will look. Preliminary analyses have shown all probable thermal-hydraulic transient behavior of importance as to opening of flap valve, minimum critical heat flux ratio, the change of flow direction, and important values of thermal-hydraulic parameters.

  18. Meteodiffusive Characterization of Algiers' Nuclear Research Reactor

    Directory of Open Access Journals (Sweden)

    Mourad Messaci

    2007-01-01

    Full Text Available In the framework of the environmental impact studies of the nuclear research reactor of Algiers, we will present the work related to the atmospheric dispersion of releases due to the installation in normal operation, which dealt with the assessment of spatial distribution of yearly average values of atmospheric dilution factor. The aim of this work is a characterization of the site in terms of diffusivity, which is basic for the radiological impact evaluation of the reactor. The meteorological statistics result from the National Office of Meteorology and concern 15 years of hourly records. According to the nature and features of these data, a Gaussian-type model with wind direction sectors was used. Values of wind speed at release height were estimated from measurement values at 10 m from ground. For the assessment of vertical dispersion coefficient, we used Briggs' formulas related to a sampling time of one hour. Areas of maximum impact were delimited and points of highest concentration within these zones were identified.

  19. Design characteristics and requirements of irradiation holes for research reactor experimental facilities

    Energy Technology Data Exchange (ETDEWEB)

    Park, Cheol; Lee, B. C.; Chae, H. T.; Lee, C. S.; Seo, C. G

    2003-07-01

    In order to be helpful for the design of a new research reactor with high performance, are summarized the applications of research reactors in various fields and the design characteristics of experimental facility such as vertical irradiation holes and beam tubes. Basic requirements of such experimental facilities are also described. Research reactor has been widely utilized in various fields such as industry, engineering, medicine, life science, environment etc., and now the application fields are gradually being expanded together with the development of technology. Looking into the research reactors which are recently constructed or in plan, it seems that to develop a multi-purpose research reactor with intensive neutron beam research capability has become tendency. In the layout of the experimental facilities, the number and configuration of irradiation and beam holes should be optimized to meet required test conditions such as neutron flux at the early design stage. But, basically high neutron flux is required to perform experiments efficiently. In this aspect, neutron flux is regarded as one of important parameters to judge the degree of research reactor performance. One of main information for a new research reactor design is utilization demands and requirements of experimental holes. So basic requirements which should be considered in a new research reactor design were summarized from the survey of experimental facilities characteristics of various research reactors with around 20 MW thermal power and the experiences of HANARO utilization. Also is suggested an example of the requirements of experimental holes such as size, number and neutron flux, which are thought as minimum, in a new research reactor for exporting to developing countries such as Vietnam.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-07-01

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

  1. Novel Technology for Phenol Wastewater Treatment Using Electrochemical Reactor

    Directory of Open Access Journals (Sweden)

    Yuncheng Xie

    2015-01-01

    Full Text Available There are various electrochemical approaches to save energy, mostly by means of equipment improvement coupled with other water treatment technologies. Replacement of DC power with pulse power, modified reactor coupled with photocatalysis can decrease cost. But more or less additional input is developed, or infrastructure has to be replaced. In this paper, an N-Step electrochemical reactor, based on stage reaction modeling, is put forward. On the basis of not changing equipment investment and by adjustment of the operating current density at different levels, power consumption decreases. This model develops a foundation of electrochemical water treatment technology for the engineering application.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-10-03

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

  3. Materials and Components Technology Division research summary, 1991

    Energy Technology Data Exchange (ETDEWEB)

    1991-04-01

    This division has the purpose of providing a R and D capability for design, fabrication, and testing of high-reliability materials, components, and instrumentation. Current divisional programs are in support of the Integral Fast Reactor, life extension for light water reactors, fuels development for the new production reactor and research and test reactors, fusion reactor first-wall and blanket technology, safe shipment of hazardous materials, fluid mechanics/materials/instrumentation for fossile energy systems, and energy conservation and renewables (including tribology, high- temperature superconductivity). Separate abstracts have been prepared for the data base.

  4. Materials and Components Technology Division research summary, 1991

    Energy Technology Data Exchange (ETDEWEB)

    1991-04-01

    This division has the purpose of providing a R and D capability for design, fabrication, and testing of high-reliability materials, components, and instrumentation. Current divisional programs are in support of the Integral Fast Reactor, life extension for light water reactors, fuels development for the new production reactor and research and test reactors, fusion reactor first-wall and blanket technology, safe shipment of hazardous materials, fluid mechanics/materials/instrumentation for fossile energy systems, and energy conservation and renewables (including tribology, high- temperature superconductivity). Separate abstracts have been prepared for the data base.

  5. Reactors

    CERN Document Server

    International Electrotechnical Commission. Geneva

    1988-01-01

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

  6. Initial decommissioning planning for the Budapest research reactor

    OpenAIRE

    Toth Gabor

    2011-01-01

    The Budapest Research Reactor is the first nuclear research facility in Hungary. The reactor is to remain in operation for at least another 13 years. At the same time, the development of a decommissioning plan is a mandatory requirement under national legislation. The present paper describes the current status of decommissioning planning which is aimed at a timely preparation for the forthcoming decommissioning of the reactor.

  7. Initial decommissioning planning for the Budapest research reactor

    Directory of Open Access Journals (Sweden)

    Toth Gabor

    2011-01-01

    Full Text Available The Budapest Research Reactor is the first nuclear research facility in Hungary. The reactor is to remain in operation for at least another 13 years. At the same time, the development of a decommissioning plan is a mandatory requirement under national legislation. The present paper describes the current status of decommissioning planning which is aimed at a timely preparation for the forthcoming decommissioning of the reactor.

  8. Materials science research for sodium cooled fast reactors

    Indian Academy of Sciences (India)

    Baldev Raj

    2009-06-01

    The paper gives an insight into basic as well as applied research being carried out at the Indira Gandhi Centre for Atomic Research for the development of advanced materials for sodium cooled fast reactors towards extending the life of reactors to nearly 100 years and the burnup of fuel to 2,00,000 MWd/t with an objective of providing fast reactor electricity at an affordable and competitive price.

  9. Research and Technology 1980

    Science.gov (United States)

    1980-01-01

    Topics are divided into three major areas: Earth resources, advanced development, and technology transfer. Topics include: aerial color infrared photography, fiber optics, lightning research, soil mechanics, corrosion prevention, image processing, and communication systems development.

  10. Contribution of CAD and PLM Research Reactors Design and Construction

    Energy Technology Data Exchange (ETDEWEB)

    Bonnetain, Xavier [AREVA TA, Paris (France)

    2013-07-01

    As all the reactors, the main stakes in the engineering of design and construction of the research reactors consist of the management and sharing of the technical data, the functional, physical and contractual interfaces data between the various contributors on the whole designs and construction cycle project. For 40 years, AREVA TA designs and builds reactors. Computer Aided Design (CAD) tools were introduced for 30 years into the engineering processes of AREVA TA, completed for 15 years by Product Lifecycle Management (PLM) tools. For 15 years AREVA TA pursues the integration since the feasibility of its newest Information Technologies (IT). In the first part, the paper presents IN the second part, the paper presents how the schematics and CAD tools support the engineering processes during the different phases of the project. CAD was used during the studies and now supports the management of the layout and design studies, including interfaces between suppliers, up to the constitution of the as built CAD mock-up. In the third part, the paper presents the relations between the various tools and the PLM solution implemented by AREVA TA to ensure the consistency between all tools and data for the benefit of the project.

  11. Recycling of the Electronic Waste Applying the Plasma Reactor Technology

    Science.gov (United States)

    Lázár, Marián; Jasminská, Natália; Čarnogurská, Mária; Dobáková, Romana

    2016-12-01

    The following paper discusses a high-temperature gasification process and melting of electronic components and computer equipment using plasma reactor technology. It analyses the marginal conditions of batch processing, as well as the formation of solid products which result from the procedure of waste processing. Attention is also paid to the impact of the emerging products on the environment.

  12. Recycling of the Electronic Waste Applying the Plasma Reactor Technology

    Directory of Open Access Journals (Sweden)

    Lázár Marián

    2016-12-01

    Full Text Available The following paper discusses a high-temperature gasification process and melting of electronic components and computer equipment using plasma reactor technology. It analyses the marginal conditions of batch processing, as well as the formation of solid products which result from the procedure of waste processing. Attention is also paid to the impact of the emerging products on the environment.

  13. Updated Generation IV Reactors Integrated Materials Technology Program Plan, Revision 2

    Energy Technology Data Exchange (ETDEWEB)

    Corwin, William R [ORNL; Burchell, Timothy D [ORNL; Halsey, William [Lawrence Livermore National Laboratory (LLNL); Hayner, George [Idaho National Laboratory (INL); Katoh, Yutai [ORNL; Klett, James William [ORNL; McGreevy, Timothy E [ORNL; Nanstad, Randy K [ORNL; Ren, Weiju [ORNL; Snead, Lance Lewis [ORNL; Stoller, Roger E [ORNL; Wilson, Dane F [ORNL

    2005-12-01

    The Department of Energy's (DOE's) Generation IV Nuclear Energy Systems Program will address the research and development (R&D) necessary to support next-generation nuclear energy systems. Such R&D will be guided by the technology roadmap developed for the Generation IV International Forum (GIF) over two years with the participation of over 100 experts from the GIF countries. The roadmap evaluated over 100 future systems proposed by researchers around the world. The scope of the R&D described in the roadmap covers the six most promising Generation IV systems. The effort ended in December 2002 with the issue of the final Generation IV Technology Roadmap [1.1]. The six most promising systems identified for next generation nuclear energy are described within the roadmap. Two employ a thermal neutron spectrum with coolants and temperatures that enable hydrogen or electricity production with high efficiency (the Supercritical Water Reactor - SCWR and the Very High Temperature Reactor - VHTR). Three employ a fast neutron spectrum to enable more effective management of actinides through recycling of most components in the discharged fuel (the Gas-cooled Fast Reactor - GFR, the Lead-cooled Fast Reactor - LFR, and the Sodium-cooled Fast Reactor - SFR). The Molten Salt Reactor (MSR) employs a circulating liquid fuel mixture that offers considerable flexibility for recycling actinides, and may provide an alternative to accelerator-driven systems. A few major technologies have been recognized by DOE as necessary to enable the deployment of the next generation of advanced nuclear reactors, including the development and qualification of the structural materials needed to ensure their safe and reliable operation. Accordingly, DOE has identified materials as one of the focus areas for Gen IV technology development.

  14. Sodium fast reactor fuels and materials : research needs.

    Energy Technology Data Exchange (ETDEWEB)

    Denman, Matthew R.; Porter, Douglas (Idaho National Laboratory, Idaho Falls, ID); Wright, Art (Argonne National Laboratory Argonne, IL); Lambert, John (Argonne National Laboratory Argonne, IL); Hayes, Steven (Idaho National Laboratory, Idaho Falls, ID); Natesan, Ken (Argonne National Laboratory Argonne, IL); Ott, Larry J. (Oak Ridge National Laboratory, Oak Ridge, TN); Garner, Frank (Radiation Effects Consulting. Richland, WA); Walters, Leon (Advanced Reactor Concepts, Idaho Falls, ID); Yacout, Abdellatif (Argonne National Laboratory Argonne, IL)

    2011-09-01

    An expert panel was assembled to identify gaps in fuels and materials research prior to licensing sodium cooled fast reactor (SFR) design. The expert panel considered both metal and oxide fuels, various cladding and duct materials, structural materials, fuel performance codes, fabrication capability and records, and transient behavior of fuel types. A methodology was developed to rate the relative importance of phenomena and properties both as to importance to a regulatory body and the maturity of the technology base. The technology base for fuels and cladding was divided into three regimes: information of high maturity under conservative operating conditions, information of low maturity under more aggressive operating conditions, and future design expectations where meager data exist.

  15. Development of a decommissioning strategy for the MR research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bylkin, Boris; Gorlinsky, Yury; Kolyadin, Vyacheslav; Pavlenko, Vitaly [RRC Kurchatov Institute, Moscow (Russian Federation); Craig, David; Fecitt, Lorna [NUKEM Limited, Dounreay (United Kingdom); Harman, Neil; Jackson, Roger [Serco Technical and Assurance Services, Warrington (United Kingdom); Lobach, Yury [Inst. for Nuclear Research of NASU, Kiev (Ukraine)

    2010-03-15

    A description of the selected decommissioning strategy for the research reactor MR at the site of the Kurchatov Institute in Moscow is presented. The MR reactor hall is planned to be used as a temporary fuel store for the other research reactors on the site. On the basis of the site-specific conditions and over-all decommissioning goals, it was identified that phased immediate decommissioning is the preferable option. The current status of the reactor, expected final conditions and the sequence of decommissioning works are shown. (orig.)

  16. Proceedings of the sixth Asian symposium on research reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-08-01

    The symposium consisted of 16 sessions with 58 submitted papers. Major fields were: (1) status and future plan of research and testing reactors, (2) operating experiences, (3) design and modification of the facility, and reactor fuels, (4) irradiation studies, (5) irradiation facilities, (6) reactor characteristics and instrumentation, and (7) neutron beam utilization. Panel discussion on the 'New Trends on Application of Research and Test Reactors' was also held at the last of the symposium. About 180 people participated from China, Korea, Indonesia, Thailand, Bangladesh, Vietnam, Chinese Taipei, Belgium, France, USA, Japan and IAEA. The 58 of the presented papers are indexed individually. (J.P.N.)

  17. Reactor pressure vessel structural integrity research

    Energy Technology Data Exchange (ETDEWEB)

    Pennell, W.E.; Corwin, W.R. [Oak Ridge National Lab., TN (United States)

    1995-04-01

    Development continues on the technology used to assess the safety of irradiation-embrittled nuclear reactor pressure vessels (RPVs) containing flaws. Fracture mechanics tests on RPV steel, coupled with detailed elastic-plastic finite-element analyses of the crack-tip stress fields, have shown that (1) constraint relaxation at the crack tip of shallows surface flaws results in increased data scatter but no increase in the lower-bound fracture toughness, (2) the nil ductility temperature (NDT) performs better than the reference temperature for nil ductility transition (RT{sub NDT}) as a normalizing parameter for shallow-flaw fracture toughness data, (3) biaxial loading can reduce the shallow-flaw fracture toughness, (4) stress-based dual-parameter fracture toughness correlations cannot predict the effect of biaxial loading on a shallow-flaw fracture toughness because in-plane stresses at the crack tip are not influenced by biaxial loading, and (5) an implicit strain-based dual-parameter fracture toughness correlation can predict the effect of biaxial loading on shallow-flaw fracture toughness. Experimental irradiation investigations have shown that (1) the irradiation-induced shift in Charpy V-notch vs temperature behavior may not be adequate to conservatively assess fracture toughness shifts due to embrittlement, and (2) the wide global variations of initial chemistry and fracture properties of a nominally uniform material within a pressure vessel may confound accurate integrity assessments that require baseline properties.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-10-01

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

  19. Proceedings of the first symposium on utilization of research reactors and JMTR

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-03-01

    The first symposium on utilization of research reactors (JRR-2, JRR-3M, JRR-4) and Japan Materials Testing Reactor (JMTR) in JAERI was held from September 29th to 30th, 1997 at Sannomaru Hotel, Mito. The purpose of this symposium is to announce contribution to progress of scientific technology as well as to promote future utilization of the research reactors and JMTR. During the symposium, 16 reports were presented on nuclear fuel and material, neutron beam experiment, medical irradiation, radioisotope production and neutron activation analysis. The present status of the research reactors and JMTR were also reported. The special lecture titled `JRR-2 and Medical Irradiation` was given by Mr. Nakamura, former editorial writer of Yomiuri. Finally, panel discussion was carried on `The Role of Research Reactors and JMTR in Scientific Technology for the future` actively by the participants and experts in every field of research reactor utilization. 250 people participated in this symposium from universities, national research institutes, private corporations and JAERI. This proceedings briefly summarizes 16 reports, the content of panel discussion and so forth. (J.P.N.)

  20. Experimental research in neutron physic and thermal-hydraulic at the CDTN Triga reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mesquita, Amir Z.; Souza, Rose Mary G.P.; Ferreira, Andrea V.; Pinto, Antonio J.; Costa, Antonio C.L.; Rezende, Hugo C., E-mail: amir@cdtn.b, E-mail: souzarm@cdtn.b, E-mail: avf@cdtn.b, E-mail: ajp@cdtn.b, E-mail: aclc@cdtn.b, E-mail: hcr@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-07-01

    The IPR-R1 TRIGA (Training, Research, Isotopes production, General Atomics) at Nuclear Technology Development Center (CDTN) is a pool type reactor cooled by natural circulation of light water and an open surface. TRIGA reactors, developed by General Atomics (GA), are the most widely used research reactor in the world and characterized by inherent safety. The IPR-R1 is the only Brazilian nuclear research reactor available and able to perform experiments in which interaction between neutronic and thermal-hydraulic areas occurs. The IPR-R1 has started up on November 11th, 1960. At that time the maximum thermal power was 30 kW. The present forced cooling system was built in the 70th and the power was upgraded to 100 kW. Recently the core configuration and instrumentation was upgraded again to 250 kW at steady state, and is awaiting the license of CNEN to operate definitely at this new power. This paper describes the experimental research project carried out in the IPR-R1 reactor that has as objective evaluate the behaviour of the reactor operational parameters, and mainly to investigate the influence of temperature on the neutronic variables. The research was supported by Research Support Foundation of the State of Minas Gerais (FAPEMIG) and Brazilian Council for Scientific and Technological Development (CNPq). The research project meets the recommendations of the IAEA, for safety, modernization and development of strategic plan for research reactors utilization. This work is in line with the strategic objectives of Brazil, which aims to design and construct the Brazilian Multipurpose research Reactor (RMB). (author)

  1. Detailed Burnup Calculations for Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Leszczynski, F. [Centro Atomico Bariloche (CNEA), 8400 S. C. de Bariloche (Argentina)

    2011-07-01

    A general method (RRMCQ) has been developed by introducing a microscopic burn up scheme which uses the Monte Carlo calculated spatial power distribution of a research reactor core and a depletion code for burn up calculations, as a basis for solving nuclide material balance equations for each spatial region in which the system is divided. Continuous energy dependent cross-section libraries and full 3D geometry of the system is input for the calculations. The resulting predictions for the system at successive burn up time steps are thus based on a calculation route where both geometry and cross-sections are accurately represented, without geometry simplifications and with continuous energy data. The main advantage of this method over the classical deterministic methods currently used is that RRMCQ System is a direct 3D method without the limitations and errors introduced on the homogenization of geometry and condensation of energy of deterministic methods. The Monte Carlo and burn up codes adopted until now are the widely used MCNP5 and ORIGEN2 codes, but other codes can be used also. For using this method, there is a need of a well-known set of nuclear data for isotopes involved in burn up chains, including burnable poisons, fission products and actinides. For fixing the data to be included on this set, a study of the present status of nuclear data is performed, as part of the development of RRMCQ method. This study begins with a review of the available cross-section data of isotopes involved in burn up chains for research nuclear reactors. The main data needs for burn up calculations are neutron cross-sections, decay constants, branching ratios, fission energy and yields. The present work includes results of selected experimental benchmarks and conclusions about the sensitivity of different sets of cross-section data for burn up calculations, using some of the main available evaluated nuclear data files. Basically, the RRMCQ detailed burn up method includes four

  2. Study on Modeling Technology in Digital Reactor System

    Institute of Scientific and Technical Information of China (English)

    刘晓平; 罗月童; 童莉莉

    2004-01-01

    Modeling is the kernel part of a digital reactor system. As an extensible platform for reactor conceptual design, it is very important to study modeling technology and develop some kind of tools to speed up preparation of all classical computing models. This paper introduces the background of the project and basic conception of digital reactor. MCAM is taken as an example for modeling and its related technologies used are given. It is an interface program for MCNP geometry model developed by FDS team (ASIPP & HUT), and designed to run on windows system. MCAM aims at utilizing CAD technology to facilitate creation of MCNP geometry model. There have been two ways for MCAM to utilize CAD technology:(1) Making use of user interface technology in aid of generation of MCNP geometry model;(2) Making use of existing 3D CAD model to accelerate creation of MCNP geometry model. This paper gives an overview of MCAM's major function. At last, several examples are given to demonstrate MCAM's various capabilities.

  3. Electro-catalytic membrane reactors and the development of bipolar membrane technology

    NARCIS (Netherlands)

    Balster, J.; Stamatialis, D.F.; Wessling, M.

    2004-01-01

    Membrane reactors are currently under extensive research and development. Hardly any concept, however, is realized yet in practice. Frequently, forgotten as membrane reactors are electro-catalytic membrane reactors where electrodes perform chemical conversations and membranes separate the locations

  4. Membrane reactor technology for C5/C6 hydroisomerization.

    Science.gov (United States)

    McLeary, E E; Buijsse, E J W; Gora, L; Jansen, J C; Maschmeyer, Th

    2005-04-15

    In this paper, we propose an improved hydroisomerization process, making use of membrane reactor technology. Linear alkanes are selectively supplied from a hydrocarbon feed (consisting of branched and linear alkanes) through an inert tubular membrane to a packed bed of catalyst. The results indicate that n-, mono- and di-branched components in a gas mixture can be separated with a selectivity factor of greater than 20 with a zeolite membrane under dedicated parameter settings. The RON-value of the product was calculated to be as high as 90 in a single pass reactor, which is 50 points higher than the feed value. The flux through the membrane could be optimized to give a STY/ATY ratio for the reactor of 877 m-1, which falls within the limits of technical feasibility.

  5. A study on future nuclear reactor technology and development strategy

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S. Y.; Kim, S. H.; Sohn, D. S.; Suk, S. D.; Zee, S. K.; Yang, M. H.; Kim, H. J.; Park, W. S

    2000-12-01

    Development of nuclear reactor and fuel cycle technology for future is essential to meet the current issues such as enhancement of nuclear power reactor safety, economically competitive with gas turbine power generation, less production of radioactive waste, proliferation resistant fuel cycle, and public acceptance in consideration of lack of energy resources in the nuclear countries worldwide as well as in Korea. This report deals with as follows, 1) Review the world energy demand and supply perspective and analyse nature of energy and sustainable development to set-up nuclear policy in Korea 2) Recaptitulate the current long term nuclear R and D activities 3) Review nuclear R and D activities and programs of USA, Japan, France, Russia, international organizations such as IAEA, OECD/NEA 4) Recommend development directions of nuclear reactors and fuels.

  6. Reactor in search for money. Cooled neutrons for unique research; Reactor zoekt geld. Gekoelde neutronen maken onderzoek mogelijk

    Energy Technology Data Exchange (ETDEWEB)

    Verdult, E.

    2010-12-17

    The modernization of radiation research at the Delft University of Technology depends on subsidies for new instruments. OYSTER (Optimised Yield for Science, Technology and Education of Radiation) is the plan of the Reactor Institute Delft (RID) to realize such a modernization. The article comprises detailed drawings of the inside of the reactor and illustrates the CNIF (Cold Neutron Irradiation Facility) to fight cancer and POSH-PALS (Positron Annihilation Lifetime Spectrometry) to visualize the atomic structure of materials. [Dutch] De modernisering van het stralingsonderzoek aan de Technische Universiteit Delft staat of valt met subsidie voor een nieuwe opzet van het instrumentarium. OYSTER (Optimised Yield for Science, Technology and Education of Radiation) is het plan van het Reactor Instituut Delft (RID) om de installatie te moderniseren. Het artikel bevat gedetailleerde tekeningen van de binnenkant van de reactor en illustreert de CNIF (Cold Neutron Irradiation Facility) voor de bestrijding van kanker en POSH-PALS (Positron Annihilation Lifetime Spectrometry) waarmee de structuur van materialen op atoomniveau inzichtelijk kan worden gemaakt.

  7. The present situations and perspectives on utilization of research reactors in Thailand

    Science.gov (United States)

    Chongkum, Somporn

    2002-01-01

    The Thai Research Reactor 1/Modification 1, a TRIGA Mark III reactor, went critical on November 7, 1977. It has been playing a central role in the development of both Office of Atomic Energy for Peace (OAEP) and nuclear application in Thailand. It has a maximum power of 2 MW (thermal) at steady state and a pulsing capacity of 2000 MW. The highest thermal neutron flux at a central thimber is 1×10 13 n/cm 2/s, which is extensively utilized for radioisotope production, neutron activation analysis and neutron beam experiments, i.e. neutron scattering, prompt gamma analysis and neutron radiography. Following the nuclear technological development, the OAEP is in the process of establishing the Ongkharak Nuclear Research Center (ONRC). The center is being built in Nakhon Nayok province, 60 km northeast of Bangkok. The centerpiece of the ONRC is a multipurpose 10 MW TRIGA research reactor. Facilities are included for the production of radioisotopes for medicine, industry and agriculture, neutron transmutation doping of silicon, and neutron capture therapy. The neutron beam facilities will also be utilized for applied research and technology development as well as training in reactor operations, performance of experiments and reactor physics. This paper describes a recent program of utilization as well as a new research reactor for enlarging the perspectives of its utilization in the future.

  8. Sodium fast reactor safety and licensing research plan. Volume I.

    Energy Technology Data Exchange (ETDEWEB)

    Sofu, Tanju (Argonne National Laboratory, Argonne, IL); LaChance, Jeffrey L.; Bari, R. (Brokhaven National Laboratory Upton, NY); Wigeland, Roald (Idaho National Laboratory, Idaho Falls, ID); Denman, Matthew R.; Flanagan, George F. (Oak Ridge National Laboratory, Oak Ridge, TN)

    2012-05-01

    This report proposes potential research priorities for the Department of Energy (DOE) with the intent of improving the licensability of the Sodium Fast Reactor (SFR). In support of this project, five panels were tasked with identifying potential safety-related gaps in available information, data, and models needed to support the licensing of a SFR. The areas examined were sodium technology, accident sequences and initiators, source term characterization, codes and methods, and fuels and materials. It is the intent of this report to utilize a structured and transparent process that incorporates feedback from all interested stakeholders to suggest future funding priorities for the SFR research and development. While numerous gaps were identified, two cross-cutting gaps related to knowledge preservation were agreed upon by all panels and should be addressed in the near future. The first gap is a need to re-evaluate the current procedures for removing the Applied Technology designation from old documents. The second cross-cutting gap is the need for a robust Knowledge Management and Preservation system in all SFR research areas. Closure of these and the other identified gaps will require both a reprioritization of funding within DOE as well as a re-evaluation of existing bureaucratic procedures within the DOE associated with Applied Technology and Knowledge Management.

  9. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2001-04-01

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

  10. IAEA Assistance in the development of new research reactor projects

    Energy Technology Data Exchange (ETDEWEB)

    Borio di Tigliole, Andrea; Bradley, Ed; Zhukova, Anastasia; Adelfang, Pablo [International Atomic Energy Agency, Research Reactor Section, Vienna (Austria); Shokr, Amgad [International Atomic Energy Agency, Research Reactor Safety Section, Vienna (Austria); Ridikas, Danas [International Atomic Energy Agency, Physics Section, Vienna (Austria)

    2015-08-15

    A research reactor (RR) project is a major undertaking that requires careful preparation, planning, implementation and investment in time, money, and human resources. In recent years, the interest of IAEA Member States in developing RR programmes has grown significantly, and currently, several Member States are in different stages of new RR projects. The majority of these countries are building their first RR as a key national facility for the development of their nuclear science and technology programmes, including nuclear power. In order to support Member States in such efforts, the IAEA in 2012 published the Nuclear Energy Series Report No. NP-T-5.1 on Specific Considerations and Milestones for a Research Reactor Project. To provide further support, the IAEA also published a document to assist Member States in the preparation of the bid invitation specification for the purchase of a RR. The IAEA will also continue to provide assistance for human resources development of the Member States establishing their first RR, and to facilitate sharing experience and knowledge among Member States through its programmatic activities including expert mission services, technical meetings, training courses and workshops addressing relevant technical and safety topics. This paper presents the IAEA assistance and services provided to the Member States considering new RRs, with particular emphasis on those establishing their first RR, including elaboration on the services mentioned above.

  11. Interim irradiated fuel storage facility for research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lolich, Jose [INVAP SE, Bariloche (Argentina)

    2002-07-01

    In most research reactors irradiated fuel discharged from the reactor is initially stored underwater inside the reactor building for along period of time. This allows for heat dissipation and fission product decay. In most cases this initial storage is done in a irradiated fuel storage facility pool located closed to the reactor core. After a certain cooling time, the fuel discharged should be relocated for long-term interim storage in a Irradiated Fuel Storage (IFS) Facility. IFS facilities are required for the safe storage of irradiated nuclear fuel before it is reprocessed or conditioned for disposal as radioactive waste. The IFS Facility described in this report is not an integral part of an operating nuclear reactor. This facility many be either co-located with nuclear facilities (such as a nuclear reactor or reprocessing plant) or sited independently of other nuclear facilities. (author)

  12. Proposed design for the PGAA facility at the TRIGA IPR-R1 research reactor

    OpenAIRE

    Guerra, Bruno T.; Jacimovic, Radojko; Menezes, Maria Angela BC; Leal,Alexandre S.

    2013-01-01

    Background This work presents an initial proposed design of a Prompt Gamma Activation Analysis (PGAA) facility to be installed at the TRIGA IPR-R1, a 60 years old research reactor of the Centre of Development of Nuclear Technology (CDTN) in Brazil. The basic characteristics of the facility and the results of the neutron flux are presented and discussed. Findings The proposed design is based on a quasi vertical tube as a neutron guide from the reactor core, inside the reactor pool, 6 m below t...

  13. Guideline for Performing Systematic Approach to Evaluate and Qualify Legacy Documents that Support Advanced Reactor Technology Activity

    Energy Technology Data Exchange (ETDEWEB)

    Honma, George [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-10-01

    The establishment of a systematic process for the evaluation of historic technology information for use in advanced reactor licensing is described. Efforts are underway to recover and preserve Experimental Breeder Reactor II and Fast Flux Test Facility historical data. These efforts have generally emphasized preserving information from data-acquisition systems and hard-copy reports and entering it into modern electronic formats suitable for data retrieval and examination. The guidance contained in this document has been developed to facilitate consistent and systematic evaluation processes relating to quality attributes of historic technical information (with focus on sodium-cooled fast reactor (SFR) technology) that will be used to eventually support licensing of advanced reactor designs. The historical information may include, but is not limited to, design documents for SFRs, research-and-development (R&D) data and associated documents, test plans and associated protocols, operations and test data, international research data, technical reports, and information associated with past U.S. Nuclear Regulatory Commission (NRC) reviews of SFR designs. The evaluation process is prescribed in terms of SFR technology, but the process can be used to evaluate historical information for any type of advanced reactor technology. An appendix provides a discussion of typical issues that should be considered when evaluating and qualifying historical information for advanced reactor technology fuel and source terms, based on current light water reactor (LWR) requirements and recent experience gained from Next Generation Nuclear Plant (NGNP).

  14. Development of safety analysis technology for integral reactor; evaluation on safety concerns of integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hee Chul; Kim, Woong Sik; Lee, J. H. [Korea Institute of Nuclear Safety, Taejeon (Korea)

    2002-03-01

    The Nuclear Desalination Plant (NDP) is being developed to produce electricity and fresh water, and is expected to locate near population zone. In the aspect of safety, it is required to protect the public and environment from the possible releases of fission products and to prevent the fresh water from the contamination of radioactivity. Thus, in this study, the safety characteristics of the integral reactor adopting passive and inherent safety features significantly different from existing nuclear power plants were investigated. Also, safety requirements applicable to the NDP were analyzed based on the regulatory requirements for current light water reactor and advanced reactor designs, and user requirements for small-medium size reactors. Based on these analyses, some safety concerns to be considered in the design stage have been identified and discussed. They include the use of proven technology for new safety features, systematic event classification and selection, strengthening containment function, and the safety impacts on desalination-related systems. The study presents the general safety requirements applicable to licensing of an integral reactor and suggests additional regulatory requirements, which need to be developed, based on the direction to resolution of the safety concerns. The efforts to identify and technically resolve the safety concerns in the design stage will provide the early confidence of SMART safety and the technical basis to evaluate the safety to designers and reviewers in the future. Suggestion on the development of additional regulatory requirements will contribute for the regulator to taking actions for licensing of an integral reactor. 66 refs., 5 figs., 24 tabs. (Author)

  15. Multi-Purpose Thermal Hydraulic Loop: Advanced Reactor Technology Integral System Test (ARTIST) Facility for Support of Advanced Reactor Technologies

    Energy Technology Data Exchange (ETDEWEB)

    James E. O' Brien; Piyush Sabharwall; SuJong Yoon

    2001-11-01

    Effective and robust high temperature heat transfer systems are fundamental to the successful deployment of advanced reactors for both power generation and non-electric applications. Plant designs often include an intermediate heat transfer loop (IHTL) with heat exchangers at either end to deliver thermal energy to the application while providing isolation of the primary reactor system. In order to address technical feasibility concerns and challenges a new high-temperature multi-fluid, multi-loop test facility “Advanced Reactor Technology Integral System Test facility” (ARTIST) is under development at the Idaho National Laboratory. The facility will include three flow loops: high-temperature helium, molten salt, and steam/water. Details of some of the design aspects and challenges of this facility, which is currently in the conceptual design phase, are discussed

  16. Integrated Decision-Making Tool to Develop Spent Fuel Strategies for Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Beatty, Randy L [ORNL; Harrison, Thomas J [ORNL

    2016-01-01

    IAEA Member States operating or having previously operated a Research Reactor are responsible for the safe and sustainable management and disposal of associated radioactive waste, including research reactor spent nuclear fuel (RRSNF). This includes the safe disposal of RRSNF or the corresponding equivalent waste returned after spent fuel reprocessing. One key challenge to developing general recommendations lies in the diversity of spent fuel types, locations and national/regional circumstances rather than mass or volume alone. This is especially true given that RRSNF inventories are relatively small, and research reactors are rarely operated at a high power level or duration typical of commercial power plants. Presently, many countries lack an effective long-term policy for managing RRSNF. This paper presents results of the International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) #T33001 on Options and Technologies for Managing the Back End of the Research Reactor Nuclear Fuel Cycle which includes an Integrated Decision Making Tool called BRIDE (Back-end Research reactor Integrated Decision Evaluation). This is a multi-attribute decision-making tool that combines the Total Estimated Cost of each life-cycle scenario with Non-economic factors such as public acceptance, technical maturity etc and ranks optional back-end scenarios specific to member states situations in order to develop a specific member state strategic plan with a preferred or recommended option for managing spent fuel from Research Reactors.

  17. Reactor Technology Options Study for Near-Term Deployment of GNEP Grid-Appropriate Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ingersoll, Daniel T [ORNL; Poore III, Willis P [ORNL

    2007-09-01

    World energy demand is projected to significantly increase over the coming decades. The International Energy Agency projects that electricity demand will increase 50% by 2015 and double by 2030, with most of the increase coming in developing countries as they experience double-digit rates of economic growth and seek to improve their standards of living. Energy is the necessary driver for human development, and the demand for energy in these countries will be met using whatever production technologies are available. Recognizing this inevitable energy demand and its implications for the United States, the U.S. National Security Strategy has proposed the Global Nuclear Energy Partnership (GNEP) to work with other nations to develop and deploy advanced nuclear recycling and reactor technologies. This initiative will help provide reliable, emission-free energy with less of the waste burden of older technologies and without making available separated plutonium that could be used by rogue states or terrorists for nuclear weapons. These new technologies will make possible a dramatic expansion of safe, clean nuclear energy to help meet the growing global energy demand. In other words, GNEP seeks to create an international regime to support large-scale growth in the worldwide use of nuclear energy without increasing the risk of nuclear weapon proliferation. This global expansion of nuclear power is strategically important to the United States for several reasons, including the following: (1) National security, by reducing the competition and potential for conflict over increasingly scarce fossil energy resources; (2) Economic security, by helping maintain stable prices for nonrenewable resources such as oil, gas, and coal; (3) Environmental security, by replacing or off-setting large-scale burning of greenhouse gas-emitting fuels for electricity production; and (4) Regaining technical leadership, through deployment of innovative U.S. technology-based reactors. Fully meeting

  18. Study on secondary shutdown systems in Tehran research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jalali, H.R.; Fadaei, A.H., E-mail: Fadaei_amir@aut.ac.ir; Gharib, M.

    2015-09-15

    Highlights: • A study was undertaken to summarize the techniques for secondary shutdown systems (SSS). • Neutronic calculation performed for proposed systems as SSS. • Dumping the heavy water stored in the reflector vessel is capable to shut down reactor. • Neutronic and transient calculation was done for validating the selected SSS. • All calculation shown that this system has advantages in safety and neutron economy. - Abstract: One important safety aspect of any research reactor is the ability to shut down the reactor. Usually, research reactors, currently in operation, have a single shutdown system based on the simultaneous insertion of the all control rods into the reactor core through gravity. Nevertheless, the International Atomic Energy Agency currently recommends use of two shutdown systems which are fully independent from each other to guarantee secure shutdown when one of them fails. This work presents an investigative study into secondary shutdown systems, which will be an important safety component in the research reactor and will provide another alternative way to shut down the reactor emergently. As part of this project, a study was undertaken to summarize the techniques that are currently used at world-wide research reactors for recognizing available techniques to consider in research reactors. Removal of the reflector, removal of the fuels, change in critical shape of reactor core and insertion of neutron absorber between the core and reflector are selected as possible techniques in mentioned function. In the next step, a comparison is performed for these methods from neutronic aspects. Then, chosen method is studied from the transient behavior point of view. Tehran research reactor which is a 5 MW open-pool reactor selected as a case study and all calculations are carried out for it. It has 5 control rods which serve the purpose of both reactivity control and shutdown of reactor under abnormal condition. Results indicated that heavy

  19. Materials and Components Technology Division research summary, 1992

    Energy Technology Data Exchange (ETDEWEB)

    1992-11-01

    The Materials and Components Technology Division (MCT) provides a research and development capability for the design, fabrication, and testing of high-reliability materials, components, and instrumentation. Current divisional programs related to nuclear energy support the development of the Integral Fast Reactor (IFR): life extension and accident analyses for light water reactors (LWRs); fuels development for research and test reactors; fusion reactor first-wall and blanket technology; and safe shipment of hazardous materials. MCT Conservation and Renewables programs include major efforts in high-temperature superconductivity, tribology, nondestructive evaluation (NDE), and thermal sciences. Fossil Energy Programs in MCT include materials development, NDE technology, and Instrumentation design. The division also has a complementary instrumentation effort in support of Arms Control Technology. Individual abstracts have been prepared for the database.

  20. An Innovative Reactor Technology to Improve Indoor Air Quality

    Energy Technology Data Exchange (ETDEWEB)

    Rempel, Jane [TIAX LLC., Lexington, MA (United States)

    2013-03-30

    As residential buildings achieve tighter envelopes in order to minimize energy used for space heating and cooling, accumulation of indoor air pollutants such as volatile organic compounds (VOCs), becomes a major concern causing poor air quality and increased health risks. Current VOC removal methods include sorbents, ultraviolet photocatalytic oxidation (UVPCO), and increased ventilation, but these methods do not capture or destroy all VOCs or are prohibitively expensive to implement. TIAX's objective in this program was to develop a new VOC removal technology for residential buildings. This novel air purification technology is based on an innovative reactor and light source design along with UVPCO properties of the chosen catalyst to purify indoor air and enhance indoor air quality (IAQ). During the program we designed, fabricated and tested a prototype air purifier to demonstrate its feasibility and effectiveness. We also measured kinetics of VOC destruction on photocatalysts, providing deep insight into reactor design.

  1. HELIAS stellarator reactor studies and related European technology studies

    Energy Technology Data Exchange (ETDEWEB)

    Grieger, G. (Max-Planck-Institut fuer Plasmaphysik, EURATOM-Association, D-85748 Garching (Germany)); Nuehrenberg, J. (Max-Planck-Institut fuer Plasmaphysik, EURATOM-Association, D-85748 Garching (Germany)); Renner, H. (Max-Planck-Institut fuer Plasmaphysik, EURATOM-Association, D-85748 Garching (Germany)); Sapper, J. (Max-Planck-Institut fuer Plasmaphysik, EURATOM-Association, D-85748 Garching (Germany)); Wobig, H. (Max-Planck-Institut fuer Plasmaphysik, EURATOM-Association, D-85748 Garching (Germany))

    1994-08-01

    Research on stellarators has been carried out in Europe since the early years of the fusion programme. Early studies of this reactor concept were done at the Culham Laboratory of the UK Atomic Energy Authority. Such classical stellarators, however, have poor reactor prospects in spite of the significant advantage of not needing a large toroidal plasma current. It seemed to be just this large toroidal plasma current which has led to intrinsic deficiencies with respect to reactor potential of the tokamak. Expecting that these deficiencies would disappear for a concept without such a current, the Institut fuer Plasmaphysik developed, in a roll-back fashion, i.e. by starting from reactor considerations, the concept of the helical axis advanced stellarator (HELIAS). The results achieved look very promising indeed. Since tokamaks and stellarators show many similarities, there appeared no need as yet for a new stand-alone stellarator reactor study. The work was rather concentrated on the few but decisive differences between the two concepts and on evaluation of their relative importance. Studies on the coil system, the stress distribution in the supporting material, the space needed for an efficient blanket system, the properties of the exhaust system, etc. have been done. Applying contemporary scaling laws, it turns out that although the aspect ratio of such advanced stellarators is larger than that of tokamaks, the plasma volume is about the same. The magnetic energy needed for plasma confinement is considerably lower and the mass utilization tends to be larger than for comparable tokamaks. It also follows that a number of reactor components needed for tokamak operation (e.g. current drive, feedback stabilization, disruption prevention) are not needed in stellarators, making this type of reactor and its operation simpler. Such results would have a large influence on selection of the final concept and the further evolution of the fusion programme.

  2. Need for space-time analyses of research reactor transients

    Energy Technology Data Exchange (ETDEWEB)

    Jatuff, F.E. [INVAP S.E., de Bariloche (Argentina)

    1997-12-01

    The success of the point-kinetics approximation to represent the time behavior of research reactors relies on the fact that research reactor cores are small enough to be neutronically tightly coupled; the core is small when measured in diffusion lengths. This fact implies that a certain change in a part of the core is immediately observed by the whole system. The propagation of changes is so fast that the core exhibits a shape function that is practically unchanged during the transient; the amplitude function, the only unknown of the problem, represents the full knowledge of the core response. One is immediately warned to look for the truth of this assumption. How small should a research reactor core be to be sure that point kinetics is a valid assumption? This question is becoming increasingly important because the tendency is to increase the size of research reactor cores to make them capable of various simultaneous uses (multipurpose characteristics), with powers in the range of tens of thermal megawatts. One of the lines of investigation at the Department of Reactor Physics is related to scenarios of Materials Test Reactor (MTR)-type research reactor transients for which space-time kinetics would bring a more profound insight than point kinetics.

  3. Development of core design and analyses technology for integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zee, Sung Quun; Lee, C. C.; Song, J. S. and others

    1999-03-01

    Integral reactors are developed for the applications such as sea water desalination, heat energy for various industries, and power sources for large container ships. In order to enhance the inherent and passive safety features, low power density concept is chosen for the integral reactor SMART. Moreover, ultra-longer cycle and boron-free operation concepts are reviewed for better plant economy and simple design of reactor system. Especially, boron-free operation concept brings about large difference in core configurations and reactivity controls from those of the existing large size commercial nuclear power plants and also causes many differences in the safety aspects. The ultimate objectives of this study include detailed core design of a integral reactor, development of the core design system and technology, and finally acquisition of the system design certificate. The goal of the first stage is the conceptual core design, that is, to establish the design bases and requirements suitable for the boron-free concept, to develop a core loading pattern, to analyze the nuclear, thermal and hydraulic characteristics of the core and to perform the core shielding design. Interface data for safety and performance analyses including fuel design data are produced for the relevant design analysis groups. Nuclear, thermal and hydraulic, shielding design and analysis code systems necessary for the core conceptual design are established through modification of the existing design tools and newly developed methodology and code modules. Core safety and performance can be improved by the technology development such as boron-free core optimization, advaned core monitoring and operational aid system. Feasiblity study on the improvement of the core protection and monitoring system will also contribute toward core safety and performance. Both the conceptual core design study and the related technology will provide concrete basis for the next design phase. This study will also

  4. Evaluation on safety concerns of integral reactor: development of safety analysis technology for integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, W. S.; Kim, W. K.; Yun, Y. G.; Ahn, H. J.; Lee, J. S.; Lee, S. G.; Sin, A. D. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-03-01

    The Nuclear Desalination Plant (NDP) is being developed to produce electricity and fresh water, and is expected to locate near population zone. In the aspect of safety, it is required to protect the public and environment from the possible releases of fission products and to prevent the fresh water from the contamination of radioactivity. Thus, in a present study, the safety characteristics of the integral reactor adopting passive and inherent safety features significantly different from existing nuclear power plants were investigated. Also, safety requirements applicable to the NDP were analyzed based on the regulatory requirements for current light water reactor and advanced reactor designs, and user requirements for small-medium size reactors. Based on these analyses, some safety concerns to be considered in the design stage have been identified and discussed. They includes the use of proven technology for new safety features, systematic event classification and selection, strengthening containment function, and the safety impacts on desalination-related systems. These efforts to identify and technically resolve the safety concerns in the design stage will provide the early confidence of SMART safety and the technical basis to evaluate the safety to designers and reviewers in the future. 62 refs., 3 figs., 21 tabs. (Author)

  5. New reactor technology: safety improvements in nuclear power systems.

    Science.gov (United States)

    Corradini, M L

    2007-11-01

    Almost 450 nuclear power plants are currently operating throughout the world and supplying about 17% of the world's electricity. These plants perform safely, reliably, and have no free-release of byproducts to the environment. Given the current rate of growth in electricity demand and the ever growing concerns for the environment, nuclear power can only satisfy the need for electricity and other energy-intensive products if it can demonstrate (1) enhanced safety and system reliability, (2) minimal environmental impact via sustainable system designs, and (3) competitive economics. The U.S. Department of Energy with the international community has begun research on the next generation of nuclear energy systems that can be made available to the market by 2030 or earlier, and that can offer significant advances toward these challenging goals; in particular, six candidate reactor system designs have been identified. These future nuclear power systems will require advances in materials, reactor physics, as well as thermal-hydraulics to realize their full potential. However, all of these designs must demonstrate enhanced safety above and beyond current light water reactor systems if the next generation of nuclear power plants is to grow in number far beyond the current population. This paper reviews the advanced Generation-IV reactor systems and the key safety phenomena that must be considered to guarantee that enhanced safety can be assured in future nuclear reactor systems.

  6. China Advanced Research Reactor Project Progress in 2011

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

    2011, China Advanced Research Reactor (CARR) Project finished the B stage commissioning and resolved the relative technical problems. Meanwhile, the acceptance items and the cold neutron source were carrying out.

  7. Status and some safety philosophies of the China advanced research reactor CARR

    Energy Technology Data Exchange (ETDEWEB)

    Luzheng Yuan [China Inst. of Atomic Energy, Beijing, BJ (China). Reactor Engineering Research and Design Dept.

    2001-07-01

    The existing two research reactors, HWRR (heavy water research reactor) and SPR (swimming pool reactor), have been operated by China Institute of Atomic Energy (CIAE) since, respectively, 1958 and 1964, and are both in extending service and facing the aging problem. It is expected that they will be out of service successively in the beginning decade of the 21{sup st} century. A new, high performance and multipurpose research reactor called China advanced research reactor (CARR) will replace these two reactors. This new reactor adopts the concept of inverse neutron trap compact core structure with light water as coolant and heavy water as the outer reflector. Its design goal is as follows: under the nuclear power of 60MW, the maximum unperturbed thermal neutron flux in peripheral D{sub 2}O reflector not less than 8 x 10{sup 14} n/cm{sup 2}. s while in central experimental channel, if the central cell to be replaced by an experimental channel, the corresponding value not less than 1 x 10{sup 15} n/cm{sup 2}. s. The main applications for this research reactor will cover RI production, neutron scattering experiments, NAA and its applications, neutron photography, NTD for monocrystaline silicon and applications on reactor engineering technology. By the end of 1999, the preliminary design of CARR was completed, then the draft of preliminary safety analysis report (PSAR) was submitted to the relevant authority at the end of 2000 for being reviewed. Now, the CARR project has entered the detail design phase and safety reviewing procedure for obtaining the construction permit from the relevant licensing authority. This paper will only briefly introduce some aspects of safety philosophy of CARR design and PSAR. (orig.)

  8. Modular Pebble Bed Reactor Project, University Research Consortium Annual Report

    Energy Technology Data Exchange (ETDEWEB)

    Petti, David Andrew

    2000-07-01

    This project is developing a fundamental conceptual design for a gas-cooled, modular, pebble bed reactor. Key technology areas associated with this design are being investigated which intend to address issues concerning fuel performance, safety, core neutronics and proliferation resistance, economics and waste disposal. Research has been initiated in the following areas: · Improved fuel particle performance · Reactor physics · Economics · Proliferation resistance · Power conversion system modeling · Safety analysis · Regulatory and licensing strategy Recent accomplishments include: · Developed four conceptual models for fuel particle failures that are currently being evaluated by a series of ABAQUS analyses. Analytical fits to the results are being performed over a range of important parameters using statistical/factorial tools. The fits will be used in a Monte Carlo fuel performance code, which is under development. · A fracture mechanics approach has been used to develop a failure probability model for the fuel particle, which has resulted in significant improvement over earlier models. · Investigation of fuel particle physio-chemical behavior has been initiated which includes the development of a fission gas release model, particle temperature distributions, internal particle pressure, migration of fission products, and chemical attack of fuel particle layers. · A balance of plant, steady-state thermal hydraulics model has been developed to represent all major components of a MPBR. Component models are being refined to accurately reflect transient performance. · A comparison between air and helium for use in the energy-conversion cycle of the MPBR has been completed and formed the basis of a master’s degree thesis. · Safety issues associated with air ingress are being evaluated. · Post shutdown, reactor heat removal characteristics are being evaluated by the Heating-7 code. · PEBBED, a fast deterministic neutronic code package suitable for

  9. Development of safety analysis technology for integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sim, Suk K.; Song, J. H.; Chung, Y. J. and others

    1999-03-01

    Inherent safety features and safety system characteristics of the SMART integral reactor are investigated in this study. Performance and safety of the SMART conceptual design have been evaluated and confirmed through the performance and safety analyses using safety analysis system codes as well as a preliminary performance and safety analysis methodology. SMART design base events and their acceptance criteria are identified to develop a preliminary PIRT for the SMART integral reactor. Using the preliminary PIRT, a set of experimental program for the thermal hydraulic separate effect tests and the integral effect tests was developed for the thermal hydraulic model development and the system code validation. Safety characteristics as well as the safety issues of the integral reactor has been identified during the study, which will be used to resolve the safety issues and guide the regulatory criteria for the integral reactor. The results of the performance and safety analyses performed during the study were used to feedback for the SMART conceptual design. The performance and safety analysis code systems as well as the preliminary safety analysis methodology developed in this study will be validated as the SMART design evolves. The performance and safety analysis technology developed during the study will be utilized for the SMART basic design development. (author)

  10. Development of Safety Analysis Technology for Integral Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sim, S. K. [Korea Atomic Energy Research Institute, Taejeon (Korea); Seul, K. W.; Kim, W. S.; Kim, W. K.; Yun, Y. G.; Ahn, H. J.; Lee, J. S.; Sin, A. D. [Korea Institute of Nuclear Safety, Taejeon (Korea)

    2000-03-01

    The Nuclear Desalination Plant(NDP) is being developed to produce electricity and fresh water, and is expected to locate near population zone. In the aspect of safety, it is required to protect the public and environment from the possible releases of fission products and to prevent the fresh water from the contamination of radioactivity. Thus, in a present study, the safety characteristics of the integral reactor adopting passive and inherent safety features significantly different from existing nuclear power plants were investigated based on the design of foreign and domestic integral reactors. Also, safety requirements applicable to the NDP were analyzed based on the regulatory requirements for current and advanced reactor designs, and use requirements for small-medium size reactors. Based on these analyses, some safety concerns to be considered in the design stage have been identified. They includes the use of proven technology for new safety systems, the systematic classification and selection of design basis accidents, and the safety assurance of desalination-related systems. These efforts to identify and resolve the safety concerns in the design stage will provide the early confidence of SMART safety to designers, and the technical basis to evaluate the safety to reviewers in the future. 8 refs., 20 figs., 4 tabs. (Author)

  11. Development of an educational nuclear research reactor simulator

    Energy Technology Data Exchange (ETDEWEB)

    Arafa, Amany Abdel Aziz; Saleh, Hassan Ibrahim [Atomic Energy Authority, Cairo (Egypt). Radiation Engineering Dept.; Ashoub, Nagieb [Atomic Energy Authority, Cairo (Egypt). Reactor Physics Dept.

    2014-12-15

    This paper introduces the development of a research reactor educational simulator based on LabVIEW that allows the training of operators and studying different accident scenarios and the effects of operational parameters on the reactor behavior. Using this simulator, the trainee can test the interaction between the input parameters and the reactor activities. The LabVIEW acts as an engine implements the reactor mathematical models. In addition, it is used as a tool for implementing the animated graphical user interface. This simulator provides the training requirements for both of the reactor staff and the nuclear engineering students. Therefore, it uses dynamic animation to enhance learning and interest for a trainee on real system problems and provides better visual effects, improved communications, and higher interest levels. The benefits of conducting such projects are to develop the expertise in this field and save costs of both operators training and simulation courses.

  12. Cooling Performance of Natural Circulation for a Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Suki; Chun, J. H.; Yum, S. B. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    This paper deals with the core cooling performance by natural circulation during normal operation and a flow channel blockage event in an open tank-in-pool type research reactor. The cooling performance is predicted by using the RELAP5/ MOD3.3 code. The core decay heat is usually removed by natural circulation to the reactor pool water in open tank-in-pool type research reactors with the thermal power less than several megawatts. Therefore, these reactors have generally no active core cooling system against a loss of normal forced flow. In reactors with the thermal power less than around one megawatt, the reactor core can be cooled down by natural circulation even during normal full power operation. The cooling performance of natural circulation in an open tank-in-pool type research reactor has been investigated during the normal natural circulation and a flow channel blockage event. It is found that the maximum powers without void generation at the hot channel are around 1.16 MW and 820 kW, respectively, for the normal natural circulation and the flow channel blockage event.

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

    Energy Technology Data Exchange (ETDEWEB)

    1993-07-01

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

  14. Oak Ridge National Laboratory Research Reactor Experimenters' Guide

    Energy Technology Data Exchange (ETDEWEB)

    Cagle, C.D. (comp.)

    1982-04-01

    The Oak Ridge National Laboratory (ORNL) operates six research reactors dedicated to research and development work as well as radioisotope production. These reactors are used by ORNL and qualified non-ORNL research and development groups. The purpose of this report is to provide information to research personnel concerning the facilities and the ORNL research and services groups available to assist in the design, fabrication, operation, and post-operation examination of irradiation assemblies. Safety and operability reviews and quality assurance requirements are also described.

  15. TORBED process reactor technology for asphalt paving recycle

    Energy Technology Data Exchange (ETDEWEB)

    Wellwood, G. A. [Torftech Limited, Reading (United Kingdom); Laughlin, R. G. [Torftech Canada Inc., Mississauga, ON (Canada)

    2001-07-01

    Fundamental principles underlying the TORBED reactor technologies are reviewed. Advantages, limitations, and the experience gained in applying the reactor to a range of material processing needs, particularly reprocessing asphalt paving waste, are discussed. The TORBED reactor consists of a compact shallow packed bed of particles suspended above an annular ring of stationary blades through which a process gas stream is passed at high velocity. Unlike fluidized beds, the process gas mass flow can be set to suit the process, i. e. a smaller gas mass flow can be used at a higher velocity at exit from the blades to keep the bed in proper motion. Advantages are: (1) smaller reactor size with rapid start-up and program change, (2) faster and more precise processing of particles, resulting in a consistent product or process, (3) low process stream pressure losses which facilitate process gas recirculation, (4) ability to process widely graded and irregularly shaped feed stocks, and (5) simplicity of operation through real time control. The 'expanded' version of the TORBED reactor also permits fuel injection to generate process gas temperatures in excess of 1600 degrees C, and allows high gas flow rates with low pressure drop. When used to reprocess asphalt paving waste, the coal tar, used in some 25 per cent of existing roads in the Netherlands, can be easily removed and destroyed and the cleaned stone and the energy generated form the destruction of the coal tar, can be reused in creating new bitumen- containing asphalt paving. Excellent results are reported to have been obtained in pilot studies. Pilot trial also have been conducted on the combustion of wood waste for power generation, with encouraging results. Other pilot projects addressed alternative cementitious materials. 5 refs.

  16. Joint research center activity in thermonuclear fusion technology

    Energy Technology Data Exchange (ETDEWEB)

    Casini, G.; Rocco, P. (Commission of the European Communities, Ispra (Italy). Joint Research Centre)

    1984-04-01

    A review of the activities in progress in the field of thermonuclear fusion technology at the Joint Research Centre of the European Communities is presented. The research areas are: (I) reactor studies, including conceptual design studies of experimental Tokamak reactors (INTOR/NET) and safety analyses; (II) experimental investigation on first wall and blanket materials and components. Emphasis has been given to those topics which are not reported in detail in the following articles of the issue.

  17. Building on knowledge base of sodium cooled fast spectrum reactors to develop materials technology for fusion reactors

    Science.gov (United States)

    Raj, Baldev; Rao, K. Bhanu Sankara

    2009-04-01

    The alloys 316L(N) and Mod. 9Cr-1Mo steel are the major structural materials for fabrication of structural components in sodium cooled fast reactors (SFRs). Various factors influencing the mechanical behaviour of these alloys and different modes of deformation and failure in SFR systems, their analysis and the simulated tests performed on components for assessment of structural integrity and the applicability of RCC-MR code for the design and validation of components are highlighted. The procedures followed for optimal design of die and punch for the near net shape forming of petals of main vessel of 500 MWe prototype fast breeder reactor (PFBR); the safe temperature and strain rate domains established using dynamic materials model for forming of 316L(N) and 9Cr-1Mo steels components by various industrial processes are illustrated. Weldability problems associated with 316L(N) and Mo. 9Cr-1Mo are briefly discussed. The utilization of artificial neural network models for prediction of creep rupture life and delta-ferrite in austenitic stainless steel welds is described. The usage of non-destructive examination techniques in characterization of deformation, fracture and various microstructural features in SFR materials is briefly discussed. Most of the experience gained on SFR systems could be utilized in developing science and technology for fusion reactors. Summary of the current status of knowledge on various aspects of fission and fusion systems with emphasis on cross fertilization of research is presented.

  18. Design and implementation progress of multi-purpose simulator for nuclear research reactor using LabVIEW

    Energy Technology Data Exchange (ETDEWEB)

    Arafa, Amany Abdel Aziz; Saleh, Hassan Ibrahim [Atomic Energy Authority, Cairo (Egypt). Radiation Engineering Dept.; Ashoub, Nagieb [Atomic Energy Authority, Cairo (Egypt). Nuclear Research Center

    2015-11-15

    This paper illustrates the neutronic and thermal hydraulic models that were implemented in the nuclear research reactor simulator based on LabVIEW. It also describes the system and transient analysis of the simulator that takes into consideration the temperature effects and poisoning. This simulator is designed to be a multi-purpose in which the operator could understand the effects of the input parameters on the reactor. A designer can study different solutions for virtual reactor accident scenarios. The main features of the simulator are the flexibility to design and maintain the interface and the ability to redesign and remodel the reactor core engine. The developed reactor simulator permits to acquire hands-on the experience of the physics and technology of nuclear reactors including reactivity control, thermodynamics, technology design and safety system design. This simulator can be easily customizable and upgradable and new opportunities for collaboration between academic groups could be conducted.

  19. Space Technology Research Grants Program

    Data.gov (United States)

    National Aeronautics and Space Administration — The Space Technology Research Grants Program will accelerate the development of "push" technologies to support the future space science and exploration...

  20. Technologies for Upgrading Light Water Reactor Outlet Temperature

    Energy Technology Data Exchange (ETDEWEB)

    Daniel S. Wendt; Piyush Sabharwall; Vivek Utgikar

    2013-07-01

    Nuclear energy could potentially be utilized in hybrid energy systems to produce synthetic fuels and feedstocks from indigenous carbon sources such as coal and biomass. First generation nuclear hybrid energy system (NHES) technology will most likely be based on conventional light water reactors (LWRs). However, these LWRs provide thermal energy at temperatures of approximately 300°C, while the desired temperatures for many chemical processes are much higher. In order to realize the benefits of nuclear hybrid energy systems with the current LWR reactor fleets, selection and development of a complimentary temperature upgrading technology is necessary. This paper provides an initial assessment of technologies that may be well suited toward LWR outlet temperature upgrading for powering elevated temperature industrial and chemical processes during periods of off-peak power demand. Chemical heat transformers (CHTs) are a technology with the potential to meet LWR temperature upgrading requirements for NHESs. CHTs utilize chemical heat of reaction to change the temperature at which selected heat sources supply or consume thermal energy. CHTs could directly utilize LWR heat output without intermediate mechanical or electrical power conversion operations and the associated thermodynamic losses. CHT thermal characteristics are determined by selection of the chemical working pair and operating conditions. This paper discusses the chemical working pairs applicable to LWR outlet temperature upgrading and the CHT operating conditions required for providing process heat in NHES applications.

  1. Validation of Reactor Physics-Thermal hydraulics Calculations for Research Reactors Cooled by the Laminar Flow of Water

    Energy Technology Data Exchange (ETDEWEB)

    Jordan, K. A.; Schubring, D. [Univ. of Florida, Florida (United States); Girardin, G.; Pautz, A. [Swiss Federal Institute of Technology, Zuerich (Switzerland)

    2013-07-01

    A collaboration between the University of Florida and the Swiss Federal Institute of Technology, Lausanne (EPFL) has been formed to develop and validate detailed coupled multiphysics models of the zero-power (100 W) CROCUS reactor at EPFL and the 100 kW University of Florida Training Reactor, for the comprehensive analysis of the reactor behavior under transient (neutronic or thermal-hydraulic induced) conditions. These two reactors differ significantly in the core design and thermal power output, but share unique heat transfer and flow characteristics. They are characterized by single-phase laminar water flow at near-atmospheric pressures in complex geometries with the possibility of mechanically entrained air bubbles. Validation experiments will be designed to expand the validation domain of these existing models, computational codes and techniques. In this process, emphasis will be placed on validation of the coupled models developed to gain confidence in their applicability for safety analysis. EPFL is responsible for the design and implementation of transient experiments to generate a database of reactor parameters (flow distribution, power profile, and power evolution) to be used to validate against code predictions. The transient experiments performed at EPFL will be simulated on the basis of developed models for these tasks. Comparative analysis will be performed with SERPENT and MCNPX reference core models. UF focuses on the generation of the coupled neutron kinetics and thermal-hydraulic models, including implementation of a TRACE/PARCS reactor simulator model, a PARET model, and development of full-field computational fluid dynamics models (using OpenFOAM) for refined thermal-hydraulics physics treatments. In this subtask of the project, the aim is to verify by means of CFD the validity of TRACE predictions for near-atmospheric pressure water flow in the presence of mechanically entrained air bubbles. The scientific understanding of these multiphysics

  2. Component and Technology Development for Advanced Liquid Metal Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, Mark [Univ. of Wisconsin, Madison, WI (United States)

    2017-01-30

    The following report details the significant developments to Sodium Fast Reactor (SFR) technologies made throughout the course of this funding. This report will begin with an overview of the sodium loop and the improvements made over the course of this research to make it a more advanced and capable facility. These improvements have much to do with oxygen control and diagnostics. Thus a detailed report of advancements with respect to the cold trap, plugging meter, vanadium equilibration loop, and electrochemical oxygen sensor is included. Further analysis of the university’s moving magnet pump was performed and included in a section of this report. A continuous electrical resistance based level sensor was built and tested in the sodium with favorable results. Materials testing was done on diffusion bonded samples of metal and the results are presented here as well. A significant portion of this work went into the development of optical fiber temperature sensors which could be deployed in an SFR environment. Thus, a section of this report presents the work done to develop an encapsulation method for these fibers inside of a stainless steel capillary tube. High temperature testing was then done on the optical fiber ex situ in a furnace. Thermal response time was also explored with the optical fiber temperature sensors. Finally these optical fibers were deployed successfully in a sodium environment for data acquisition. As a test of the sodium deployable optical fiber temperature sensors they were installed in a sub-loop of the sodium facility which was constructed to promote the thermal striping effect in sodium. The optical fibers performed exceptionally well, yielding unprecedented 2 dimensional temperature profiles with good temporal resolution. Finally, this thermal striping loop was used to perform cross correlation velocimetry successfully over a wide range of flow rates.

  3. Activities for extending the lifetime of MINT research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bokhari, Adnan; Kassim, Mohammad Suhaimi [Malaysian Inst. for Nuclear Technology Research (MINT), Bangi, Kajang (Malaysia)

    1998-10-01

    MINT TRIGA Reactor is a 1-MW swimming pool nuclear reactor commissioned in June 1982. Since then, it has been used for research, isotope production, neutron activation, neutron radiography and manpower training. The total operating time till the end on September 1997 is 16968 hours with cumulative total energy release of 11188 MW-hours. After more than fifteen years of successful operation, some deterioration in components and associated systems has been observed. This paper describes some of the activities carried out to increase the lifetime and to reduce the shutdown time of the reactor. (author)

  4. Neutron beam facilities at the Australian Replacement Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kennedy, Shane; Robinson, Robert; Hunter, Brett [Physics Division, ANSTO, Lucas Heights (Australia)

    2001-03-01

    Australia is building a research reactor to replace the HIFAR reactor at Lucas Heights by the end of 2005. Like HIFAR, the Replacement Research Reactor will be multipurpose with capabilities for both neutron beam research and radioisotope production. It will be a pool-type reactor with thermal neutron flux (unperturbed) of 4 x 10{sup 14} n/cm{sup 2}/sec and a liquid D{sub 2} cold neutron source. Cold and thermal neutron beams for neutron beam research will be provided at the reactor face and in a large neutron guide hall. Supermirror neutron guides will transport cold and thermal neutrons to the guide hall. The reactor and the associated infrastructure, with the exception of the neutron beam instruments, is to be built by INVAP S.E. under contract. The neutron beam instruments will be developed by ANSTO, in consultation with the Australian user community. This status report includes a review the planned scientific capabilities, a description of the facility and a summary of progress to date. (author)

  5. Background Radiation Measurements at High Power Research Reactors

    CERN Document Server

    Ashenfelter, J; Baldenegro, C X; Band, H R; Barclay, G; Bass, C D; Berish, D; Bowden, N S; Bryan, C D; Cherwinka, J J; Chu, R; Classen, T; Davee, D; Dean, D; Deichert, G; Dolinski, M J; Dolph, J; Dwyer, D A; Fan, S; Gaison, J K; Galindo-Uribarri, A; Gilje, K; Glenn, A; Green, M; Han, K; Hans, S; Heeger, K M; Heffron, B; Jaffe, D E; Kettell, S; Langford, T J; Littlejohn, B R; Martinez, D; McKeown, R D; Morrell, S; Mueller, P E; Mumm, H P; Napolitano, J; Norcini, D; Pushin, D; Romero, E; Rosero, R; Saldana, L; Seilhan, B S; Sharma, R; Stemen, N T; Surukuchi, P T; Thompson, S J; Varner, R L; Wang, W; Watson, S M; White, B; White, C; Wilhelmi, J; Williams, C; Wise, T; Yao, H; Yeh, M; Yen, Y -R; Zhang, C; Zhang, X

    2016-01-01

    Research reactors host a wide range of activities that make use of the intense neutron fluxes generated at these facilities. Recent interest in performing measurements with relatively low event rates, e.g. reactor antineutrino detection, at these facilities necessitates a detailed understanding of background radiation fields. Both reactor-correlated and naturally occurring background sources are potentially important, even at levels well below those of importance for typical activities. Here we describe a comprehensive series of background assessments at three high-power research reactors, including $\\gamma$-ray, neutron, and muon measurements. For each facility we describe the characteristics and identify the sources of the background fields encountered. The general understanding gained of background production mechanisms and their relationship to facility features will prove valuable for the planning of any sensitive measurement conducted therein.

  6. Background radiation measurements at high power research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ashenfelter, J. [Wright Laboratory, Department of Physics, Yale University, New Haven, CT 06520 (United States); Balantekin, B. [Department of Physics, University of Wisconsin, Madison, WI 53706 (United States); Baldenegro, C.X. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Band, H.R. [Wright Laboratory, Department of Physics, Yale University, New Haven, CT 06520 (United States); Barclay, G. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Bass, C.D. [Department of Chemistry and Physics, Le Moyne College, Syracuse, NY 13214 (United States); Berish, D. [Department of Physics, Temple University, Philadelphia, PA 19122 (United States); Bowden, N.S., E-mail: nbowden@llnl.gov [Nuclear and Chemical Sciences Division, Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Bryan, C.D. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Cherwinka, J.J. [Physical Sciences Laboratory, University of Wisconsin, Madison, WI 53706 (United States); Chu, R. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Department of Physics, University of Tennessee, Knoxville, TN 37996 (United States); Classen, T. [Nuclear and Chemical Sciences Division, Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Davee, D. [Department of Physics, College of William and Mary, Williamsburg, VA 23187 (United States); Dean, D.; Deichert, G. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Dolinski, M.J. [Department of Physics, Drexel University, Philadelphia, PA 19104 (United States); Dolph, J. [Physics Department, Brookhaven National Laboratory, Upton, NY 11973 (United States); Dwyer, D.A. [Physics Division, Lawrence Berkeley National Laboratory, Berkeley, CA 94720 (United States); Fan, S. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Department of Physics, University of Tennessee, Knoxville, TN 37996 (United States); and others

    2016-01-11

    Research reactors host a wide range of activities that make use of the intense neutron fluxes generated at these facilities. Recent interest in performing measurements with relatively low event rates, e.g. reactor antineutrino detection, at these facilities necessitates a detailed understanding of background radiation fields. Both reactor-correlated and naturally occurring background sources are potentially important, even at levels well below those of importance for typical activities. Here we describe a comprehensive series of background assessments at three high-power research reactors, including γ-ray, neutron, and muon measurements. For each facility we describe the characteristics and identify the sources of the background fields encountered. The general understanding gained of background production mechanisms and their relationship to facility features will prove valuable for the planning of any sensitive measurement conducted therein.

  7. Model Based Cyber Security Analysis for Research Reactor Protection System

    Energy Technology Data Exchange (ETDEWEB)

    Sho, Jinsoo; Rahman, Khalil Ur; Heo, Gyunyoung [Kyung Hee Univ., Yongin (Korea, Republic of); Son, Hanseong [Joongbu Univ., Geumsan (Korea, Republic of)

    2013-07-01

    The study on the qualitative risk due to cyber-attacks into research reactors was performed using bayesian Network (BN). This was motivated to solve the issues of cyber security raised due to digitalization of instrumentation and control (I and C) system. As a demonstrative example, we chose the reactor protection system (RPS) of research reactors. Two scenarios of cyber-attacks on RPS were analyzed to develop mitigation measures against vulnerabilities. The one is the 'insertion of reactor trip' and the other is the 'scram halt'. The six mitigation measures are developed for five vulnerability for these scenarios by getting the risk information from BN.

  8. Neutron scattering at Australia's replacement research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kennedy, S.J.; Robinson, R.A.; Hunter, B.A. [Physics Division, ANSTO, Lucas Heights, NSW (Australia)

    2001-03-01

    On August 25{sup th} 1999, the Australian government gave final approval to build a research reactor to replace the existing HIFAR reactor at Lucas Heights. The replacement reactor, which will commence operation in 2005, will be multipurpose with capabilities for both neutron beam research and radioisotope production. Cold, and thermal neutron sources are to be installed and supermirror guides will transport cold and thermal neutron beams into a large modern guide hall. The reactor and associated infrastructure is to be built by INVAP, SE and subcontractors under contract. The neutron beam instruments will be developed by ANSTO in consultation with the Australian user community and interested overseas parties. We review the planned scientific capabilities, give a description of the facility and a status report on the activities so far. (author)

  9. PHYSICS AND SAFETY ANALYSIS FOR THE NIST RESEARCH REACTOR.

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, L.; Diamond, D.; Xu, J.; Carew, J.; Rorer, D.

    2004-03-31

    Detailed reactor physics and safety analyses have been performed for the 20 MW D{sub 2}O moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The analyses provide an update to the Final Safety Analysis Report (FSAR) and employ state-of-the-art calculational methods. Three-dimensional Monte Carlo neutron and photon transport calculations were performed with the MCNP code to determine the safety parameters for the NBSR. The core depletion and determination of the fuel compositions were performed with MONTEBURNS. MCNP calculations were performed to determine the beginning, middle, and end-of-cycle power distributions, moderator temperature coefficient, and shim safety arm, beam tube and void reactivity worths. The calculational model included a plate-by-plate description of each fuel assembly, axial mid-plane water gap, beam tubes and the tubular geometry of the shim safety arms. The time-dependent analysis of the primary loop was determined with a RELAP5 transient analysis model that includes the pump, heat exchanger, fuel element geometry, and flow channels for both the six inner and twenty-four outer fuel elements. The statistical analysis used to assure protection from critical heat flux (CHF) was performed using a Monte Carlo simulation of the uncertainties contributing to the CHF calculation. The power distributions used to determine the local fuel conditions and margin to CHF were determined with MCNP. Evaluations were performed for the following accidents: (1) the control rod withdrawal startup accident, (2) the maximum reactivity insertion accident, (3) loss-of-flow resulting from loss of electrical power, (4) loss-of-flow resulting from a primary pump seizure, (5) loss-of-flow resulting from inadvertent throttling of a flow control valve, (6) loss-of-flow resulting from failure of both shutdown cooling pumps and (7) misloading of a fuel element. In both the startup and maximum reactivity insertion accidents, the

  10. Nigerian Journal of Technological Research

    African Journals Online (AJOL)

    The Nigerian Journal of Technological Research is a pure scientific journal with a ... technology to its immediate environs and the international community. ... and communication Technology; Management and Entrepreneurship sciences.

  11. Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hogerton, John

    1964-01-01

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

  12. Strategic Need for Multi-Purpose Thermal Hydraulic Loop for Support of Advanced Reactor Technologies

    Energy Technology Data Exchange (ETDEWEB)

    James E. O' Brien; Piyush Sabharwall; Su-Jong Yoon; Gregory K. Housley

    2014-09-01

    This report presents a conceptual design for a new high-temperature multi fluid, multi loop test facility for the INL to support thermal hydraulic, materials, and thermal energy storage research for nuclear and nuclear-hybrid applications. In its initial configuration, the facility will include a high-temperature helium loop, a liquid salt loop, and a hot water/steam loop. The three loops will be thermally coupled through an intermediate heat exchanger (IHX) and a secondary heat exchanger (SHX). Research topics to be addressed with this facility include the characterization and performance evaluation of candidate compact heat exchangers such as printed circuit heat exchangers (PCHEs) at prototypical operating conditions, flow and heat transfer issues related to core thermal hydraulics in advanced helium-cooled and salt-cooled reactors, and evaluation of corrosion behavior of new cladding materials and accident-tolerant fuels for LWRs at prototypical conditions. Based on its relevance to advanced reactor systems, the new facility has been named the Advanced Reactor Technology Integral System Test (ARTIST) facility. Research performed in this facility will advance the state of the art and technology readiness level of high temperature intermediate heat exchangers (IHXs) for nuclear applications while establishing the INL as a center of excellence for the development and certification of this technology. The thermal energy storage capability will support research and demonstration activities related to process heat delivery for a variety of hybrid energy systems and grid stabilization strategies. Experimental results obtained from this research will assist in development of reliable predictive models for thermal hydraulic design and safety codes over the range of expected advanced reactor operating conditions. Proposed/existing IHX heat transfer and friction correlations and criteria will be assessed with information on materials compatibility and instrumentation

  13. Strategic need for a multi-purpose thermal hydraulic loop for support of advanced reactor technologies

    Energy Technology Data Exchange (ETDEWEB)

    O' Brien, James E. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sabharwall, Piyush [Idaho National Lab. (INL), Idaho Falls, ID (United States); Yoon, Su -Jong [Idaho National Lab. (INL), Idaho Falls, ID (United States); Housley, Gregory K. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-09-01

    This report presents a conceptual design for a new high-temperature multi fluid, multi loop test facility for the INL to support thermal hydraulic, materials, and thermal energy storage research for nuclear and nuclear-hybrid applications. In its initial configuration, the facility will include a high-temperature helium loop, a liquid salt loop, and a hot water/steam loop. The three loops will be thermally coupled through an intermediate heat exchanger (IHX) and a secondary heat exchanger (SHX). Research topics to be addressed with this facility include the characterization and performance evaluation of candidate compact heat exchangers such as printed circuit heat exchangers (PCHEs) at prototypical operating conditions, flow and heat transfer issues related to core thermal hydraulics in advanced helium-cooled and salt-cooled reactors, and evaluation of corrosion behavior of new cladding materials and accident-tolerant fuels for LWRs at prototypical conditions. Based on its relevance to advanced reactor systems, the new facility has been named the Advanced Reactor Technology Integral System Test (ARTIST) facility. Research performed in this facility will advance the state of the art and technology readiness level of high temperature intermediate heat exchangers (IHXs) for nuclear applications while establishing the INL as a center of excellence for the development and certification of this technology. The thermal energy storage capability will support research and demonstration activities related to process heat delivery for a variety of hybrid energy systems and grid stabilization strategies. Experimental results obtained from this research will assist in development of reliable predictive models for thermal hydraulic design and safety codes over the range of expected advanced reactor operating conditions. Proposed/existing IHX heat transfer and friction correlations and criteria will be assessed with information on materials compatibility and instrumentation

  14. Liquid metal reactor development. Development of LMR coolant technology

    Energy Technology Data Exchange (ETDEWEB)

    Nam, H. Y.; Choi, S. K.; Hwang, J. s.; Lee, Y. B.; Choi, B. H.; Kim, J. M.; Kim, Y. G.; Kim, M. J.; Lee, S. D.; Kang, Y. H.; Maeng, Y. Y.; Kim, T. R.; Park, J. H.; Park, S. J.; Cha, J. H.; Kim, D. H.; Oh, S. K.; Park, C. G.; Hong, S. H.; Lee, K. H.; Chun, M. H.; Moon, H. T.; Chang, S. H.; Lee, D. N.

    1997-07-15

    Following studies have been performed during last three years as the 1.2 phase study of the mid and long term nuclear technology development plan. First, the small scale experiments using the sodium have been performed such as the basic turbulent mixing experiment which is related to the design of a compact reactor, the flow reversal characteristics experiment by natural circulation which is necessary for the analysis of local flow reversal when the electromagnetic pump is installed, the feasibility test of the decay heat removal by wall cooling and the operation of electromagnetic pump. Second, the technology of operation mechanism of sodium facility is developed and the technical analysis and fundamental experiments of sodium measuring technology has been performed such as differential pressure measuring experiment, local flow rate measuring experimenter, sodium void fraction measuring experiment, under sodium facility, the free surface movement experiment and the side orifice pressure drop experiment. A new bounded convection scheme was introduced to the ELBO3D thermo-hydraulic computer code designed for analysis of experimental result. A three dimensional computer code was developed for the analysis of free surface movement and the analysis model of transmission of sodium void fraction was developed. Fourth, the small scale key components are developed. The submersible-in-pool type electromagnetic pump which can be used as primary pump in the liquid metal reactor is developed. The SASS which uses the Curie-point electromagnet and the mock-up of Pantograph type IVTM were manufactured and their feasibility was evaluated. Fifth, the high temperature characteristics experiment of stainless steel which is used as a major material for liquid metal reactor and the material characteristics experiment of magnet coil were performed. (author). 126 refs., 98 tabs., 296 figs.

  15. Development of system integration technology for integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Moon Hee; Kang, D. J.; Kim, K. K. and others

    1999-03-01

    The objective of this report is to integrate the conceptual design of an integral reactor, SMART producing thermal energy of 330 MW, which will be utilized to supply energy for seawater desalination and small-scale power generation. This project also aims to develop system integration technology for effective design of the reactor. For the conceptual design of SMART, preliminary design requirements including the top-tier requirements and design bases were evaluated and established. Furthermore, in the view of the application of codes and standards to the SMART design, existing laws, codes and standards were analyzed and evaluated with respect to its applicability. As a part of this evaluation, directions and guidelines were proposed for the development of new codes and standards which shall be applied to the SMART design. Regarding the integration of SMART conceptual designs, major design activities and interfaces between design departments were established and coordinated through the design process. For the effective management of all design schedules, a work performance evaluation system was developed and applied to the design process. As the results of this activity, an integrated output of SMART designs was produced. Two additional scopes performed in this project include the preliminary economic analysis on the SMART utilization for seawater desalination, and the planning of verification tests for technology implemented into SMART and establishing development plan of the computer codes to be used for SMART design in the next phase. The technical cooperation with foreign country and international organization for securing technologies for integral reactor design and its application was coordinated and managed through this project. (author)

  16. Status and future of the WWR-M research reactor in Kiev

    Energy Technology Data Exchange (ETDEWEB)

    Bazavov, D.A.; Gavrilyuk, V.I.; Kirischuk, V.I.; Kochetkov, V.V.; Lysenko, M.V.; Makarovskiy, V.N.; Scherbachenko, A.M.; Shevel, V.N.; Slisenko, V.I. [Institute for Nuclear Research, Kiev (Ukraine)

    2001-07-01

    Kiev WWR-M Research Reactor, operated at maximum power of 10 MW, was put into operation in 1960 and during its 40-years history has been used to perform numerous studies in different areas of science and technology. Due to a number of technical problems the Research Reactor, the only one in Ukraine, was shut down in 1993 and then put into operation in 1999 again. Now there is an intention to reconstruct Kiev Research Reactor. The upgraded Research Reactor would allow solving such problems as the safe operation of Ukrainian NPPs, radioisotope production and, naturally, fundamental and applied research. The main problem for the successful operation of Kiev Research Reactor is the management and storage of spent fuel at the site, since after core unloading the spent fuel storage appears to be practically completed. So it is absolutely necessary to ship the most part of the spent fuel for reprocessing and as soon as possible. Besides, there is a need to build up the new spent fuel storage, because the tank of available storage requires careful inspection for corrosion. (author)

  17. Research on physical and chemical parameters of coolant in Light-Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Reis, Isabela C.; Mesquita, Amir Z., E-mail: icr@cdtn.br, E-mail: amir@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEM-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    The coolant radiochemical monitoring of light-water reactors, both power reactor as research reactors is one most important tasks of the system safe operation. The last years have increased the interest in the coolant chemical studying to optimize the process, to minimize the corrosion, to ensure the primary system materials integrity, and to reduce the workers exposure radiation. This paper has the objective to present the development project in Nuclear Technology Development Center (CDTN), which aims to simulate the primary water physical-chemical parameters of light-water-reactors (LWR). Among these parameters may be cited: the temperature, the pressure, the pH, the electric conductivity, and the boron concentration. It is also being studied the adverse effects that these parameters can result in the reactor integrity. The project also aims the mounting of a system to control and monitoring of temperature, electric conductivity, and pH of water in the Installation of Test in Accident Conditions (ITCA), located in the Thermal-Hydraulic Laboratory at CDTN. This facility was widely used in the years 80/90 for commissioning of several components that were installed in Angra 2 containment. In the test, the coolant must reproduce the physical and chemical conditions of the primary. It is therefore fundamental knowledge of the main control parameters of the primary cooling water from PWR reactors. Therefore, this work is contributing, with the knowledge and the reproduction with larger faithfulness of the reactors coolant in the experimental circuits. (author)

  18. Research and Development Methodology for Practical Use of Accident Tolerant Fuel in Light Water Reactors

    OpenAIRE

    Masaki Kurata

    2016-01-01

    Research and development (R&D) methodology for the practical use of accident tolerant fuel (ATF) in commercial light water reactors is discussed in the present review. The identification and quantification of the R&D-metrics and the attribute of candidate ATF-concepts, recognition of the gap between the present R&D status and the targeted practical use, prioritization of the R&D, and technology screening schemes are important for achieving a common understanding on technology screening proces...

  19. The rehabilitation/upgrading of Philippine Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Renato, T. Banaga [Philippines Nuclear Research Inst., Quezon (Philippines)

    1998-10-01

    The Philippine Research Reactor (PRR-1) is the only research reactor in the Philippines. It was acquired through the Bilateral Agreement with the United States of America. The General Electric (G.E.) supplied PRR-1 first become operational in 1963 and used MTR plate type fuel. The original one-megawatt G.E. reactor was shutdown and converted into a 3 MW TRIGA PULSING REACTOR in 1984. The conversion includes the upgrading of the cooling system, replacement of new reactor coolant pumps, heat exchanger, cooling tower, replacement of new nuclear instrumentation and standard TRIGA console, TRIGA fuel supplied by General Atomic (G.A.). Philippine Nuclear Research Institute (PNRI) provided the old reactor, did the detailed design of the new cooling system, provided the new non-nuclear instrumentation and electrical power supply system and performed all construction, installation and modification work on site. The TRIGA conversion fuel is contained in a shrouded 4-rod cluster which fit into the original grid plate. The new fuel is a E{sub 1}-U-Z{sub 1}-H{sub 1.6} TRIGA fuel, has a 20% wt Uranium loading with 19.7% U-235 enrichment and about 0.5 wt % Erbium. The Start-up, calibration and Demonstration of Pulsing and Full Power Operation were completed during a three week start-up phase which were performed last March 1968. A few days after, a leak in the pool liner was discovered. The reactor was shutdown again for repair and up to present the reactor is still in the process of rehabilitation. This paper will describe the rehabilitation/upgrading done on the PRR-1 since 1988 up to present. (author)

  20. Component-Level Prognostics Health Management Framework for Passive Components - Advanced Reactor Technology Milestone: M2AT-15PN2301043

    Energy Technology Data Exchange (ETDEWEB)

    Ramuhalli, Pradeep; Roy, Surajit; Hirt, Evelyn H.; Prowant, Matthew S.; Pitman, Stan G.; Tucker, Joseph C.; Dib, Gerges; Pardini, Allan F.

    2015-06-19

    This report describes research results to date in support of the integration and demonstration of diagnostics technologies for prototypical advanced reactor passive components (to establish condition indices for monitoring) with model-based prognostics methods. Achieving this objective will necessitate addressing several of the research gaps and technical needs described in previous technical reports in this series.

  1. Analysis on approach of safeguards implementation at research reactor handling item count and bulk material

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyun Jo; Lee, Sung Ho; Lee, Byung Doo; Jung, Juang [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    KiJang research reactor (KJRR) will be constructed to produce the radioisotope such as Mo-99 etc., provide the neutron transmutation doping (NTD) service of silicon, and develop the core technologies of research reactor. In this paper, the features of the process and nuclear material flow are reviewed and the material balance area (MBA) and key measurement point (KMP) are established based on the nuclear material flow. Also, this paper reviews the approach on safeguards and nuclear material accountancy at the facility level for Safeguards-by-Design at research reactor handling item count and bulk material. In this paper, MBA and KMPs are established through the analysis on facility features and major process at KJRR handling item count and bulk material. Also, this paper reviews the IAEA safeguards implementation and nuclear material accountancy at KJRR. It is necessary to discuss the safeguards approach on the fresh FM target assemblies and remaining uranium in the intermediate level liquid wastes.

  2. Report from the Light Water Reactor Sustainability Workshop on On-Line Monitoring Technologies

    Energy Technology Data Exchange (ETDEWEB)

    Thomas Baldwin; Magdy Tawfik; Leonard Bond

    2010-06-01

    In support of expanding the use of nuclear power, interest is growing in methods of determining the feasibility of longer term operation for the U.S. fleet of nuclear power plants, particularly operation beyond 60 years. To help establish the scientific and technical basis for such longer term operation, the DOE-NE has established a research and development (R&D) objective. This objective seeks to develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of current reactors. The Light Water Reactor Sustainability (LWRS) Program, which addresses the needs of this objective, is being developed in collaboration with industry R&D programs to provide the technical foundations for licensing and managing the long-term, safe, and economical operation of nuclear power plants. The LWRS Program focus is on longer-term and higher-risk/reward research that contributes to the national policy objectives of energy and environmental security. In moving to identify priorities and plan activities, the Light Water Reactor Sustainability Workshop on On-Line Monitoring (OLM) Technologies was held June 10–12, 2010, in Seattle, Washington. The workshop was run to enable industry stakeholders and researchers to identify the nuclear industry needs in the areas of future OLM technologies and corresponding technology gaps and research capabilities. It also sought to identify approaches for collaboration that would be able to bridge or fill the technology gaps. This report is the meeting proceedings, documenting the presentations and discussions of the workshop and is intended to serve as a basis for a plan which is under development that will enable the I&C research pathway to achieve its goals. Benefits to the nuclear industry accruing from On Line Monitoring Technology cannot be ignored. Information gathered thus far has contributed significantly to the Department of Energy’s Light Water Reactor Sustainability Program. DOE has

  3. Research on Key Technology of Small Natural Ci rculation Fast Reactor Integrated with AMTEC%结合AMTEC的小型自然循环快堆的关键技术研究

    Institute of Scientific and Technical Information of China (English)

    陆道纲; 张勋; 李宗洋; 郭超

    2016-01-01

    The conceptual design of a small modular reactor (SMR) integrated with alkali metal thermal to electric converter (AMTEC ) units , called SMR‐AMTEC system ,was proposed internationally in order to meet the reliability in power supplies subjected to extremes of operating conditions . Three key technologies of the small modular reactor in the conceptual design were developed in the paper .They were reactor core physical control technology based on rotating drums , this primary loop natural circulation technology under normal power and the heat removal technology based on natural circulation .T he preparation technologies of three key components of the small multi‐tube circulating AMTEC unit coupled with the conceptual design were particularly developed .They were fabrication technology of porous TiN film electrode of AMTEC , the sealing technology of BASE (Beta″‐alumina solid electrolytes ) components and the development and test technologies of wick components . The feasibility of the SMR‐AMTEC system is preliminarily verified through R&D of the technologies above .%为满足远距离无人值守化等极端环境下的电源可靠供给,本文提出了一种结合碱金属热电转换器(AMTEC)的小型模块化反应堆(SMR)的概念,即SMR‐AMTEC系统。针对该小型模块化反应堆的概念设计,本文研发了3项关键技术,即:基于转鼓的堆物理控制技术;正常功率条件下一回路全自然循环技术;基于自然循环的余热排出技术。针对与该小型模块化反应堆相耦合的小型多管循环式 A M T EC单元,本文重点开展了3项关键部件制备技术的研发,即:AMTEC的TiN多孔薄膜电极制备技术;β″氧化铝固体电解质组件封接技术;吸液芯组件的制备及测试技术。通过对以上技术的研究与开发,初步验证了SMR‐AMTEC系统的可行性。

  4. Operating manual for the Health Physics Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    1985-11-01

    This manual is intended to serve as a guide in the operation and maintenance of the Health Physics Researh Reactor (HPRR) of the Health Physics Dosimetry Applications Research (DOSAR) Facility. It includes descriptions of the HPRR and of associated equipment such as the reactor positioning devises and the derrick. Procedures for routine operation of the HPRR are given in detail, and checklists for the various steps are provided where applicable. Emergency procedures are similarly covered, and maintenance schedules are outlined. Also, a bibliography of references giving more detailed information on the DOSAR Facility is included. Changes to this manual will be approved by at least two of the following senior staff members: (1) the Operations Division Director, (2) the Reactor Operations Department Head, (3) the Supervisor of Reactor Operations TSF-HPRR Areas. The master copy and the copy of the manual issued to the HPRR Operations Supervisor will always reflect the latest revision. 22 figs.

  5. Implementation Plan for Qualification of Sodium-Cooled Fast Reactor Technology Information

    Energy Technology Data Exchange (ETDEWEB)

    Moe, Wayne [Idaho National Lab. (INL), Idaho Falls, ID (United States); Honma, George [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-03-01

    This document identifies and discusses implementation elements that can be used to facilitate consistent and systematic evaluation processes relating to quality attributes of technical information (with focus on SFR technology) that will be used to support licensing of advanced reactor designs. Information may include, but is not limited to, design documents for SFRs, research-and-development (R&D) data and associated documents, test plans and associated protocols, operations and test data, international research data, technical reports, and information associated with past U.S. Nuclear Regulatory Commission (NRC) reviews of SFR designs. The approach for determining acceptability of test data, analysis, and/or other technical information is based on guidance provided in INL/EXT-15-35805, “Guidance on Evaluating Historic Technology Information for Use in Advanced Reactor Licensing.” The implementation plan can be adopted into a working procedure at each of the national laboratories performing data qualification, or by applicants seeking future license application for advanced reactor technology.

  6. Development of fluid system design technology for integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, D. J.; Chang, M. H.; Kang, D. J. and others

    1999-03-01

    This study presents the technology development of the system design concepts of SMART, a multi-purposed integral reactor with enhanced safety and operability, for use in diverse usages and applications of the nuclear energy. This report contains the following; - Design characteristics - Performance and safety related design criteria - System description: Primary system, Secondary system, Residual heat removal system, Make-up system, Component cooling system, Safety system - Development of design computer code: Steam generator performance(ONCESG), Pressurizer performance(COLDPZR), Steam generator flow instability(SGINS) - Development of component module and modeling using MMS computer code - Design calculation: Steam generator thermal sizing, Analysis of feed-water temperature increase at a low flow rate, Evaluation of thermal efficiency in the secondary system, Inlet orifice throttling coefficient for the prevention of steam generator flow instability, Analysis of Nitrogen gas temperature in the pressurizer during heat-up process, evaluation of water chemistry and erosion etc. The results of this study can be utilized not only for the foundation technology of the next phase basic system design of the SMART but also for the basic model in optimizing the system concepts for future advanced reactors. (author)

  7. Reduced enrichment for research and test reactors: Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    1988-05-01

    The international effort to develop new research reactor fuel materials and designs based on the use of low-enriched uranium, instead of highly-enriched uranium, has made much progress during the eight years since its inception. To foster direct communication and exchange of ideas among the specialist in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at the Argonne National Laboratory, sponsored this meeting as the ninth of a series which began in 1978. All previous meetings of this series are listed on the facing page. The focus of this meeting was on the LEU fuel demonstration which was in progress at the Oak Ridge Research (ORR) reactor, not far from where the meeting was held. The visit to the ORR, where a silicide LEU fuel with 4.8 g A/cm/sup 3/ was by then in routine use, illustrated how far work has progressed.

  8. A probabilistic safety analysis of incidents in nuclear research reactors.

    Science.gov (United States)

    Lopes, Valdir Maciel; Agostinho Angelo Sordi, Gian Maria; Moralles, Mauricio; Filho, Tufic Madi

    2012-06-01

    This work aims to evaluate the potential risks of incidents in nuclear research reactors. For its development, two databases of the International Atomic Energy Agency (IAEA) were used: the Research Reactor Data Base (RRDB) and the Incident Report System for Research Reactor (IRSRR). For this study, the probabilistic safety analysis (PSA) was used. To obtain the result of the probability calculations for PSA, the theory and equations in the paper IAEA TECDOC-636 were used. A specific program to analyse the probabilities was developed within the main program, Scilab 5.1.1. for two distributions, Fischer and chi-square, both with the confidence level of 90 %. Using Sordi equations, the maximum admissible doses to compare with the risk limits established by the International Commission on Radiological Protection (ICRP) were obtained. All results achieved with this probability analysis led to the conclusion that the incidents which occurred had radiation doses within the stochastic effects reference interval established by the ICRP-64.

  9. Prevention of ageing of research reactors by design

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-12-31

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

  10. Radiation protection personnel training in Research Reactors; Capacitacion en proteccion radiologica para reactores de investigacion

    Energy Technology Data Exchange (ETDEWEB)

    Fernandez, Carlos Dario; Lorenzo, Nestor Pedro de [Comision Nacional de Energia Atomica, Rio Negro (Argentina). Centro Atomico Bariloche. Instituto Balseiro

    1996-07-01

    The RA-6 research reactor is considering the main laboratory in the training of different groups related with radiological protection. The methodology applied to several courses over 15 years of experience is shown in this work. The reactor is also involved in the construction, design, start-up and sell of different installation outside Argentina for this reason several theoretical and practical courses had been developed. The acquired experience obtained is shown in this paper and the main purpose is to show the requirements to be taken into account for every group (subjects, goals, on-job training, etc) (author)

  11. RADIATION DOSIMETRY AT THE BNL HIGH FLUX BEAM REACTOR AND MEDICAL RESEARCH REACTOR.

    Energy Technology Data Exchange (ETDEWEB)

    HOLDEN,N.E.

    1999-09-10

    RADIATION DOSIMETRY MEASUREMENTS HAVE BEEN PERFORMED OVER A PERIOD OF MANY YEARS AT THE HIGH FLUX BEAM REACTOR (HFBR) AND THE MEDICAL RESEARCH REACTOR (BMRR) AT BROOKHAVEN NATIONAL LABORATORY TO PROVIDE INFORMATION ON THE ENERGY DISTRIBUTION OF THE NEUTRON FLUX, NEUTRON DOSE RATES, GAMMA-RAY FLUXES AND GAMMA-RAY DOSE RATES. THE MCNP PARTICLE TRANSPORT CODE PROVIDED MONTE CARLO RESULTS TO COMPARE WITH VARIOUS DOSIMETRY MEASUREMENTS PERFORMED AT THE EXPERIMENTAL PORTS, AT THE TREATMENT ROOMS AND IN THE THIMBLES AT BOTH HFBR AND BMRR.

  12. LIGHT WATER REACTOR SUSTAINABILITY PROGRAM ADVANCED INSTRUMENTATION, INFORMATION, AND CONTROL SYSTEMS TECHNOLOGIES TECHNICAL PROGRAM PLAN FOR 2013

    Energy Technology Data Exchange (ETDEWEB)

    Hallbert, Bruce; Thomas, Ken

    2014-07-01

    Reliable instrumentation, information, and control (II&C) systems technologies are essential to ensuring safe and efficient operation of the U.S. light water reactor (LWR) fleet. These technologies affect every aspect of nuclear power plant (NPP) and balance-of-plant operations. In 1997, the National Research Council conducted a study concerning the challenges involved in modernization of digital instrumentation and control systems in NPPs. Their findings identified the need for new II&C technology integration.

  13. Light Water Reactor Sustainability Program Advanced Instrumentation, Information, and Control Systems Technologies Technical Program Plan for FY 2016

    Energy Technology Data Exchange (ETDEWEB)

    Hallbert, Bruce Perry [Idaho National Laboratory; Thomas, Kenneth David [Idaho National Laboratory

    2015-10-01

    Reliable instrumentation, information, and control (II&C) systems technologies are essential to ensuring safe and efficient operation of the U.S. light water reactor (LWR) fleet. These technologies affect every aspect of nuclear power plant (NPP) and balance-of-plant operations. In 1997, the National Research Council conducted a study concerning the challenges involved in modernization of digital instrumentation and control systems in NPPs. Their findings identified the need for new II&C technology integration.

  14. Light Water Reactor Sustainability Program Advanced Instrumentation, Information, and Control Systems Technologies Technical Program Plan for 2013

    Energy Technology Data Exchange (ETDEWEB)

    Hallbert, Bruce [Idaho National Laboratory (INL), Idaho Falls, ID (United States); Thomas, Ken [Idaho National Laboratory (INL), Idaho Falls, ID (United States)

    2014-09-01

    Reliable instrumentation, information, and control (II&C) systems technologies are essential to ensuring safe and efficient operation of the U.S. light water reactor (LWR) fleet. These technologies affect every aspect of nuclear power plant (NPP) and balance-of-plant operations. In 1997, the National Research Council conducted a study concerning the challenges involved in modernization of digital instrumentation and control systems in NPPs. Their findings identified the need for new II&C technology integration.

  15. Reactor Safety Research: Semiannual report, July-December 1986

    Energy Technology Data Exchange (ETDEWEB)

    1987-11-01

    Sandia National Laboratories is conducting, under USNRC sponsorship, phenomenological research related to the safety of commercial nuclear power reactors. The research includes experiments to simulate the phenomenology of the accident conditions and the development of analytical models, verified by experiment, which can be used to predict reactor and safety systems performance and behavior under abnormal conditions. The objective of this work is to provide NRC requisite data bases and analytical methods to (1) identify and define safety issues, (2) understand the progression of risk-significant accident sequences, and (3) conduct safety assessments. The collective NRC-sponsored effort at Sandia National Laboratories is directed at enhancing the tehcnology base supporting licensing decisions.

  16. SoLid: Search for Oscillation with a 6Li Detector at the BR2 research reactor

    CERN Document Server

    Michiels, Ianthe

    2016-01-01

    In the past decades, various nuclear reactor neutrino experiments have measured a deficit in the flux of antineutrinos coming from the reactor at short reactor-detector distances, when compared to theoretical calculations. One of the experiments designed to investigate this reactor antineutrino anomaly is the SoLid experiment. It uses the compact BR2 research reactor from the SCK-CEN in Mol, Belgium, to perform reactor antineutrino flux measurements at very short baseline. These proceedings discuss the general detection concepts of the SoLid experiment and its novel detector technology. The performance of the SoLid design is demonstrated with some results of the analysis of the data gathered with the experiment's first large scale test module, SM1.

  17. Decontamination and decommissioning project of the TRIGA Mark-2 and 3 research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Jung, K. J.; Baik, S. T.; Chung, U. S.; Jung, K. H.; Park, S. K.; Lee, B. J.; Kim, J. K.; Yang, S. H

    2000-01-01

    During the review on the decommissioning plan and environmental impact assessment report by the KINS, the number of the inquired items were two hundred and fifty one, and the answers were made and sent until September 10, 1999, as the screened review results were reported to Ministry of Science and Technology(MOST) in December 14, 1999, all the reviews on the licence were over. Radioactive liquid wastes of 400 tons generated during the operation of the research reactors including reactor vessels are stored in the facility of the research reactor 1 and 2. Those liquid wastes have the low-level-radioactivity which can be discharged to the surroundings, but was wholly treated to be vaporized naturally by means of the increased numbers of the natural vaporization disposal facilities with the annual capacity of 200 tons for the purpose of the minimized environmental contamination.

  18. Pilot-scale tests of an innovative 'serial self-turning reactor' composting technology in Thailand.

    Science.gov (United States)

    Sungsomboon, Praj-ya; Chaisomphob, Taweep; Bongochgetsakul, Nattakorn; Ishida, Tetsuya

    2013-02-01

    Composting facilities in Thailand have faced various operational problems, resulting in the emission of odours, incomplete digestion of waste organics, and higher than desired costs. Composting technologies imported from developed countries tend to be sized for larger communities and are otherwise not suited for the rural communities that comprise more than 80% of all communities in Thailand. This article addresses the research and development of a novel composting technology aimed at filling this observed need. The study was divided into two parts: (1) the development of a new composting technology and fabrication of a prototype configuration of equipment; and (2) scale-up and study on a pilot-scale using real rubbish. The proposed technology, called 'serial self-turning reactor (STR)', entailed a vertical flow composting system that consisted of a set of aerobic reactors stacked on a set of gravity fed turning units. In-vessel bioreactor technology enables the operator to control composting conditions. The researchers found that a tower-like STR results in flexibility in size scale and waste processing residence time. The pilot-scale experiments showed that the proposed system can produce good quality compost while consuming comparatively little energy and occupying a compact space, compared to traditional land-intensive windrow composting facilities.

  19. Fuel burnup analysis for Thai research reactor by using MCNPX computer code

    Science.gov (United States)

    Sangkaew, S.; Angwongtrakool, T.; Srimok, B.

    2017-06-01

    This paper presents the fuel burnup analysis of the Thai research reactor (TRR-1/M1), TRIGA Mark-III, operated by Thailand Institute of Nuclear Technology (TINT) in Bangkok, Thailand. The modelling software used in this analysis is MCNPX (MCNP eXtended) version 2.6.0, a Fortran90 Monte Carlo radiation transport computer code. The analysis results will cover the core excess reactivity, neutron fluxes at the irradiation positions and neutron detector tubes, power distribution, fuel burnup, and fission products based on fuel cycle of first reactor core arrangement.

  20. Replacement Nuclear Research Reactor: Draft Environmental Impact Statement. Vol. 1. Main report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-07-01

    The Draft Environmental Impact Statement (EIS) for the replacement of the Australian Research reactor has been released. An important objective of the EIS process is to ensure that all relevant information has been collected and assessed so that the Commonwealth Government can make an informed decision on the proposal. The environmental assessment of the proposal to construct and operate a replacement reactor described in the Draft EIS has shown that the scale of environmental impacts that would occur would be acceptable, provided that the management measures and commitments made by ANSTO are adopted. Furthermore, construction and operation of the proposed replacement reactor would result in a range of benefits in health care, the national interest, scientific achievement and industrial capability. It would also result in a range of benefits derived from increased employment and economic activity. None of the alternatives to the replacement research reactor considered in the Draft EIS can meet all of the objectives of the proposal. The risk from normal operations or accidents has been shown to be well within national and internationally accepted risk parameters. The dose due to reactor operations would continue to be small and within regulatory limits. For the replacement reactor, the principle of `As Low As Reasonably Achievable` would form an integral part of the design and licensing process to ensure that doses to operators are minimized. Costs associated with the proposal are $286 million (in 1997 dollars) for design and construction. The annual operating and maintenance costs are estimated to be $12 million per year, of which a significant proportion will be covered by commercial activities. The costs include management of the spent fuel from the replacement reactor as well as the environmental management costs of waste management, safety and environmental monitoring. Decommissioning costs for the replacement reactor would arise at the end of its lifetime

  1. Hafnium as a neutron absorber in research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Jatuff, F.E.; Villarino, E.A. [INVAP S.E., S.C. de Bariloche (Argentina)

    1995-12-31

    Research reactors usually require the periodic replacement of control rods because of the burnup of the neutron absorber material, even if devices have been mainly withdrawn. The purpose of this paper is the study of the reactivity worth of hafnium as a neutron absorber in the multi-purpose-type research reactor (MPTRR). The MPTRR design develops 20 MW, and it is configured by five rows of six low-enriched uranium (200% enrichment) materials test reactor fuels inside a Zircaloy chimney. The standard control plates are given by two rows of three plates of silver-indium-cadmium alloy (85% silver, 10% indium, and 5% cadmium, in weight percents; 0.26 cm of maximum absorber width), separating the central three rows of fuel elements from the peripheral first and fifth rows. The burnup of control plates decreases the reactivity worth below the licensing criteria in {approximately}6 months.

  2. Neutron flux optimization in irradiation channels at NUR research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Meftah, B. [Division Reacteur, Centre de Recherche Nucleaire Draria (CRND), BP 43 Sebala DRARIA, Alger (Algeria)]. E-mail: b_meftah@yahoo.com; Zidi, T. [Division Reacteur, Centre de Recherche Nucleaire Draria (CRND), BP 43 Sebala DRARIA, Alger (Algeria); Bousbia-Salah, A. [Dipartimento di Ingegneria Meccanica, Nucleari e della Produzione, Facolta di Ingegneria, Universita di Pisa, Via Diotisalvi, 2 - 56126 Pisa (Italy)

    2006-09-15

    Optimization of neutron fluxes in experimental channels is of great concern in research reactor utilization. The general approach used at the NUR research reactor for neutron flux optimization in irradiation channels is presented. The approach is essentially based upon a judicious optimization of the core configuration combined with the improvement of reflector characteristics. The method allowed to increase the thermal neutron flux for radioisotope production purposes by more than 800%. Increases of up to 60% are also observed in levels of useful fluxes available for neutron diffraction experiments (small angle neutron scattering (SANS), neutron reflectometry, etc.). Such improvements in the neutronic characteristics of the NUR reactor opened new perspectives in terms of its utilization. More particularly, it is now possible to produce at industrial scales major radio-isotopes for medicine and industry and to perform, for the first time, material testing experiments. The cost of the irradiations in the optimized configuration is generally small when compared to those performed in the old configuration and an average reduction factor of about of 10 is expected in the case of production of Molybdenum-99 (isotope required for the manufacturing of Technetium-99 medical kits). In addition to these important results, safety analysis studies showed that the more symmetrical nature of the core geometry leads to a more adequately balanced reactivity control system and contributes quite efficiently to the operational safety of the NUR reactor. Results of comparisons between calculations and measurements for a series of parameters of importance in reactor operation and safety showed good agreement.

  3. Experimental Design for Evaluating Selected Nondestructive Measurement Technologies - Advanced Reactor Technology Milestone: M3AT-16PN2301043

    Energy Technology Data Exchange (ETDEWEB)

    Ramuhalli, Pradeep [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Hirt, Evelyn H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Pitman, Stan G. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Dib, Gerges [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Roy, Surajit [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Good, Morris S. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Walker, Cody M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2016-07-16

    The harsh environments in advanced reactors (AdvRx) increase the possibility of degradation of safety-critical passive components, and therefore pose a particular challenge for deployment and extended operation of these concepts. Nondestructive evaluation technologies are an essential element for obtaining information on passive component condition in AdvRx, with the development of sensor technologies for nondestructively inspecting AdvRx passive components identified as a key need. Given the challenges posed by AdvRx environments and the potential needs for reducing the burden posed by periodic in-service inspection of hard-to-access and hard-to-replace components, a viable solution may be provided by online condition monitoring of components. This report identifies the key challenges that will need to be overcome for sensor development in this context, and documents an experimental plan for sensor development, test, and evaluation. The focus of initial research and development is on sodium fast reactors, with the eventual goal of the research being developing the necessary sensor technology, quantifying sensor survivability and long-term measurement reliability for nondestructively inspecting critical components. Materials for sensor development that are likely to withstand the harsh environments are described, along with a status on the fabrication of reference specimens, and the planned approach for design and evaluation of the sensor and measurement technology.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-06-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-06-01

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

  6. Energy-technological complex with reactor for torrefaction

    Science.gov (United States)

    Kuzmina, J. S.; Director, L. B.; Zaichenko, V. M.

    2016-11-01

    To eliminate shortcomings of raw plant materials pelletizing process with thermal treatment (low-temperature pyrolysis or torrefaction) can be applied. This paper presents a mathematical model of energy-technological complex (ETC) for combined production of heat, electricity and solid biofuels torrefied pellets. According to the structure the mathematical model consists of mathematical models of main units of ETC and the relationships between them and equations of energy and material balances. The equations describe exhaust gas straining action through a porous medium formed by pellets. Decomposition rate of biomass was calculated by using the gross-reaction diagram, which is responsible for the disintegration of raw material. A mathematical model has been tested according to bench experiments on one reactor module. From nomographs, designed for a particular configuration of ETC it is possible to determine the basic characteristics of torrefied pellets (rate of weight loss, heating value and heat content) specifying only two parameters (temperature and torrefaction time). It is shown that the addition of reactor for torrefaction to gas piston engine can improve the energy efficiency of power plant.

  7. Defining the "proven technology" technical criterion in the reactor technology assessment for Malaysia's nuclear power program

    Science.gov (United States)

    Anuar, Nuraslinda; Kahar, Wan Shakirah Wan Abdul; Manan, Jamal Abdul Nasir Abd

    2015-04-01

    Developing countries that are considering the deployment of nuclear power plants (NPPs) in the near future need to perform reactor technology assessment (RTA) in order to select the most suitable reactor design. The International Atomic Energy Agency (IAEA) reported in the Common User Considerations (CUC) document that "proven technology" is one of the most important technical criteria for newcomer countries in performing the RTA. The qualitative description of five desired features for "proven technology" is relatively broad and only provides a general guideline to its characterization. This paper proposes a methodology to define the "proven technology" term according to a specific country's requirements using a three-stage evaluation process. The first evaluation stage screens the available technologies in the market against a predefined minimum Technology Readiness Level (TRL) derived as a condition based on national needs and policy objectives. The result is a list of technology options, which are then assessed in the second evaluation stage against quantitative definitions of CUC desired features for proven technology. The potential technology candidates produced from this evaluation is further narrowed down to obtain a list of proven technology candidates by assessing them against selected risk criteria and the established maximum allowable total score using a scoring matrix. The outcome of this methodology is the proven technology candidates selected using an accurate definition of "proven technology" that fulfills the policy objectives, national needs and risk, and country-specific CUC desired features of the country that performs this assessment. A simplified assessment for Malaysia is carried out to demonstrate and suggest the use of the proposed methodology. In this exercise, ABWR, AP1000, APR1400 and EPR designs assumed the top-ranks of proven technology candidates according to Malaysia's definition of "proven technology".

  8. Instructional Technology: The Research Field.

    Science.gov (United States)

    Gagne, Robert M.

    1986-01-01

    Reflects upon opportunities for research in instructional technology provided by present state of media hardware technology and educational requirements. Prospects for research in incidental learning, including learning from television, are discussed, as well as traditional learning research on intentional learning, including possibilities for…

  9. China Advanced Research Reactor Project Progress in 2012

    Institute of Scientific and Technical Information of China (English)

    ZHAO; Tie-jun

    2012-01-01

    <正>In 2012, all the commissioning for the China Advanced Research Reactor (CARR) had been finished and the diffraction pattern had been successfully obtained on the neutron scattering spectrometer. Meanwhile, the cold neutron source project and the acceptance items of CARR project had been carrying out.

  10. Neutron spectrometric methods for core inventory verification in research reactors

    CERN Document Server

    Ellinger, A; Hansen, W; Knorr, J; Schneider, R

    2002-01-01

    In consequence of the Non-Proliferation Treaty safeguards, inspections are periodically made in nuclear facilities by the IAEA and the EURATOM Safeguards Directorate. The inspection methods are permanently improved. Therefore, the Core Inventory Verification method is being developed as an indirect method for the verification of the core inventory and to check the declared operation of research reactors.

  11. 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.

  12. Inorganic membrane reactor technology CRADA {number_sign}1176; Final report and assessment of membrane technology

    Energy Technology Data Exchange (ETDEWEB)

    Schwartz, R.W.; Collins, J.P.; Ng, M.F. [and others

    1997-04-01

    This project focused on the fabrication and evaluation of supported inorganic membranes for hydrogen and oxygen separation in petrochemical processes. A variety of fabrication techniques, including CVD (Chemical Vapor Deposition), electroless plating, solution deposition and conventional ceramic processing methods were used for membrane fabrication. For the oxygen separation membrane materials studied, the high surface roughness of the commercially available (and chemically compatible) MgO supports for high flux oxygen materials (SrCo{sub 0.5}FeO{sub x} and SrCo{sub 0.8}Fe{sub 0.2}O{sub x}) hindered the development of supported membranes of these materials. More encouraging results were obtained for the supported hydrogen separation membranes. Both dense palladium (prepared by CVD and electroless plating) and ultramicroporous silica (prepared by solution deposition) membranes were fabricated onto porous alumina supports. Gas separation characteristics and reactor performance of the membranes were both studied. Of the two classes of membranes, when incorporated into a membrane reactor the silica membranes demonstrated the best performance. Propane and isobutane dehydrogenation processes were studied and the silica membrane reactors displayed modest improvements in performance compared to the conventional reactors. In propane dehydrogenation, an increase in propylene yield of 34% was obtained with the membrane reactor (compared to the conventional reactor); in isobutane dehydrogenation, an increase in isobutylene yield of 40% at 525 C was obtained. However, these performance gains decreased somewhat with time on stream, due to membrane instability. Further improvements in membrane stability and permselectivity, as well as catalyst stability are needed before membrane reactors can be considered as a realistic alternative to the existing conventional technology.

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

    CERN Document Server

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

    1990-01-01

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

  14. Fuel shuffling optimization for the Delft research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Geemert, R. van; Hoogenboom, J.E.; Gibcus, H.P.M. [Delft Univ. of Technology, Interfaculty Reactor Inst., Delft (Netherlands); Quist, A.J. [Delft Univ., Fac. of Applied Mathematics and Informatics, Delft (Netherlands)

    1997-07-01

    A fuel shuffling optimization procedure is proposed for the Hoger Onderwijs Reactor (HOR) in Delft, the Netherlands, a 2 MWth swimming-pool type research reactor. In order to cope with the fluctuatory behaviour of objective functions in loading pattern optimization, the proposed cyclic permutation optimization procedure features a gradual transition from global to local search behaviour via the introduction of stochastic tests for the number of fuel assemblies involved in a cyclic permutation. The possible objectives and the safety and operation constraints, as well as the optimization procedure, are discussed, followed by some optimization results for the HOR. (author)

  15. Liquid metal reactor development. Development of LMR design technology

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Cheol; Kim, Y. I.; Kim, Y. G.; Kim, E. K.; Song, H.; Chung, H. T.; Sim, Y. S.; Min, B. T.; Kim, Y. S.; Wi, M. H.; Yoo, B.; Lee, J. H.; Lee, H. Y.; Kim, J. B.; Koo, G. H.; Hahn, D. H.; Na, B. C.; Hwang, W.; Nam, C.; Ryu, W. S.; Lim, G. S.; Kim, D. H.; Kim, J. D.; Gil, C. S.

    1997-07-01

    This project was performed in five parts, the scope and contents of which are as follows: The nuclear data processing system was established and the KFS group constant library was improved and verified. Basic computation system was constructed by either developing or adding its function. Input/output (I/O) interface processing was developed to establish an integrated calculation system for LMR core nuclear rand thermal-hydraulic design and analysis. An experimental data analysis was performed to validate the constructed core neutronic calculation system. Using the established core calculation system and design technology, preliminary core design and performance analysis on the domestic LMR core design concept were carried out. To develop the basic technology of the LMR system analysis, LMR system behavior characteristics evaluation, thermal -fluid system analysis in the reactor pool, preliminary overall plant analysis and computer codes development have been performed. A porous model and simple one-dimensional model have been evaluated for the reactor pool analysis. The evaluation of the residual heat removal system on different design concepts has been also conducted. For the development of high temperature structural analysis, the heat transfer and thermal stress analyses were performed using finite element program with user subroutine that has been developed with an implementation of the Chaboche constitutive model for inelastic analysis capability, and the evaluation of creep-fatigue and ratcheting behavior of high temperature structure was carried out using this program. for development of the seismic isolation system and to predict the shear behavior for the laminated rubber bearing were established. And the behavior tests of isolation bearing and rubber specimens were carried out, and the seismic response tests for the isolation model structure were performed using the 30 ton shaking table. (author). 369 refs., 119 tabs., 320 figs.

  16. ENGINEERING DEVELOPMENT OF SLURRY BUBBLE COLUMN REACTOR (SBCR) TECHNOLOGY

    Energy Technology Data Exchange (ETDEWEB)

    Bernard A. Toseland, Ph.D.

    2002-01-01

    The major technical objectives of this program are threefold: (1) to develop the design tools and a fundamental understanding of the fluid dynamics of a slurry bubble column reactor to maximize reactor productivity, (2) to develop the mathematical reactor design models and gain an understanding of the hydrodynamic fundamentals under industrially relevant process conditions, and (3) to develop an understanding of the hydrodynamics and their interaction with the chemistries occurring in the bubble column reactor. Successful completion of these objectives will permit more efficient usage of the reactor column and tighter design criteria, increase overall reactor efficiency, and ensure a design that leads to stable reactor behavior when scaling up to large diameter reactors.

  17. Engineering Development of Slurry Bubble Column Reactor (SBCR) Technology

    Energy Technology Data Exchange (ETDEWEB)

    Toseland, B.A.

    1998-10-29

    The major technical objectives of this program are threefold: (1) to develop the design tools and a fundamental understanding of the fluid dynamics of a slurry bubble column reactor to maximize reactor productivity, (2) to develop the mathematical reactor design models and gain an understanding of the hydrodynamic fundamentals under industrially relevant process conditions, and (3) to develop an understanding of the hydrodynamics and their interaction with the chemistries occurring in the bubble column reactor. Successful completion of these objectives will permit more efficient usage of the reactor column and tighter design criteria, increase overall reactor efficiency, and ensure a design that leads to stable reactor behavior when scaling up to large diameter reactors.

  18. ENGINEERING DEVELOPMENT OF SLURRY BUBBLE COLUMN REACTOR (SBCR) TECHNOLOGY

    Energy Technology Data Exchange (ETDEWEB)

    Bernard A. Toseland

    2002-09-30

    The major technical objectives of this program are threefold: (1) to develop the design tools and a fundamental understanding of the fluid dynamics of a slurry bubble column reactor to maximize reactor productivity, (2) to develop the mathematical reactor design models and gain an understanding of the hydrodynamic fundamentals under industrially relevant process conditions, and (3) to develop an understanding of the hydrodynamics and their interaction with the chemistries occurring in the bubble column reactor. Successful completion of these objectives will permit more efficient usage of the reactor column and tighter design criteria, increase overall reactor efficiency, and ensure a design that leads to stable reactor behavior when scaling up to large diameter reactors.

  19. Iaea Activities Supporting the Applications of Research Reactors in 2013

    Science.gov (United States)

    Peld, Nathan D.; Ridikas, Danas

    2014-02-01

    As the underutilization of research reactors around the world persists as a primary topic of concern among facility owners and operators, the IAEA responded in 2013 with a broad range of activities to address the planning, execution and improvement of many experimental techniques. The revision of two critical documents for planning and diversifying a facility's portfolio of applications, TECDOC 1234 “The Applications of Research Reactors” and TECDOC 1212 “Strategic Planning for Research Reactors”, is in progress in order to keep this information relevant, corresponding to the dynamism of experimental techniques and research capabilities. Related to the latter TECDOC, the IAEA convened a meeting in 2013 for the expert review of a number of strategic plans submitted by research reactor operators in developing countries. A number of activities focusing on specific applications are either continuing or beginning as well. In neutron activation analysis, a joint round of inter-comparison proficiency testing sponsored by the IAEA Technical Cooperation Department will be completed, and facility progress in measurement accuracy is described. Also, a training workshop in neutron imaging and Coordinated Research Projects in reactor benchmarks, automation of neutron activation analysis and neutron beam techniques for material testing intend to advance these activities as more beneficial services to researchers and other users.

  20. Yale High Energy Physics Research: Precision Studies of Reactor Antineutrinos

    Energy Technology Data Exchange (ETDEWEB)

    Heeger, Karsten M. [Yale Univ., New Haven, CT (United States)

    2014-09-13

    This report presents experimental research at the intensity frontier of particle physics with particular focus on the study of reactor antineutrinos and the precision measurement of neutrino oscillations. The experimental neutrino physics group of Professor Heeger and Senior Scientist Band at Yale University has had leading responsibilities in the construction and operation of the Daya Bay Reactor Antineutrino Experiment and made critical contributions to the discovery of non-zero$\\theta_{13}$. Heeger and Band led the Daya Bay detector management team and are now overseeing the operations of the antineutrino detectors. Postdoctoral researchers and students in this group have made leading contributions to the Daya Bay analysis including the prediction of the reactor antineutrino flux and spectrum, the analysis of the oscillation signal, and the precision determination of the target mass yielding unprecedented precision in the relative detector uncertainty. Heeger's group is now leading an R\\&D effort towards a short-baseline oscillation experiment, called PROSPECT, at a US research reactor and the development of antineutrino detectors with advanced background discrimination.

  1. An evaluation of alternative reactor vessel cutting technologies for the experimental boiling water reactor at Argonne National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Boing, L.E.; Henley, D.R. (Argonne National Lab., IL (USA)); Manion, W.J.; Gordon, J.W. (Nuclear Energy Services, Inc., Danbury, CT (USA))

    1989-12-01

    Metal cutting techniques that can be used to segment the reactor pressure vessel of the Experimental Boiling Water Reactor (EBWR) at Argonne National Laboratory (ANL) have been evaluated by Nuclear Energy Services. Twelve cutting technologies are described in terms of their ability to perform the required task, their performance characteristics, environmental and radiological impacts, and cost and schedule considerations. Specific recommendations regarding which technology should ultimately be used by ANL are included. The selection of a cutting method was the responsibility of the decommissioning staff at ANL, who included a relative weighting of the parameters described in this document in their evaluation process. 73 refs., 26 figs., 69 tabs.

  2. A reload and startup plan for conversion of the NIST research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Diamond, D. J. [Brookhaven National Lab. (BNL), Upton, NY (United States)

    2016-03-31

    The National Institute of Standards and Technology operates a 20 MW research reactor for neutron-based research. The heavy-water moderated and cooled reactor is fueled with high-enriched uranium (HEU) but a program to convert the reactor to low-enriched uranium (LEU) fuel is underway. Among other requirements, a reload and startup test plan must be submitted to the U.S. Nuclear Regulatory Commission (NRC) for their approval. The NRC provides guidance for what should be in the plan to ensure that the licensee has sufficient information to operate the reactor safely. Hence, a plan has been generated consisting of two parts. The reload portion of the plan specifies the fuel management whereby initially only two LEU fuel elements are in the core for eight fuel cycles. This is repeated until a point when the optimum approach is to place four fresh LEU elements into the reactor each cycle. This final transition is repeated and after eight cycles the reactor is completely fueled with LEU. By only adding two LEU fuel elements initially, the plan allows for the consumption of HEU fuel elements that are expected to be in storage at the time of conversion and provides additional qualification of production LEU fuel under actual operating conditions. Because the reload is to take place over many fuel cycles, startup tests will be done at different stages of the conversion. The tests, to be compared with calculations to show that the reactor will operate as planned, are the measurement of critical shim arm position and shim arm and regulating rod reactivity worths. An acceptance criterion for each test is specified based on technical specifications that relate to safe operation. Additional tests are being considered that have less safety significance but may be of interest to bolster the validation of analysis tools.

  3. Needs and Requirements for Future Research Reactors (ORNL Perspectives)

    Energy Technology Data Exchange (ETDEWEB)

    Ilas, Germina [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bryan, Chris [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gehin, Jess C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-02-10

    The High Flux Isotope Reactor (HFIR) is a vital national and international resource for neutron science research, production of radioisotopes, and materials irradiation. While HFIR is expected to continue operation for the foreseeable future, interest is growing in understanding future research reactors features, needs, and requirements. To clarify, discuss, and compile these needs from the perspective of Oak Ridge National Laboratory (ORNL) research and development (R&D) missions, a workshop, titled “Needs and Requirements for Future Research Reactors”, was held at ORNL on May 12, 2015. The workshop engaged ORNL staff that is directly involved in research using HFIR to collect valuable input on the reactor’s current and future missions. The workshop provided an interactive forum for a fruitful exchange of opinions, and included a mix of short presentations and open discussions. ORNL staff members made 15 technical presentations based on their experience and areas of expertise, and discussed those capabilities of the HFIR and future research reactors that are essential for their current and future R&D needs. The workshop was attended by approximately 60 participants from three ORNL directorates. The agenda is included in Appendix A. This document summarizes the feedback provided by workshop contributors and participants. It also includes information and insights addressing key points that originated from the dialogue started at the workshop. A general overview is provided on the design features and capabilities of high performance research reactors currently in use or under construction worldwide. Recent and ongoing design efforts in the US and internationally are briefly summarized, followed by conclusions and recommendations.

  4. Development of materials and manufacturing technologies for Indian fast reactor programme

    Energy Technology Data Exchange (ETDEWEB)

    Raj, Baldev; Jayakumar, T.; Bhaduri, A.K.; Mandal, Sumantra [Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

    2010-07-01

    Fast Breeder Reactors (FBRs) are vital towards meeting security and sustainability of energy for the growing economy of India. The development of FBRs necessitates extensive research and development in domains of materials and manufacturing technologies in association with a wide spectrum of disciplines and their inter-twining to meet the challenging technology. The paper highlight the work and the approaches adopted for the successful deployment of materials, manufacturing and inspection technologies for the in-core and structural components of current and future Indian Fast Breeder Reactor Programme. Indigenous development of in-core materials viz. Titanium modified austenitic stainless steel (Alloy D9) and its variants, ferritic/martensitic oxide-dispersion strengthened (ODS) steels as well as structural materials viz. 316L(N) stainless steel and modified 9Cr-1Mo have been achieved through synergistic interactions between Indira Gandhi Centre for Atomic Research (IGCAR), education and research institutes and industries. Robust manufacturing technology has been established for forming and joining of various components of 500 MWe Prototype Fast Breeder Reactor (PFBR) through 'science-based technology' approach. To achieve the strict quality standards of formed parts in terms of geometrical tolerances, residual stresses and microstructural defects, FEM-based modelling and experimental validation was carried out for estimation of spring-back during forming of multiple curvature thick plantes. Optimization of grain boundary character distribution in Alloy D9 was carried out by adopting the grain boundary engineering approach to reduce radiation induced segregation. Extensive welding is involved in the fabrication of reactor vessels, piping, steam generators, fuel sub-assemblies etc. Activated Tungsten Inert Gas Welding process along with activated flux developed at IGCAR has been successfully used in fabrication of dummy fuel subassemblies (DFSA) required

  5. A Study on Comparison of HANARO and KIJANG Research Reactor in Nuclear Safeguards

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Juang; Lee, Sung Ho; Kim, Hyun-Jo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    As one of major national projects for nuclear science and engineering in Korea, the KIJANG Research Reactor(KJRR) project was commenced in order to develop the core research reactor(RR) technologies for strengthening the competitiveness of the RR export and also to stabilize the supply of key radioisotopes for medical and industrial applications. This paper is about applying IAEA safeguards at new nuclear facility (KJRR). The beginning of this project is comparing of HANARO and KIJANG research reactor in nuclear safeguards for nuclear material accountancy method. As mentioned before, research reactor is basically item counting facility. In Fig 1, first two processes are belonging to item counting. But last two processes are for bulk handling. So KIJANG RR would be treated item counting facility as well as bulk handling facility by fission moly production facility. For this reason, nuclear material accountancy method for KJRR is not easy compared to existing one. This paper accounted for solution of KJRR nuclear material accountancy briefly. Future study on the suitable nuclear material accountancy method for mixed facility between item counting facility and bulk handling facility will be conducted more specifically.

  6. Reactor-produced radionuclides at the University of Missouri Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ketring, A.R.; Evans-Blumer, M.S.; Ehrhardt, G.J. [University of Missouri Research Reactor, Colombia (United States). Departments of Radiology, Chemistry and Nuclear Engineering

    1997-10-01

    Nuclear medicine has primarily been a diagnostic science for many years, but today is facing considerable challenges from other modalities in this area. However, these competing techniques (magnetic resonance imaging, ultrasound, and computer-assisted tomography) in general are not therapeutic. Although early nuclear medicine therapy was of limited efficacy, in recent years a revolution in radiotherapy has been developing base don more sophisticated targeting methods, including radioactive intra-arterial microspheres, chemically-guided bone agents, labelled monoclonal antibodies, and isotopically-tagged polypeptide receptor-binding agents. Although primarily used for malignancies, therapeutic nuclear medicine is also applicable to the treatment of rheumatoid arthritis and possibly coronary artery re closure following angioplasty. The isotopes of choice for these applications are reactor-produced beta emitters such as Sm-153, Re-186, Re-188, Ho-166, Lu-177, and Rh-105. Although alpha emitters possess greater cell toxicity due to their high LET, the greater range of beta emitters and the typically inhomogeneous deposition of radiotherapy agents in lesions leads to greater beta `crossfire` and better overall results. The University of Missouri Research Reactor (MURR) has been in the forefront of research into means of preparing, handling and supplying these high-specific-activity isotopes in quantities appropriate not only for research, but also for patient trials in the US and around the world. Researchers at MURR in collaboration with others at the University of Missouri (MU) developed Sm-153 Quadramet{sup TM}, a drug recently approved in the US for palliation of bone tumor pain. In conjunction with researchers at the University of Missouri-Rolla, MURR also developed Y-90 TheraSphere{sup TM}, an agent for the treatment of liver cancer now approved in Canada. Considerable effort has been expended to develop techniques for irradiation, handling, and shipping isotopes

  7. 75 FR 27368 - Aerotest Operations, Inc., Aerotest Radiography and Research Reactor; Notice of Consideration of...

    Science.gov (United States)

    2010-05-14

    ... COMMISSION Aerotest Operations, Inc., Aerotest Radiography and Research Reactor; Notice of Consideration of... INFORMATION CONTACT: Cindy Montgomery, Project Manager, Research and Test Reactors Licensing Branch, Division... Operating License No. R-98 for the Aerotest Radiography and Research Reactor (ARRR), currently held by...

  8. Major update of Safety Analysis Report for Thai Research Reactor-1/Modification 1

    Energy Technology Data Exchange (ETDEWEB)

    Tippayakul, Chanatip [Thailand Institute of Nuclear Technology, Bangkok (Thailand)

    2013-07-01

    Thai Research Reactor-1/Modification 1 (TRR-1/M1) was converted from a Material Testing Reactor in 1975 and it had been operated by Office of Atom for Peace (OAP) since 1977 until 2007. During the period, Office of Atom for Peace had two duties for the reactor, that is, to operate and to regulate the reactor. However, in 2007, there was governmental office reformation which resulted in the separation of the reactor operating organization from the regulatory body in order to comply with international standard. The new organization is called Thailand Institute of Nuclear Technology (TINT) which has the mission to promote peaceful utilization of nuclear technology while OAP remains essentially the regulatory body. After the separation, a new ministerial regulation was enforced reflecting a new licensing scheme in which TINT has to apply for a license to operate the reactor. The safety analysis report (SAR) shall be submitted as part of the license application. The ministerial regulation stipulates the outlines of the SAR almost equivalent to IAEA standard 35-G1. Comparing to the IAEA 35-G1 standard, there were several incomplete and missing chapters in the original SAR of TRR1/M1. The major update of the SAR was therefore conducted and took approximately one year. The update work included detail safety evaluation of core configuration which used two fuel element types, the classification of systems, structures and components (SSC), the compilation of detail descriptions of all SSCs and the review and evaluation of radiation protection program, emergency plan and emergency procedure. Additionally, the code of conduct and operating limits and conditions were revised and finalized in this work. A lot of new information was added to the SAR as well, for example, the description of commissioning program, information on environmental impact assessment, decommissioning program, quality assurance program and etc. Due to the complexity of this work, extensive knowledge was

  9. Preliminary conceptual design for electrical and I and C system of a new research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Hoan Sung; Kim, Y. K.; Kim, M. J.; Kim, H. K.; Ryu, J. S

    2004-01-01

    The core type and the process system design will be varied according to the reactor's application and capacity. A New research reactor is being designed by KAERI since 2002 and the process systems are not fixed yet. But control and instrument systems are similar to each other even though the application and the size are not same. So the C and I system that encompasses reactor protection system, reactor control system, and computer system was designed conceptually according to the requirements based on new digital technology and HANARO's proven design. The plant electrical system consists of off-site system that delivers bulk electrical power to the reactor site and on-site system that distributes and controls electrical power at the facility. The electrical system includes building service system that consist of lighting, communication, fire detection, grounding, cathodic protection, etc. also. This report describes the design requirements of on-site and off-site electric power system that set up from the codes and standards and the conceptual design based on the design requirements.

  10. Safety culture and quality management of Kartini research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Syarip [Yogyakarta Nuclear Research Centre, Yogyakarta (Indonesia); Hauptmanns, Ulrich [Department of Plant Design and Safety, Otto-Von-Guericke-University, Magdeburg (Germany)

    1999-10-01

    The evaluation for assessing the safety culture and quality of safety management of Kartini research reactor is presented. The method is based on the concept of management control of safety (audit) as well as by using the developed method i.e. the questionnaires concerning areas of relevance which have to be answered with value statements. There are seven statements or qualifiers in answering the questions. Since such statements are vague, they are represented by fuzzy numbers. The weaknesses can be identified from the different areas contemplated. The evaluation result show that the quality of safety management of Kartini research reactor is globally rated as 'Average'. The operator behavior in the implementation of 'safety culture' concept is found as a weakness, therefore this area should be improved. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-04-15

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

  12. Technology, Safety and Costs of Decommissioning Nuclear Reactors At Multiple-Reactor Stations

    Energy Technology Data Exchange (ETDEWEB)

    Wittenbrock, N. G.

    1982-01-01

    Safety and cost information is developed for the conceptual decommissioning of large (1175-MWe) pressurized water reactors (PWRs) and large (1155-MWe) boiling water reactors {BWRs) at multiple-reactor stations. Three decommissioning alternatives are studied: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and ENTOMB (entombment). Safety and costs of decommissioning are estimated by determining the impact of probable features of multiple-reactor-station operation that are considered to be unavailable at a single-reactor station, and applying these estimated impacts to the decommissioning costs and radiation doses estimated in previous PWR and BWR decommissioning studies. The multiple-reactor-station features analyzed are: the use of interim onsite nuclear waste storage with later removal to an offsite nuclear waste disposal facility, the use of permanent onsite nuclear waste disposal, the dedication of the site to nuclear power generation, and the provision of centralized services. Five scenarios for decommissioning reactors at a multiple-reactor station are investigated. The number of reactors on a site is assumed to be either four or ten; nuclear waste disposal is varied between immediate offsite disposal, interim onsite storage, and immediate onsite disposal. It is assumed that the decommissioned reactors are not replaced in one scenario but are replaced in the other scenarios. Centralized service facilities are provided in two scenarios but are not provided in the other three. Decommissioning of a PWR or a BWR at a multiple-reactor station probably will be less costly and result in lower radiation doses than decommissioning an identical reactor at a single-reactor station. Regardless of whether the light water reactor being decommissioned is at a single- or multiple-reactor station: • the estimated occupational radiation dose for decommissioning an LWR is lowest for SAFSTOR and highest for DECON • the estimated

  13. Experimental facility for development of high-temperature reactor technology: instrumentation needs and challenges

    Directory of Open Access Journals (Sweden)

    Sabharwall Piyush

    2015-01-01

    Full Text Available A high-temperature, multi-fluid, multi-loop test facility is under development at the Idaho National Laboratory for support of thermal hydraulic materials, and system integration research for high-temperature reactors. The experimental facility includes a high-temperature helium loop, a liquid salt loop, and a hot water/steam loop. The three loops will be thermally coupled through an intermediate heat exchanger (IHX and a secondary heat exchanger (SHX. Research topics to be addressed include the characterization and performance evaluation of candidate compact heat exchangers such as printed circuit heat exchangers (PCHEs at prototypical operating conditions. Each loop will also include an interchangeable high-temperature test section that can be customized to address specific research issues associated with each working fluid. This paper also discusses needs and challenges associated with advanced instrumentation for the multi-loop facility, which could be further applied to advanced high-temperature reactors. Based on its relevance to advanced reactor systems, the new facility has been named the Advanced Reactor Technology Integral System Test (ARTIST facility. A preliminary design configuration of the ARTIST facility will be presented with the required design and operating characteristics of the various components. The initial configuration will include a high-temperature (750 °C, high-pressure (7 MPa helium loop thermally integrated with a molten fluoride salt (KF-ZrF4 flow loop operating at low pressure (0.2 MPa, at a temperature of ∼450 °C. The salt loop will be thermally integrated with the steam/water loop operating at PWR conditions. Experiment design challenges include identifying suitable materials and components that will withstand the required loop operating conditions. The instrumentation needs to be highly accurate (negligible drift in measuring operational data for extended periods of times, as data collected will be

  14. On use of ZPR research reactors and associated instrumentation and measurement methods for reactor physics studies

    Energy Technology Data Exchange (ETDEWEB)

    Chauvin, J.P. [CEA,DEN, DER, SPEX, Experimental Physics Service, Cadarache, F-13108 St-Paul-Lez-Durance (France); Blaise, P. [CEA, DEN, DER, SPEX Experimental Programs Laboratory, Cadarache, F-13108 St-Paul-Lez-Durance (France); Lyoussi, A. [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-Lez-Durance (France)

    2015-07-01

    The French atomic and alternative energies -CEA- is strongly involved in research and development programs concerning the use of nuclear energy as a clean and reliable source of energy and consequently is working on the present and future generation of reactors on various topics such as ageing plant management, optimization of the plutonium stockpile, waste management and innovative systems exploration. Core physics studies are an essential part of this comprehensive R and D effort. In particular, the Zero Power Reactor (ZPR) of CEA: EOLE, MINERVE and MASURCA play an important role in the validation of neutron (as well photon) physics calculation tools (codes and nuclear data). The experimental programs defined in the CEA's ZPR facilities aim at improving the calculation routes by reducing the uncertainties of the experimental databases. They also provide accurate data on innovative systems in terms of new materials (moderating and decoupling materials) and new concepts (ADS, ABWR, new MTR (e.g. JHR), GENIV) involving new fuels, absorbers and coolant materials. Conducting such interesting experimental R and D programs is based on determining and measuring main parameters of phenomena of interest to qualify calculation tools and nuclear data 'libraries'. Determining these parameters relies on the use of numerous and different experimental techniques using specific and appropriate instrumentation and detection tools. Main ZPR experimental programs at CEA, their objectives and challenges will be presented and discussed. Future development and perspectives regarding ZPR reactors and associated programs will be also presented. (authors)

  15. Technological status of reactor coolant pumps in generation III+ pressurized nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Brecht, Bernhard; Bross, Stephan [KSB Aktiengesellschaft, Frankenthal (Germany)

    2016-05-15

    KSB has been developing and producing pumps for thermal power plants for nearly 90 years. Consequently, KSB also started to develop and manufacture pumps for all kinds of nuclear power plants from the very beginning of the civil use of nuclear energy. This is especially true for reactor coolant pumps for pressurized water reactors. For the generation of advanced evolutionary reactors (Generation III+ reactors), KSB developed an advanced shaft seal system which is also able to fulfill the requirements of station blackout conditions. The tests in the KSB test rigs, which were successfully completed in December 2015, proved the full functionality of the new design. For generation III+ passive plant reactors KSB developed a new reactor coolant pump type called RUV, which is based on the experience of classic reactor coolant pumps and reactor internal pumps. It is a very compact, hermetically sealed vertical pump-motor unit with a wet winding motor. A full scale prototype successfully passed the 1st stage qualification test program in October 2015.

  16. Dismantling design for a reference research reactor of the WWR type

    Energy Technology Data Exchange (ETDEWEB)

    Lobach, Yu.N., E-mail: lobach@kinr.kiev.ua [Institute for Nuclear Research, Pr. Nauki, 47, Kiev 03680 (Ukraine); Cross, M.T., E-mail: Martin.Cross@nuvia.co.uk [Nuvia Ltd., Robinson House, Crow Park Way, Westlakes Science and Technology Park, Moor Row, Cumbria CA24 3HY (United Kingdom)

    2014-01-15

    Highlights: • Design features of WWRs relevant to decommissioning have been analysed. • The technical basis for the preparation and implementation of dismantling has been established for a reference WWR. • The applicability of existing proven dismantling technologies has been established. -- Abstract: A decommissioning study has been carried out for a reference research reactor of the WWR type. Many such reactors were constructed more than 50 years ago and most of them are still in operation. Decommissioning has now become an important consideration. This paper summarizes the main decommissioning steps and, on the basis of the reactor design features, technical aspects of the dismantling and removal of the contaminated/activated components have been analysed. The advisability of the removal of large components, such as the reactor vessel and the heat-exchangers, as one piece items has also been demonstrated. Additionally, a work schedule and an estimation of the collective dose for the preparation and implementation of dismantling have been established. The applicability of existing proven dismantling technologies has been identified together with some additional features for the dismantling.

  17. Preparation Before Signature of Upgrade of Algeria Heavy Water Research Reactor Contract

    Institute of Scientific and Technical Information of China (English)

    LI; Song; ZAN; Huai-qi; XU; Qi-guo; JIA; Yu-wen

    2012-01-01

    <正>Algeria heavy water research reactor (Birine) is a multiple-purpose research reactor, which was constructed with the help of China more than 20 years ago. By request of Algeria, China will upgrade the research reactor; so as to improve the status of current reactor such as equipment ageing, shortage of spare parts, several systems do not meet requirements of current standards and criteria etc.

  18. Characterization of radioactive aerosols in Tehran research reactor containment

    Directory of Open Access Journals (Sweden)

    Moradi Gholamreza

    2015-01-01

    Full Text Available The objectives of this research were to determine the levels of radioactivity in the Tehran research reactor containment and to investigate the mass-size distribution, composition, and concentration of radionuclides during operation of the reactor. A cascade impactor sampler was used to determine the size-activity distributions of radioactive aerosols in each of the sampling stations. Levels of a and b activities were determined based on a counting method using a liquid scintillation counter and smear tests. The total average mass fractions of fine particles (particle diameter dp < 1 mm in all of the sampling stations were approximately 26.75 %, with the mean and standard deviation of 52.15 ± 19.75 mg/m3. The total average mass fractions of coarse particles were approximately 73.2%, with the mean and standard deviation of 71.34 ± 24.57 mg/m3. In addition to natural radionuclides, artificial radionuclides, such as 24Na, 91Sr, 131I, 133I, 103Ru, 82Br, and 140La, may be released into the reactor containment structure. Maximum activity was associated with accumulation-mode particles with diameters less than 400 nm. The results obtained from liquid scintillation counting suggested that the mean specific activity of alpha particles in fine and coarse-modes were 89.7 % and 10.26 %, respectively. The mean specific activity of beta particles in fine and coarse-modes were 81.15 % and 18.51 %, respectively. A large fraction of the radionuclides' mass concentration in the Tehran research reactor containment was associated with coarse-mode particles, in addition, a large fraction of the activity in the aerosol particles was associated with accumulation-mode particles.

  19. Replacement Nuclear Research Reactor: Draft Environmental Impact Statement. Vol. 2. Appendices

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-07-01

    The appendices contains additional relevant information on: Environment Australia EIS Guidelines, composition of the Study Team, Consultation Activities and Resuits, Relevant Legislation and Regulatory Requirements, Exampies of Multi-Purpose Research Reactors, Impacts of Radioactive Emissions and Wastes Generated at Lucas Heights Science and Technology Centre, Technical Analysis of the Reference Accident, Flora and Fauna Species Lists, Summary of Environmental Commitments and an Outline of the Construction Environmental Management Plan Construction Environmental Management Plan figs., ills., refs. Prepared for Australian Nuclear Science and Technology Organisation (ANSTO)

  20. Present status and future perspectives of research and test reactor in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Kaneko, Yoshihiko [Atomic Energy Research Laboratory, Musashi Institute of Technology, Kawasaki, Kanagawa (Japan); Kaieda, Keisuke [Department of Research Reactor, Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan)

    2000-10-01

    Since 1957, Japan Atomic Energy Research Institute (JAERI) has constructed several research and test reactors to fulfill a major role in the study of nuclear energy and fundamental research. At present four reactors, the Japan Research Reactor No. 3 and No. 4 (JRR-3M and JRR-4 respectively), the Japan Materials Testing Reactor (JMTR) and the Nuclear Safety Research Reactor (NSRR) are in operation, and a new High Temperature Engineering Test Reactor (HTTR) has recently reached first criticality and now in the power up test. In 1966, the Kyoto University built the Kyoto University Reactor (KUR) and started its operation for joint use program of the Japanese universities. This paper introduces these reactors and describes their present operational status and also efforts for aging management. The recent tendency of utilization and future perspectives is also reported. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

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

  2. Feasibility of Thermoelectric Waste Heat Recovery from Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Byunghee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    A thermoelectric generator has the most competitive method to regenerate the waste heat from research reactors, because it has no limitation on operating temperature. In addition, since the TEG is a solid energy conversion device converting heat to electricity directly without moving parts, the regenerating power system becomes simple and highly reliable. In this regard, a waste heat recovery using thermoelectric generator (TEG) from 15-MW pool type research reactor is suggested and the feasibility is demonstrated. The producible power from waste heat is estimated with respect to the reactor parameters, and an application of the regenerated power is suggested by performing a safety analysis with the power. The producible power from TEG is estimated with respect to the LMTD of the HX and the required heat exchange area is also calculated. By increasing LMTD from 2 K to 20K, the efficiency and the power increases greatly. Also an application of the power regeneration system is suggested by performing a safety analysis with the system, and comparing the results with reference case without the power regeneration.

  3. Status of research and development on reduced-moderation water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Iwamura, Takamichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2002-12-01

    To improve uranium utilization, a design study of the Reduced-Moderation Water Reactor (RMWR) has been carried out intensively since 1998 at the Japan Atomic Energy Research Institute (JAERI). In this reactor, the nuclear fission reaction is designed to be realized mainly by high energy neutrons. To achieve this, the volume of water used to cool the fuel rods is decreased by reducing the gap width between the fuel rods. Conversion ratio greater than 1.0 is expected whether the core i-s cooled by boiling water or pressurized water and whether the core size is small or large. Status of the RMWR design is reviewed and planning of R and D for future deployment of this reactor after 20-20 is presented. To improve economics of this reactor, development of fuel cans for high burnup and low-cost reprocessing technology of mixed oxide spect fuels are highly needed. R and D has been conducted under the cooperation with utilities, industry, research organization and academia. (T. Tanaka)

  4. Developing strategic plans for effective utilization of research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ridikas, Danas [International Atomic Energy Agency, Vienna (Austria). Dept. of Nuclear Sciences and Applications

    2015-12-15

    Strategic plans are indispensable documents for research reactors (RRs) to ensure their efficient, optimized and well managed utilization. A strategic plan provides a framework for increasing utilization, while helping to create a positive safety culture, a motivated staff, a clear understanding of real costs and a balanced budget. A strategic plan should be seen as an essential tool for a responsible manager of any RR, from the smallest critical facility to the largest reactor. Results and lessons learned are shown from the IAEA efforts to help the RR facilities developing strategic plans, provide review and advise services, organize national and regional stakeholder/user workshops, prepare further guidance and recommendations, document and publish guidance documents and other supporting materials.

  5. Status of reactor shielding research in the United States

    Energy Technology Data Exchange (ETDEWEB)

    Bartine, D.E.

    1983-01-01

    Shielding research in the United States continues to place emphasis on: (1) the development and refinement of shielding design calculational methods and nuclear data; and (2) the performance of confirmation experiments, both to evaluate specific design concepts and to verify specific calculational techniques and input data. The successful prediction of the radiation levels observed within the now-operating Fast Flux Test Facility (FFTF) has demonstrated the validity of this two-pronged approach, which has since been applied to US fast breeder reactor programs and is now being used to determine radiation levels and possible further shielding needs at operating light water reactors, especially under accident conditions. A similar approach is being applied to the back end of the fission fuel cycle to verify that radiation doses at fuel element storage and transportation facilities and within fuel reprocessing plants are kept at acceptable levels without undue economic penalties.

  6. Reactivity feedback coefficients Pakistan research reactor-1 using PRIDE code

    Energy Technology Data Exchange (ETDEWEB)

    Mansoor, Ali; Ahmed, Siraj-ul-Islam; Khan, Rustam [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan). Dept. of Nuclear Engineering; Inam-ul-Haq [Comsats Institute of Information Technology, Islamabad (Pakistan). Dept. of Physics

    2017-05-15

    Results of the analyses performed for fuel, moderator and void's temperature feedback reactivity coefficients for the first high power core configuration of Pakistan Research Reactor - 1 (PARR-1) are summarized. For this purpose, a validated three dimensional model of PARR-1 core was developed and confirmed against the reference results for reactivity calculations. The ''Program for Reactor In-Core Analysis using Diffusion Equation'' (PRIDE) code was used for development of global (3-dimensional) model in conjunction with WIMSD4 for lattice cell modeling. Values for isothermal fuel, moderator and void's temperature feedback reactivity coefficients have been calculated. Additionally, flux profiles for the five energy groups were also generated.

  7. Design requirement for electrical system of an advanced research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Hoan Sung; Kim, H. K.; Kim, Y. K.; Wu, J. S.; Ryu, J. S

    2004-12-01

    An advanced research reactor is being designed since 2002 and the conceptual design has been completed this year for the several types of core. Also the fuel was designed for the potential cores. But the process system, the I and C system, and the electrical system design are under pre-conceptual stage. The conceptual design for those systems will be developed in the next year. Design requirements for the electrical system set up to develop conceptual design. The same goals as reactor design - enhance safety, reliability, economy, were applied for the development of the requirements. Also the experience of HANARO design and operation was based on. The design requirements for the power distribution, standby power supply, and raceway system will be used for the conceptual design of electrical system.

  8. Return of spent fuel from the Portuguese research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ramalho, A.J.G.; Marques, J.G.; Cardeira, F.M. [Instituto Tecnologico e Nuclear, PO-2686-953 Sacavem (Portugal)

    2000-07-01

    Thirty-nine spent MTR fuel assemblies from the Portuguese Research Reactor were recently returned to the US. Prior to the shipment all assemblies were inspected for corrosion and sipped for determination of fission product leakage. Limitations on the floor loading of the reactor building and on the capacity of the crane prevented the placement and loading of the Transnucleaire IU04 transport cask inside the containment building. The transport cask was thus placed outside, under permanent surveillance, in a support structure built around it. A small transfer cask was used to carry individually the assemblies from the storage racks to the transport cask. A forklift was used as a shuttle between the pool and the IU04. A detailed description of the procedures is given. (author)

  9. Control Rod Reactivity Curves for the Annular Core Research Reactor

    Science.gov (United States)

    Depriest, K. Russell; Kajder, Karen C.; Frye, Jason N.; Denman, Matthew R.

    2009-08-01

    Experiments were conducted at the Annular Core Research Reactor (ACRR) to increase the fidelity of the control rod integral reactivity worth curve. This experiment series was designed to refine the integral reactivity curve used for pulse yield prediction and eliminate the need for operator compensation in the pulse setup. The experiment series consisted of delayed critical and positive period measurements with various ACRR cavity configurations. An improved integral reactivity worth curve for the ACRR control rods has been constructed using the positive period measurements, the delayed critical measurements, and radiation transport modeling of the reactor. A series of prompt period measurements is used to validate that the new control rod curve more accurately predicts the energy yield of the pulse operations. The new reactivity worth curve is compared with the current curve that was developed using traditional approaches.

  10. Advanced Reactor Safety Research Division. Quarterly progress report, January 1-March 31, 1980

    Energy Technology Data Exchange (ETDEWEB)

    Agrawal, A.K.; Cerbone, R.J.; Sastre, C.

    1980-06-01

    The Advanced Reactor Safety Research Programs quarterly progress report describes current activities and technical progress in the programs at Brookhaven National Laboratory sponsored by the USNRC Division of Reactor Safety Research. The projects reported each quarter are the following: HTGR Safety Evaluation, SSC Code Development, LMFBR Safety Experiments, and Fast Reactor Safety Code Validation.

  11. Advanced Reactor Safety Research Division. Quarterly progress report, April 1-June 30, 1980

    Energy Technology Data Exchange (ETDEWEB)

    Romano, A.J.

    1980-01-01

    The Advanced Reactor Safety Research Programs Quarterly Progress Report describes current activities and technical progress in the programs at Brookhaven National Laboratory sponsored by the USNRC Division of Reactor Safety Research. The projects reported each quarter are the following: HTGR safety evaluation, SSC Code Development, LMFBR Safety Experiments, and Fast Reactor Safety Code Validation.

  12. User research & technology, pt.2

    CERN Document Server

    Greifeneder, Elke

    2011-01-01

    This e-book is Part 2 on the theme "User Research and Technology". The research covers the testing of online digital library resources using various methods. Library and information science as a field is changing and the requirements for top quality research are growing more stringent. This is typical of the experience of other professional fields as they have moved from practitioners advising practitioners to researchers building on past results. This e-book contains 12 papers on this theme.

  13. Operation experience of the Indonesian multipurpose research reactor RSG-GAS

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-08-01

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

  14. Decommissioning of the ASTRA research reactor: Planning, executing and summarizing the project

    Directory of Open Access Journals (Sweden)

    Meyer Franz

    2010-01-01

    Full Text Available The decommissioning of the ASTRA research reactor at the Austrian Research Centres Seibersdorf was described within three technical papers already released in Nuclear Technology & Radiation Protection throughout the years 2003, 2006, and 2008. Following a suggestion from IAEA the project was investigated well after the files were closed regarding rather administrative than technical matters starting with the project mission, explaining the project structure and identifying the key factors and the key performance indicators. The continuous documentary and reporting system as implemented to fulfil the informational needs of stake-holders, management, and project staff alike is described. Finally the project is summarized in relationship to the performance indicators.

  15. Next Generation Nuclear Plant Reactor Pressure Vessel Materials Research and Development Plan (PLN-2803)

    Energy Technology Data Exchange (ETDEWEB)

    J. K. Wright; R. N. Wright

    2010-07-01

    The U.S. Department of Energy (DOE) has selected the High-Temperature Gas-cooled Reactor (HTGR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production, with an outlet gas temperature in the range of 750°C, and a design service life of 60 years. The reactor design will be a graphite-moderated, helium-cooled, prismatic, or pebble bed reactor and use low-enriched uranium, Tri-Isotopic (TRISO)-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. This technology development plan details the additional research and development (R&D) required to design and license the NGNP RPV, assuming that A 508/A 533 is the material of construction. The majority of additional information that is required is related to long-term aging behavior at NGNP vessel temperatures, which are somewhat above those commonly encountered in the existing database from LWR experience. Additional data are also required for the anticipated NGNP environment. An assessment of required R&D for a Grade 91 vessel has been retained from the first revision of the R&D plan in Appendix B in somewhat less detail. Considerably more development is required for this steel compared to A 508/A 533 including additional irradiation testing for expected NGNP operating temperatures, high-temperature mechanical properties, and extensive studies of long-term microstructural stability.

  16. Oak Ridge National Laboratory Support of Non-light Water Reactor Technologies: Capabilities Assessment for NRC Near-term Implementation Action Plans for Non-light Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Belles, Randy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jain, Prashant K. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-04-01

    The Oak Ridge National Laboratory (ORNL) has a rich history of support for light water reactor (LWR) and non-LWR technologies. The ORNL history involves operation of 13 reactors at ORNL including the graphite reactor dating back to World War II, two aqueous homogeneous reactors, two molten salt reactors (MSRs), a fast-burst health physics reactor, and seven LWRs. Operation of the High Flux Isotope Reactor (HFIR) has been ongoing since 1965. Expertise exists amongst the ORNL staff to provide non-LWR training; support evaluation of non-LWR licensing and safety issues; perform modeling and simulation using advanced computational tools; run laboratory experiments using equipment such as the liquid salt component test facility; and perform in-depth fuel performance and thermal-hydraulic technology reviews using a vast suite of computer codes and tools. Summaries of this expertise are included in this paper.

  17. Radioisotope Production Plan and Strategy of Kijang Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kye Hong; Lee, Jun Sig [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    This reactor will be located at Kijang, Busan, Korea and be dedicated to produce mainly medical radioisotopes. Tc-99m is very important isotope for diagnosis and more than 80% of radiation diagnostic procedures in nuclear medicine depend on this isotope. There were, however, several times of insecure production of Mo-99 due to the shutdown of major production reactors worldwide. OECD/NEA is leading member countries to resolve the shortage of this isotope and trying to secure the international market of Mo-99. The radioisotope plan and strategy of Kijang Research Reactor (KJRR) should be carefully established to fit not only the domestic but also international demand on Mo-99. The implementation strategy of 6 principles of HLG-MR should be established that is appropriate to national environments. Ministry of Science, ICT and Future Planning and Ministry of Health and welfare should cooperate well to organize the national radioisotope supply structure, to set up the reasonable and competitive pricing of radioisotopes, and to cope with the international supply strategy.

  18. Progress with OPAL, the new Australian research reactor

    Indian Academy of Sciences (India)

    R A Robinson

    2008-11-01

    Australian science is entering a new `golden age', with the start-up of bright new neutron and photon sources in Sydney and Melbourne, in 2006 and 2007 respectively. The OPAL reactor and the Australian Synchrotron can be considered as the greatest single investment in scientific infrastructure in Australia's history. They will essentially be `sister' facilities, with a common open user ethos, and a vision to play a major role in international science. Fuel was loaded into the reactor in August 2006, and full power was (20 MW) achieved in November 2006. The first call for proposals was made in 2007, and commissioning experiments have taken place well before then. The first three instruments in operation are high-resolution powder diffractometer (for materials discovery), high-intensity powder diffractometer (for kinetics experiments and small samples) and a strain scanner (for mechanical engineering and industrial applications). These are closely followed by four more instruments with broad application in nanoscience, condensed matter physics and other scientific disciplines. Instrument performance will be competitive with the best research-reactor facilities anywhere. To date there is committed funding for nine instruments, with a capacity to install a total of ∼ 18 beamlines. An update will be given on the status of OPAL, its thermal and cold neutron sources, its instruments and the first results.

  19. Space and nuclear research and technology

    Science.gov (United States)

    1975-01-01

    A fact sheet is presented on the space and nuclear research and technology program consisting of a research and technology base, system studies, system technology programs, entry systems technology, and experimental programs.

  20. Study on the License Requirements for the SRO/RO of the Kijang Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Subeom; Shin, Taemyung [Korea Nat. University of Transportation, Seoul (Korea, Republic of); Chae, H. T.; Ahn, G. H. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, S. J.; Gam, S. C. [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2016-10-15

    The purpose of the study is to propose an appropriate regulatory position for the Kijang reactor operator license requirement by the review of the applicability and compatibility of HANARO SRO/RO license holders for Kijang reactor operation. As the area using radioactive isotope became gradually enlarged both inside and outside of the country, the Kijang research reactor is planned and now under construction next to the HANARO research reactor now being operated in Taejon. In this paper, therefore, an establishment of revised operator license system is discussed for the new research reactor. The design and operation characteristics of the two (HANARO and Kijang) reactors are concluded to be very similar to each other, however, there still exist slight differences in some minor portions. It is recommendable to allow an independent license for each reactor if two reactors of the same power level have recognizable differences in the design and operation characteristics.

  1. ENGINEERING DEVELOPMENT OF SLURRY BUBBLE COLUMN REACTOR (SBCR) TECHNOLOGY

    Energy Technology Data Exchange (ETDEWEB)

    Bernard A. Toseland

    2000-12-31

    The major technical objectives of this program are threefold: (1) to develop the design tools and a fundamental understanding of the fluid dynamics of a slurry bubble column reactor to maximize reactor productivity, (2) to develop the mathematical reactor design models and gain an understanding of the hydrodynamic fundamentals under industrially relevant process conditions, and (3) to develop an understanding of the hydrodynamics and their interaction with the chemistries occurring in the bubble column reactor. Successful completion of these objectives will permit more efficient usage of the reactor column and tighter design criteria, increase overall reactor efficiency, and ensure a design that leads to stable reactor behavior when scaling up to large-diameter reactors. Washington University's work during the reporting period involved the implementation of the automated calibration device, which will provide an advanced method of determining liquid and slurry velocities at high pressures. This new calibration device is intended to replace the original calibration setup, which depended on fishing lines and hooks to position the radioactive particle. The report submitted by Washington University contains a complete description of the new calibration device and its operation. Improvements to the calibration program are also discussed. Iowa State University utilized air-water bubble column simulations in an effort to determine the domain size needed to represent all of the flow scales in a gas-liquid column at a high superficial velocity. Ohio State's report summarizes conclusions drawn from the completion of gas injection phenomena studies, specifically with respect to the characteristics of bubbling-jetting at submerged single orifices in liquid-solid suspensions.

  2. Review of the status of low power research reactors and considerations for its development

    Energy Technology Data Exchange (ETDEWEB)

    Lim, In Cheol; Wu, Sang Ik; Lee, Byung Chul; Ha, Jae Joo [KAERI, Daejeon (Korea, Republic of)

    2012-10-15

    At present, 232 research reactors in the world are in operation and two thirds of them have a power less than 1 MW. Many countries have used research reactors as the tools for educating and training students or engineers and for scientific service such as neutron activation analysis. As the introduction of a research reactor is considered a stepping stone for a nuclear power development program, many newcomers are considering having a low power research reactor. The IAEA has continued to provide forums for the exchange of information and experiences regarding low power research reactors. Considering these, the Agency is recently working on the preparation of a guide for the preparation of technical specification possibly for a member state to use when wanting to purchase a low power research reactor. In addition, ANS has stated that special consideration should be given to the continued national support to maintain and expand research and test reactor programs and to the efforts in identifying and addressing the future needs by working toward the development and deployment of next generation nuclear research and training facilities. Thus, more interest will be given to low power research reactors and its role as a facility for education and training. Considering these, the status of low power research reactors was reviewed, and some aspects to be considered in developing a low power research reactor were studied.

  3. Virtual reality technology as a tool for human factors requirements evaluation in design of the nuclear reactors control desks

    Energy Technology Data Exchange (ETDEWEB)

    Grecco, Claudio H.S.; Santos, Isaac J.A.L.; Mol, Antonio C.A.; Carvalho, Paulo V.R.; Silva, Antonio C.F.; Ferreira, Francisco J.O.; Dutra, Marco A.M. [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)]. E-mail: grecco@ien.gov.br; luquetti@ien.gov.br; mol@ien.gov.br; paulov@ien.gov.br; tonico@ien.gov.br; fferreira@ien.gov.br; dutra@ien.gov.br

    2007-07-01

    The Virtual Reality (VR) is an advanced computer interface technology that allows the user to internet or to explore a three-dimensional environment through the computer, as was part of the virtual world. This technology presents great applicability in the most diverse areas of the human knowledge. This paper presents a study on the use of the VR as tool for human factors requirements evaluation in design of the nuclear reactors control desks. Moreover, this paper presents a case study: a virtual model of the control desk, developed using virtual reality technology to be used in the human factors requirements evaluation. This case study was developed in the Virtual Reality Laboratory at IEN, and understands the stereo visualization of the Argonauta research nuclear reactor control desk for a static ergonomic evaluation using check-lists, in accordance to the standards and human factors nuclear international guides (IEC 1771, NUREG-0700). (author)

  4. Material characteristics and construction methods for a typical research reactor concrete containment in Iran

    Energy Technology Data Exchange (ETDEWEB)

    Ebrahimia, Mahsa; Suha, Kune Y. [Seoul National Univ., Seoul (Korea, Republic of); Eghbalic, Rahman; Jahan, Farzaneh Asadi malek [School of Architecture and Urbanism, Qazvin (Iran, Islamic Republic of)

    2012-10-15

    Generally selecting an appropriate material and also construction style for a concrete containment due to its function and special geometry play an important role in applicability and also construction cost and duration decrease in a research reactor (RR) project. The reactor containment enclosing the reactor vessel comprises physical barriers reflecting the safety design and construction codes, regulations and standards so as to prevent the community and the environment from uncontrolled release of radioactive materials. It is the third and the last barrier against radioactivity release. It protects the reactor vessel from such external events as earthquake and aircraft crash as well. Thus, it should be designed and constructed in such a manner as to withstand dead and live loads, ground and seismic loads, missiles and aircraft loads, and thermal and shrinkage loads. This study aims to present a construction method for concrete containment of a typical RR in Iran. The work also presents an acceptable characteristic for concrete and reinforcing re bar of a typical concrete containment. The current study has evaluated the various types of the RR containments. The most proper type was selected in accordance with the current knowledge and technology of Iran.

  5. Research and Practice in Green Chemical Technologies

    Institute of Scientific and Technical Information of China (English)

    Yin Yingwu

    2004-01-01

    environmental pollution. Insight has facilities producing nearly 20 photo-initiators and developed the new technique for light-curing lubricity, water-born resin, the technique for water-bon coating and their associated applications. The successfully developed technology of new high performance emulsion will promote the technology advancement in the coating industry and the wide spread of new coating in China.Insight has developed a series of new technologies, including a novel reactor technology - using a newly designed patented blade, has been characterized as high efficiency & energy saving reactor.The novel technology research and development are widely applied in INSIGHT's ten-year practice.We studied and confirmed the non-thermal catalyzing effect of microwaves, proposed a "weak microwave" theory, and a chemical reaction rate equation. We proposed that in the existence of microwave magnesium ions and phosphate, amino acid can be activated to form peptide which maybe the path of the origin of life.A 3000mt/a pilot plant in Chongqing using carbon monoxide and hydrogen to produce dimethyl ether in a single step has been built via collaboration with the Dept. of Chemical Engineering at Tsinghua University. The large scale, low cost production of dimethyl ether may possible through comprehensive technology improvement. This technology will take full advantage of current ammonia production facilities, and try to solve the global fuel shortage problem.

  6. Next Generation Nuclear Plant Reactor Pressure Vessel Materials Research and Development Plan (PLN-2803)

    Energy Technology Data Exchange (ETDEWEB)

    J. K. Wright; R. N. Wright

    2008-04-01

    The U.S. Department of Energy has selected the High Temperature Gas-cooled Reactor design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production. It will have an outlet gas temperature in the range of 900°C and a plant design service life of 60 years. The reactor design will be a graphite moderated, helium-cooled, prismatic, or pebble-bed reactor and use low-enriched uranium, Tri-Isotopic-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The NGNP Materials Research and Development Program is responsible for performing research and development on likely NGNP materials in support of the NGNP design, licensing, and construction activities. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. Studies of potential Reactor Pressure Vessel (RPV) steels have been carried out as part of the pre-conceptual design studies. These design studies generally focus on American Society of Mechanical Engineers (ASME) Code status of the steels, temperature limits, and allowable stresses. Three realistic candidate materials have been identified by this process: conventional light water reactor RPV steels A508/533, 2¼Cr-1Mo in the annealed condition, and modified 9Cr 1Mo ferritic martenistic steel. Based on superior strength and higher temperature limits, the modified 9Cr-1Mo steel has been identified by the majority of design engineers as the preferred choice for the RPV. All of the vendors have

  7. Upgrading of neutron radiography/tomography facility at research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Abd El Bar, Waleed; Mongy, Tarek [Atomic Energy Authority, Cairo (Egypt). ETRR-2; Kardjilov, Nikolay [Helmholtz Zentrum Berlin (HZB) for Materials and Energy, Berlin (Germany)

    2014-03-15

    A state-of-the-art neutron tomography imaging system was set up at the neutron radiography beam tube at the Egypt Second Research Reactor (ETRR-2) and was successfully commissioned in 2013. This study presents a set of tomographic experiments that demonstrate a high quality tomographic image formation. A computer technique for data processing and 3D image reconstruction was used to see inside a copy module of an ancient clay article provided by the International Atomic Energy Agency (IAEA). The technique was also able to uncover tomographic imaging details of a mummified fish and provided a high resolution tomographic image of a defective fire valve. (orig.)

  8. Reduced enrichment for research and test reactors: Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    1993-08-01

    November 9--10, 1978, marked the first of what has become an annual event--the International Meeting on Reduced Enrichment for Research and Test Reactors (RERTR). The meeting brought together for the first time many people who became major program participants in later years. This first meeting emphasized fuel development, and it established the basis for all later meetings. Believing that the proceedings of this first meeting are important as a historical record of the beginning of the international RERTR effort. This report provides presentations and discussions of this original meeting. Individual papers have been cataloged separately.

  9. UCLA research reactor relicensing, or guilty until proven innocent

    Energy Technology Data Exchange (ETDEWEB)

    Wegst, W.F.

    1985-11-01

    This paper briefly reviews the history and experiences of the University of California, Los Angeles (UCLA) in attempting to relicense its 100-kW Argonaut research reactor. The process of intervention in US Nuclear Regulatory Commission (NRC) licensing hearings is briefly reviewed. The intervention in the UCLA case, by an antinuclear group called the Committee to Bridge the Gap (CBG), is described. The outcome of the entire proceeding is summarized and opinions are presented on the validity and viability of the licensing/intervention process.

  10. Oak Ridge National Laboratory Research Reactor Experimenters' Guide

    Energy Technology Data Exchange (ETDEWEB)

    Cagle, C.D. (comp.)

    1982-10-01

    The Oak Ridge National Laboratory has three multipurpose research reactors which accommodate testing loops, target irradiations, and beam-type experiments. Since the experiments must share common or similar facilities and utilities, be designed and fabricated by the same groups, and meet the same safety criteria, certain standards for these have been developed. These standards deal only with those properties from which safety and economy of time and money can be maximized and do not relate to the intent of the experiment or quality of the data obtained. The necessity for, and the limitations of, the standards are discussed; and a compilation of general standards is included.

  11. Preliminary Accident Analyses for Conversion of the Massachusetts Institute of Technology Reactor (MITR) from Highly Enriched to Low Enriched Uranium

    Energy Technology Data Exchange (ETDEWEB)

    Dunn, Floyd E. [Argonne National Lab. (ANL), Argonne, IL (United States); Olson, Arne P. [Argonne National Lab. (ANL), Argonne, IL (United States); Wilson, Erik H. [Argonne National Lab. (ANL), Argonne, IL (United States); Sun, Kaichao S. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Newton, Jr., Thomas H. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Hu, Lin-wen [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)

    2013-09-30

    The Massachusetts Institute of Technology Reactor (MITR-II) is a research reactor in Cambridge, Massachusetts designed primarily for experiments using neutron beam and in-core irradiation facilities. It delivers a neutron flux comparable to current LWR power reactors in a compact 6 MW core using Highly Enriched Uranium (HEU) fuel. In the framework of its non-proliferation policies, the international community presently aims to minimize the amount of nuclear material available that could be used for nuclear weapons. In this geopolitical context most research and test reactors, both domestic and international, have started a program of conversion to the use of LEU fuel. A new type of LEU fuel based on an alloy of uranium and molybdenum (U-Mo) is expected to allow the conversion of U.S. domestic high performance reactors like MITR. This report presents the preliminary accident analyses for MITR cores fueled with LEU monolithic U-Mo alloy fuel with 10 wt% Mo. Preliminary results demonstrate adequate performance, including thermal margin to expected safety limits, for the LEU accident scenarios analyzed.

  12. Outline of the safety research results, in the power reactor field, fiscal year 1996

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-11-01

    The Power Reactor and Nuclear Fuel Development Corporation (PNC) has promoted the safety research in fiscal year of 1996 according to the Fundamental Research on Safety Research (fiscal year 1996 to 2000) prepared on March, 1996. Here is described on the research results in fiscal year 1996, the first year of the 5 years programme, and whole outline of the fundamental research on safety research, on the power reactor field (whole problems on the new nuclear converter and the fast breeder reactor field and problems relating to the power reactor in the safety for earthquake and probability theoretical safety evaluation field). (G.K.)

  13. Management of historical waste from research reactors: the Dutch experience

    Energy Technology Data Exchange (ETDEWEB)

    Van Heek, Aliki; Metz, Bert; Janssen, Bas; Groothuis, Ron [NRG, Petten (Netherlands)

    2013-07-01

    Most radioactive waste emerges as well-defined waste streams from operating power reactors. The management of this is an on-going practice, based on comprehensive (IAEA) guidelines. A special waste category however consists of the historical waste from research reactors, mostly originating from various experiments in the early years of the nuclear era. Removal of the waste from the research site, often required by law, raises challenges: the waste packages must fulfill the acceptance criteria from the receiving storage site as well as the criteria for nuclear transports. Often the aged waste containers do not fulfill today's requirements anymore, and their contents are not well documented. Therefore removal of historical waste requires advanced characterization, sorting, sustainable repackaging and sometimes conditioning of the waste. This paper describes the Dutch experience of a historical waste removal campaign from the Petten High Flux research reactor. The reactor is still in operation, but Dutch legislation asks for central storage of all radioactive waste at the COVRA site in Vlissingen since the availability of the high- and intermediate-level waste storage facility HABOG in 2004. In order to comply with COVRA's acceptance criteria, the complex and mixed inventory of intermediate and low level waste must be characterized and conditioned, identifying the relevant nuclides and their activities. Sorting and segregation of the waste in a Hot Cell offers the possibility to reduce the environmental footprint of the historical waste, by repackaging it into different classes of intermediate and low level waste. In this way, most of the waste volume can be separated into lower level categories not needing to be stored in the HABOG, but in the less demanding LOG facility for low-level waste instead. The characterization and sorting is done on the basis of a combination of gamma scanning with high energy resolution of the closed waste canister and low

  14. The technology development for surveillance test of reactor vessel materials

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai; Choi, Sun Phil; Park, Day Young; Choi, Kwen Jai

    1997-12-01

    Benchmark test was performed in accordance with the requirement of US NRC Reg. Guide DG-1053 for Kori unit-1 in order to determine best-estimated fast neutron fluence irradiated into reactor vessel. Since the uncertainty of radiation analysis comes from the calculation error due to neutron cross-section data, reactor core geometrical dimension, core source, mesh density, angular expansion and convergence criteria, evaluation of calculational uncertainty due to analytical method was performed in accordance with the regulatory guide and the proof was performed for entire analysis by comparing the measurement value obtained by neutron dosimetry located in surveillance capsule. Best-estimated neutron fluence in reactor vessel was calculated by bias factor, neutron flux measurement value/calculational value, from reanalysis result from previous 1st through 4th surveillance testing and finally fluence prediction was performed for the end of reactor life and the entire period of plant life extension. Pressurized thermal shock analysis was performed in accordance with 10 CFR 50.61 using the result of neutron fluence analysis in order to predict the life of reactor vessel material and the criteria of safe operation for Kori unit 1 was reestablished. (author). 55 refs., 55 figs.

  15. Analysis of Nigeria research reactor-1 thermal power calibration methods

    Energy Technology Data Exchange (ETDEWEB)

    Agbo, Sunday Arome; Ahmed, Yusuf Aminu; Ewa, Ita Okon; Jibrin, Yahaya [Ahmadu Bello University, Zaria (Nigeria)

    2016-06-15

    This paper analyzes the accuracy of the methods used in calibrating the thermal power of Nigeria Research Reactor-1 (NIRR-1), a low-power miniature neutron source reactor located at the Centre for Energy Research and Training, Ahmadu Bello University, Zaria, Nigeria. The calibration was performed at three different power levels: low power (3.6 kW), half power (15 kW), and full power (30 kW). Two methods were used in the calibration, namely, slope and heat balance methods. The thermal power obtained by the heat balance method at low power, half power, and full power was 3.7 ± 0.2 kW, 15.2 ± 1.2 kW, and 30.7 ± 2.5 kW, respectively. The thermal power obtained by the slope method at half power and full power was 15.8 ± 0.7 kW and 30.2 ± 1.5 kW, respectively. It was observed that the slope method is more accurate with deviations of 4% and 5% for calibrations at half and full power, respectively, although the linear fit (slope method) on average temperature-rising rates during the thermal power calibration procedure at low power (3.6 kW) is not fitting. As such, the slope method of power calibration is not suitable at lower power for NIRR-1.

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

    Energy Technology Data Exchange (ETDEWEB)

    Rosenbalm, K.F. [comp.

    1995-12-31

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

  17. Research Progress of Isotope Technology

    Institute of Scientific and Technical Information of China (English)

    Department; of; Isotope

    2015-01-01

    Radioactive isotope is one of the origins of nonnuclear power technology.In the 12th Five Year Plan period,CIAE made breakthrough progresses on several important fields such as research and development of preparation of radioactive nuclides,preparation of radioactive source and study of radiopharmaceuticals relied on different financial support,successfully

  18. Technology in Education: Research Says!!

    Science.gov (United States)

    Canuel, Ron

    2011-01-01

    A large amount of research existed in the field of technology in the classroom; however, almost all was focused on the impact of desktop computers and the infamous "school computer room". However, the activities in a classroom represent a multitude of behaviours and interventions, including personal dynamics, classroom management and…

  19. Antineutrino emission and gamma background characteristics from a thermal research reactor

    CERN Document Server

    Bui, V M; Fallot, M; Communeau, V; Cormon, S; Estienne, M; Lenoir, M; Peuvrel, N; Shiba, T; Cucoanes, A S; Elnimr, M; Martino, J; Onillon, A; Porta, A; Pronost, G; Remoto, A; Thiolliere, N; Yermia, F; Zakari-Issoufou, A -A

    2016-01-01

    The detailed understanding of the antineutrino emission from research reactors is mandatory for any high sensitivity experiments either for fundamental or applied neutrino physics, as well as a good control of the gamma and neutron backgrounds induced by the reactor operation. In this article, the antineutrino emission associated to a thermal research reactor: the OSIRIS reactor located in Saclay, France, is computed in a first part. The calculation is performed with the summation method, which sums all the contributions of the beta decay branches of the fission products, coupled for the first time with a complete core model of the OSIRIS reactor core. The MCNP Utility for Reactor Evolution code was used, allowing to take into account the contributions of all beta decayers in-core. This calculation is representative of the isotopic contributions to the antineutrino flux which can be found at research reactors with a standard 19.75\\% enrichment in $^{235}$U. In addition, the required off-equilibrium correction...

  20. Reprocessing of research reactor fuel the Dounreay option

    Energy Technology Data Exchange (ETDEWEB)

    Cartwright, P.

    1997-08-01

    Reprocessing is a proven process for the treatment of spent U/Al Research Reactor fuel. At Dounreay 12679 elements have been reprocessed during the past 30 years. For reactors converting to LEU fuel the uranium recovered in reprocessing can be blended down to less than 20% U{sub 235}, enrichment and be fabricated into new elements. For reactors already converted to LEU it is technically possible to reprocess spent silicide fuel to reduce the U{sub 235} burden and present to a repository only stable conditioned waste. The main waste stream from reprocessing which contains the Fission products is collected in underground storage tanks where it is kept for a period of at least five years before being converted to a stable solid form for return to the country of origin for subsequent storage/disposal. Discharges to the environment from reprocessing are low and are limited to the radioactive gases contained in the spent fuel and a low level liquid waste steam. Both of these discharges are independently monitored, and controlled within strict discharge limits set by the UK Government`s Scottish Office. Transportation of spent fuel to Dounreay has been undertaken using many routes from mainland Europe and has utilised over the past few years both chartered and scheduled vessel services. Several different transport containers have been handled and are currently licensed in the UK. This paper provides a short history of MTR reprocessing at Dounreay, and provides information to show reprocessing can satisfy the needs of MTR operators, showing that reprocessing is a valuable asset in non-proliferation terms, offers a complete solution and is environmentally acceptable.

  1. Relevant thermal-hydraulic aspects in the design of the RRR (Replacement Research Reactor)

    Energy Technology Data Exchange (ETDEWEB)

    Doval, Alicia S.; Mazufri, Claudio M. [INVAP SE, Bariloche (Argentina)

    2002-07-01

    A description of the main thermal-hydraulic features and challenges of the Replacement Research Reactor, for the Australian Nuclear Science and Technology Organization (ANSTO), is presented. Different hydraulic and thermal-hydraulic aspects are considered, core cooling during full power operation and the way it affects the design, design criteria, engineered safety features and computational tools, amongst others. A special section is devoted to the thermal-hydraulic aspects inside the reflector tank, as well as the cooling of irradiation facilities, particularly, the Molybdenum production facility. (author)

  2. Fluoride Salt-Cooled High-Temperature Reactor Technology Development and Demonstration Roadmap

    Energy Technology Data Exchange (ETDEWEB)

    Holcomb, David Eugene [ORNL; Flanagan, George F [ORNL; Mays, Gary T [ORNL; Pointer, William David [ORNL; Robb, Kevin R [ORNL; Yoder Jr, Graydon L [ORNL

    2013-11-01

    Fluoride salt-cooled High-temperature Reactors (FHRs) are an emerging reactor class with potentially advantageous performance characteristics, and fully passive safety. This roadmap describes the principal remaining FHR technology challenges and the development path needed to address the challenges. This roadmap also provides an integrated overview of the current status of the broad set of technologies necessary to design, evaluate, license, construct, operate, and maintain FHRs. First-generation FHRs will not require any technology breakthroughs, but do require significant concept development, system integration, and technology maturation. FHRs are currently entering early phase engineering development. As such, this roadmap is not as technically detailed or specific as would be the case for a more mature reactor class. The higher cost of fuel and coolant, the lack of an approved licensing framework, the lack of qualified, salt-compatible structural materials, and the potential for tritium release into the environment are the most obvious issues that remain to be resolved.

  3. Application Method of Anthropometric Data for Operator Console of Exportable Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Goo Hyun; Lee, Jun Hun; Jeng, Ja Won; Lee, Youn Sang; Kim, Min Gyu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    This paper studied the method to apply the anthropometric data to operator console and large display that used to control room of the exportable research reactor. It is difficult to provide an appropriate operation environment personally to all operators. Therefore, this paper studied method to provide comfortable operation space common to most operators. In the future, it will be possible to enhance the completeness through conformity assessment of the design based on this paper. Therefore, the results of this paper will be an important basic data to design suitable for body size of the user for exportable products such as large display and operator console. Nuclear-related domestic technology has been exported overseas, starting with the JRTR (Jordan Research and Training Reactor) which is currently on its development scheduled to operate in March 2015. It means that Korean nuclear technology has reached the global level already. Therefore, design standards of Human Factors Engineering (HFE) are needed for good products to make more comfortable and suitable for export products. In addition, U. S. Nuclear Regulatory Commission (NRC) reported that the Three Mile Island (TMI) accident in 1979 has been caused by inappropriate design of control panel, human errors, and incorrect procedures. Accordingly, the importance of HFE was raised. In this paper, we studied the application of anthropometric data for operator console and large display of exportable research reactor. Research for nuclear power has been active around the world with environment friendly image. Therefore, it is also very important to study the HFE as a big part in the field of nuclear safety.

  4. System Requirements Analysis for a Computer-based Procedure in a Research Reactor Facility

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jaek Wan; Jang, Gwi Sook; Seo, Sang Moon; Shin, Sung Ki [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    This can address many of the routine problems related to human error in the use of conventional, hard-copy operating procedures. An operating supporting system is also required in a research reactor. A well-made CBP can address the staffing issues of a research reactor and reduce the human errors by minimizing the operator's routine tasks. A CBP for a research reactor has not been proposed yet. Also, CBPs developed for nuclear power plants have powerful and various technical functions to cover complicated plant operation situations. However, many of the functions may not be required for a research reactor. Thus, it is not reasonable to apply the CBP to a research reactor directly. Also, customizing of the CBP is not cost-effective. Therefore, a compact CBP should be developed for a research reactor. This paper introduces high level requirements derived by the system requirements analysis activity as the first stage of system implementation. Operation support tools are under consideration for application to research reactors. In particular, as a full digitalization of the main control room, application of a computer-based procedure system has been required as a part of man-machine interface system because it makes an impact on the operating staffing and human errors of a research reactor. To establish computer-based system requirements for a research reactor, this paper addressed international standards and previous practices on nuclear plants.

  5. Materials technology for an advanced space power nuclear reactor concept: Program summary

    Science.gov (United States)

    Gluyas, R. E.; Watson, G. K.

    1975-01-01

    The results of a materials technology program for a long-life (50,000 hr), high-temperature (950 C coolant outlet), lithium-cooled, nuclear space power reactor concept are reviewed and discussed. Fabrication methods and compatibility and property data were developed for candidate materials for fuel pins and, to a lesser extent, for potential control systems, reflectors, reactor vessel and piping, and other reactor structural materials. The effects of selected materials variables on fuel pin irradiation performance were determined. The most promising materials for fuel pins were found to be 85 percent dense uranium mononitride (UN) fuel clad with tungsten-lined T-111 (Ta-8W-2Hf).

  6. An Account of Oak Ridge National Laboratory's Thirteen Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rosenthal, Murray Wilford [ORNL

    2009-08-01

    The Oak Ridge National Laboratory has built and operated 13 nuclear reactors in its 66-year history. The first was the graphite reactor, the world's first operational nuclear reactor, which served as a plutonium production pilot plant during World War II. It was followed by two aqueous-homogeneous reactors and two red-hot molten-salt reactors that were parts of power-reactor development programs and by eight others designed for research and radioisotope production. One of the eight was an all-metal fast burst reactor used for health physics studies. All of the others were light-water cooled and moderated, including the famous swimming-pool reactor that was copied dozens of times around the world. Two of the reactors were hoisted 200 feet into the air to study the shielding needs of proposed nuclear-powered aircraft. The final reactor, and the only one still operating today, is the High Flux Isotope Reactor (HFIR) that was built particularly for the production of californium and other heavy elements. With the world's highest flux and recent upgrades that include the addition of a cold neutron source, the 44-year-old HFIR continues to be a valuable tool for research and isotope production, attracting some 500 scientific visitors and guests to Oak Ridge each year. This report describes all of the reactors and their histories.

  7. Design and Construction of Operation Bridge for Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Kwangsub; Choi, Jinbok; Lee, Jongmin; Oh, Jinho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    The operation bridge contains a lower working deck mounted on a saddle that travels on rails. Upright members are mounted on the saddle to support the upper structure and two hoist monorails. The saddle contains an anti-derail system that is composed of seismic lugs and guide rollers. The operation bridge travels along the rails to transport the fuel assembly, irradiated object, and reactor components in the pools by using tools. Hoists are installed at the top girder. The hoist is suspended from the monorail by means of a motor driven trolley that runs along the monorail. Movements of hoist and trolley are controlled by using the control pendant switch. Processes of design and construction of the operation bridge for the research reactor are introduced. The operation bridge is designed under consideration of functions of handling equipment in the pool and operational limits for safety. Structural analysis is carried out to evaluate the structural integrity in the seismic events. Tests and inspections are also performed during fabrication and installation to confirm the function and safety of the operation bridge.

  8. Energy Technology Division research summary 1997.

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-10-21

    The Energy Technology Division provides materials and engineering technology support to a wide range of programs important to the US Department of Energy. As shown on the preceding page, the Division is organized into ten sections, five with concentrations in the materials area and five in engineering technology. Materials expertise includes fabrication, mechanical properties, corrosion, friction and lubrication, and irradiation effects. Our major engineering strengths are in heat and mass flow, sensors and instrumentation, nondestructive testing, transportation, and electromechanics and superconductivity applications. The Division Safety Coordinator, Environmental Compliance Officers, Quality Assurance Representative, Financial Administrator, and Communication Coordinator report directly to the Division Director. The Division Director is personally responsible for cultural diversity and is a member of the Laboratory-wide Cultural Diversity Advisory Committee. The Division's capabilities are generally applied to issues associated with energy production, transportation, utilization or conservation, or with environmental issues linked to energy. As shown in the organization chart on the next page, the Division reports administratively to the Associate Laboratory Director (ALD) for Energy and Environmental Science and Technology (EEST) through the General Manager for Environmental and Industrial Technologies. While most of our programs are under the purview of the EEST ALD, we also have had programs funded under every one of the ALDs. Some of our research in superconductivity is funded through the Physical Research Program ALD. We also continue to work on a number of nuclear-energy-related programs under the ALD for Engineering Research. Detailed descriptions of our programs on a section-by-section basis are provided in the remainder of this book. This Overview highlights some major trends. Research related to the operational safety of commercial light water

  9. Investigation of Classification and Design Requirements for Digital Software for Advanced Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Park, Gee Young; Jung, H. S.; Ryu, J. S.; Park, C

    2005-06-15

    As the digital technology is being developed drastically, it is being applied to various industrial instrumentation and control (I and C) fields. In the nuclear power plants, I and C systems are also being installed by digital systems replacing their corresponding analog systems installed previously. There had been I and C systems constructed by analog technology especially for the reactor protection system in the research reactor HANARO. Parallel to the pace of the current trend for digital technology, it is desirable that all I and C systems including the safety critical and non-safety systems in an advanced research reactor is to be installed based on the computer based system. There are many attractable features in using digital systems against existing analog systems in that the digital system has a superior performance for a function and it is more flexible than the analog system. And any fruit gained from the newly developed digital technology can be easily incorporated into the existing digital system and hence, the performance improvement of a computer based system can be implemented conveniently and promptly. Moreover, the capability of high integrity in electronic circuits reduces the electronic components needed to construct the processing device and makes the electronic board simple, and this fact reveals that the hardware failure itself are unlikely to occur in the electronic device other than some electric problems. Balanced the fact mentioned above are the roles and related issues of the software loaded on the digital integrated hardware. Some defects in the course of software development might induce a severe damage on the computer system and plant systems and therefore it is obvious that comprehensive and deep considerations are to be placed on the development of the software in the design of I and C system for use in an advanced research reactor. The work investigates the domestic and international standards on the classifications of digital

  10. Issues and future direction of thermal-hydraulics research and development in nuclear power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Saha, P., E-mail: pradip.saha@ge.com [GE Hitachi Nuclear Energy, Wilmington, NC (United States); Aksan, N. [GRNSPG Group, University of Pisa (Italy); Andersen, J. [GE Hitachi Nuclear Energy, Wilmington, NC (United States); Yan, J. [Westinghouse Electric Co., Columbia, SC (United States); Simoneau, J.P. [AREVA, Lyon (France); Leung, L. [Atomic Energy of Canada Ltd., Chalk River, Ontario (Canada); Bertrand, F. [CEA, DEN, DER, F-13108 Saint-Paul-Lez-Durance (France); Aoto, K.; Kamide, H. [Japan Atomic Energy Agency, Chiyoda-ku, Tokyo (Japan)

    2013-11-15

    The paper archives the proceedings of an expert panel discussion on the issues and future direction of thermal-hydraulic research and development in nuclear power reactors held at the NURETH-14 conference in Toronto, Canada, in September 2011. Thermal-hydraulic issues related to both operating and advanced reactors are presented. Advances in thermal-hydraulics have significantly improved the performance of operating reactors. Further thermal-hydraulics research and development is continuing in both experimental and computational areas for operating reactors, reactors under construction or ready for near-term deployment, and advanced Generation-IV reactors. As the computing power increases, the fine-scale multi-physics computational models, coupled with the systems analysis code, are expected to provide answers to many challenging problems in both operating and advanced reactor designs.

  11. Research technology as a barrier to research

    Directory of Open Access Journals (Sweden)

    J. M. Scheepers

    1994-06-01

    Full Text Available There appears to be considerable concern regarding the progress which masters and doctoral students make in their studies and the underlying causes are thought to relate to the absence of a suitable research climate on campuses. The aim of this study was to identify factors which relate to the research output of tutors. With this in mind a questionnaire was developed and handed to 120 tutors in the human sciences for completion. The means and variances of the variables in the questionnaire were computed/ the variables were intercorrelated and subjected to a principal factor analysis. Four factors were extracted and identified as: knowledge of research technology research output/ knowledge of research methodology and ability to conduct research, and teaching experience in research methodology Two regression analyses were done. The first against a "soft" criterion (perceived ability to conduct empirical research and the second against a "hard" criterion (the real research output of tutors. Multiple correlation coefficients of 0,7257 and 0/5824 respectively, were obtained. Opsomming Die vordering van magister en doktorale studente skyn aansienlike kommer te wek, en die oorsake daarvan word in die afwesigheid van 'n gepaste navorsingsklimaat op kampusse, gesoek. Die doel van hierdie studie was om faktore te identifiseer wat verband hou met die navorsingsuitset van dosente. Met die oog hierop is 'n vraelys opgestel en aan 120 dosente in die geesteswetenskappe vir voltooiing oorhandig. Die gemiddeldes en variansies van die veranderlikes is bereken, die veranderlikes is gemterkorreleer en aan 'n hooffaktorontleding onderwerp. Vier faktore is onttrek en geidentifiseer as: kennis van navorsingstegnologie, navorsingsuitset, kennis van navor-singsmetodologie en die vermoe om navorsing te doen, en onderrigervaring m navorsingsmetodologie. Twee prcgressie-ontledings is gedoen. Die eerste teen 'n "sagte" kriterium (persepsie van dosente van hul vermoe om

  12. Development of fusion blanket technology for the DEMO reactor.

    Science.gov (United States)

    Colling, B R; Monk, S D

    2012-07-01

    The viability of various materials and blanket designs for use in nuclear fusion reactors can be tested using computer simulations and as parts of the test blanket modules within the International Thermonuclear Experimental Reactor (ITER) facility. The work presented here focuses on blanket model simulations using the Monte Carlo simulation package MCNPX (Computational Physics Division Los Alamos National Laboratory, 2010) and FISPACT (Forrest, 2007) to evaluate the tritium breeding capability of a number of solid and liquid breeding materials. The liquid/molten salt breeders are found to have the higher tritium breeding ratio (TBR) and are to be considered for further analysis of the self sufficiency timing. Copyright © 2011 Elsevier Ltd. All rights reserved.

  13. Development of reactor design aid tool using virtual reality technology

    Energy Technology Data Exchange (ETDEWEB)

    Mizuguchi, N. [National Institute for Fusion Science, 322-6 Orochi-cho, Toki, Gifu 509-5292 (Japan)]. E-mail: mizu@nifs.ac.jp; Tamura, Y. [National Institute for Fusion Science, 322-6 Orochi-cho, Toki, Gifu 509-5292 (Japan); Imagawa, S. [National Institute for Fusion Science, 322-6 Orochi-cho, Toki, Gifu 509-5292 (Japan); Sagara, A. [National Institute for Fusion Science, 322-6 Orochi-cho, Toki, Gifu 509-5292 (Japan); Hayashi, T. [National Institute for Fusion Science, 322-6 Orochi-cho, Toki, Gifu 509-5292 (Japan)

    2006-11-15

    A new type of aid system for fusion reactor design, to which the virtual reality (VR) visualization and sonification techniques are applied, is developed. This system provides us with an intuitive interaction environment in the VR space between the observer and the designed objects constructed by the conventional 3D computer-aided design (CAD) system. We have applied the design aid tool to the heliotron-type fusion reactor design activity FFHR2m [A. Sagara, S. Imagawa, O. Mitarai, T. Dolan, T. Tanaka, Y. Kubota, et al., Improved structure and long -life blanket concepts for heliotron reactors, Nucl. Fusion 45 (2005) 258-263] on the virtual reality system CompleXcope [Y. Tamura, A. Kageyama, T. Sato, S. Fujiwara, H. Nakamura, Virtual reality system to visualize and auralize numerical imulation data, Comp. Phys. Comm. 142 (2001) 227-230] of the National Institute for Fusion Science, Japan, and have evaluated its performance. The tool includes the functions of transfer of the observer, translation and scaling of the objects, recording of the operations and the check of interference.

  14. Treatemnt of Wastewater with Modified Sequencing Batch Biofilm Reactor Technology

    Institute of Scientific and Technical Information of China (English)

    胡龙兴; 刘宇陆

    2002-01-01

    This paper describes the removel of COD and nitrogen from wastewater with modified sequencing batch biofilm reactor,The strategy of simultaneous feeding and draining was explored.The results show that introduction of a new batch of wastewater and withdrawal of the purifeid water can be conducted simultaneously with the maximum volumetric exchange rate of about 70%,Application of this feeding and draining mode leads to the reduction of the cycle time,the increase of the utilization of the reactor volume and the simplification of the reactor structure.The treatment of a synthetic wastewater containing COD and nitrogen was investigated.The operation mode of F(D)-O(i.e.,simultaneous feeding and draining followed by the aerobic condition)was adopted.It was found that COD was degraded very fast in the initial reaction period of time,then reduced slowly and the ammonia nitrogen and nitrate nitrogen concentrations decreased and increased with time respectively,while the nitrite nitrogen level increased first and then reduced.The relationship between the COD or ammonia nitrogen loading and its removal rate was examined,and the removal of COD,ammonia nitrogen and total nitrogen could exceed 95%,90%and 80% respectively,The fact that nitrogen could e removed more completely under constant aeration(aerobic condition)of the SBBR operation mode is very interesting and could be explained in several respects.

  15. Investigation of radiations penetration in aluminum used in reactor technology

    Energy Technology Data Exchange (ETDEWEB)

    El-Sayed Abdo, A.; Bashter, I.I.; Makarious, A.S.; Kansouh, W.A. [Atomic Energy Authority, Cairo (Egypt). Nuclear Research Center

    1996-07-01

    Measurements have been carried out to determine the spatial and energy distributions of fast neutrons, total gamma rays and secondary gamma rays in aluminum as a reactor wide structural material experimental work was performed using one of the horizontal channels of the ET-RR-1 reactor. For secondary gamma ray investigations, measurements were carried out for bar, cadmium filtered and boron carbide filtered reactor beams. Aluminum cylinders with a diameter of 100mm and of different thicknesses have been used. Fast neutrons and gamma rays were measured using a neutron-gamma spectrometer with a stilbene scintillator. Discrimination against undesired pulses of neutrons and gamma rays was achieved by a pulse shape discrimination based on zero cross over technique. The measured pulse amplitude distributions of fast neutrons or gamma rays were transformed to neutron or gamma ray energy distributions using the differentiation method and the matrix correction method, respectively. The obtained data were displayed in the form of spatial fluxes and energy distributions and attenuation curves. Total fast neutron macroscopic cross sections, gamma rays linear attenuation coefficients and the relaxation lengths were obtained for the whole energy range and at different energies. (Author).

  16. Updating of PGAA system at HANARO research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jeo, H. J.; Kim, S. H.; Moon, J. H.; Jeong, Y. S.; Kim, Y. J. [KAERI, Taejon (Korea, Republic of)

    2003-10-01

    In this study, updating of Prompt Gamma-ray neutron Activation Analysis system (PGAA) has been carried out to obtain the best, optimal condition through the improvement of neutron flux and reduction of background of PGAA facility which is installed on the ST 1 horizontal beam port at HANARO research reactor. Both diffracted beam profiling's conditions and the neutron diffraction of pyrolytic graphite crystals are investigated by BF{sub 3} counter, laser and optical diffraction angle control method to confirm the beam convergence rate. Also, the effects of interference materials such as aluminum sample holder, teflon holder and Teflon wire appeared from analyzing elemental constituent are investigated with single - and Compton mode. After readjusting of system, the neutron flux measured was 8.1{+-}0.2 x 10{sup 7} n{center_dot}cm{sup -2}{center_dot}s{sup -1} increasing about 40%, to be expected the improved analytical sensitivity.

  17. Radiopharmaceuticals developed at the University of Missouri research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ketring, A.R.; Ehrhardt, G.J. [Univ. of Missouri, Columbia, MO (United States); Day, D.E. [Univ. of Missouri, Rolla, MO (United States)

    1997-12-01

    The University of Missouri Research Reactor (MURR) has put a great deal of effort in the last two decades into development of radiotherapeutic beta emitters as nuclear medicine radiotherapeutics for malignancies. This paper describes the development of two of these drugs, {sup 153}Sm ethylenediaminetetra-methylene phosphonic acid (EDTMP) (Quadramet{trademark}) and {sup 90}Y glass microspheres (TheraSphere{trademark}). Samarium-153 EDTMP is a palliative used to treat the pain of metastatic bone cancer without the side effects of narcotic pain killers. Yttrium-90 glass microspheres are delivered via hepatic artery catheter to embolize the capillaries of liver tumors and deliver a large radiation dose for symptom palliation and life prolonging purposes.

  18. The neutron texture diffractometer at the China Advanced Research Reactor

    Science.gov (United States)

    Li, Mei-Juan; Liu, Xiao-Long; Liu, Yun-Tao; Tian, Geng-Fang; Gao, Jian-Bo; Yu, Zhou-Xiang; Li, Yu-Qing; Wu, Li-Qi; Yang, Lin-Feng; Sun, Kai; Wang, Hong-Li; Santisteban, J. r.; Chen, Dong-Feng

    2016-03-01

    The first neutron texture diffractometer in China has been built at the China Advanced Research Reactor, due to strong demand for texture measurement with neutrons from the domestic user community. This neutron texture diffractometer has high neutron intensity, moderate resolution and is mainly applied to study texture in commonly used industrial materials and engineering components. In this paper, the design and characteristics of this instrument are described. The results for calibration with neutrons and quantitative texture analysis of zirconium alloy plate are presented. The comparison of texture measurements with the results obtained in HIPPO at LANSCE and Kowari at ANSTO illustrates the reliability of the texture diffractometer. Supported by National Nature Science Foundation of China (11105231, 11205248, 51327902) and International Atomic Energy Agency-TC program (CPR0012)

  19. Light Water Reactor Sustainability Program: Digital Technology Business Case Methodology Guide

    Energy Technology Data Exchange (ETDEWEB)

    Thomas, Ken [Idaho National Lab. (INL), Idaho Falls, ID (United States); Lawrie, Sean [ScottMadden, Inc., Raleigh, NC (United States); Hart, Adam [ScottMadden, Inc., Raleigh, NC (United States); Vlahoplus, Chris [ScottMadden, Inc., Raleigh, NC (United States)

    2014-09-01

    The Department of Energy’s (DOE’s) Light Water Reactor Sustainability Program aims to develop and deploy technologies that will make the existing U.S. nuclear fleet more efficient and competitive. The program has developed a standard methodology for determining the impact of new technologies in order to assist nuclear power plant (NPP) operators in building sound business cases. The Advanced Instrumentation, Information, and Control (II&C) Systems Technologies Pathway is part of the DOE’s Light Water Reactor Sustainability (LWRS) Program. It conducts targeted research and development (R&D) to address aging and reliability concerns with the legacy instrumentation and control and related information systems of the U.S. operating light water reactor (LWR) fleet. This work involves two major goals: (1) to ensure that legacy analog II&C systems are not life-limiting issues for the LWR fleet and (2) to implement digital II&C technology in a manner that enables broad innovation and business improvement in the NPP operating model. Resolving long-term operational concerns with the II&C systems contributes to the long-term sustainability of the LWR fleet, which is vital to the nation’s energy and environmental security. The II&C Pathway is conducting a series of pilot projects that enable the development and deployment of new II&C technologies in existing nuclear plants. Through the LWRS program, individual utilities and plants are able to participate in these projects or otherwise leverage the results of projects conducted at demonstration plants. Performance advantages of the new pilot project technologies are widely acknowledged, but it has proven difficult for utilities to derive business cases for justifying investment in these new capabilities. Lack of a business case is often cited by utilities as a barrier to pursuing wide-scale application of digital technologies to nuclear plant work activities. The decision to move forward with funding usually hinges on

  20. Development of in-vessel type control rod drive mechanism for a innovative small reactor (Contract research)

    Energy Technology Data Exchange (ETDEWEB)

    Yoritsune, Tsutomu; Ishida, Toshihisa [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    Although the control rod drive mechanism of an existing large scale light water reactor is generally installed outside the reactor vessel, an in-vessel type control rod drive mechanism (INV-CRDM) is installed inside the reactor vessel. The INV-CRDM contributes to compactness and simplicity of the reactor system, and it can eliminate the possibility of a rod ejection accident. Therefore, INV-CRDM is an important technology adopted in an innovative small reactor. Japan Atomic Energy Research Institute (JAERI) has developed this type of CRDM driven by an electric motor, which can work under high temperature and high pressure water for the advanced marine reactor. On the basis of this research result, a driving motor coil and a bearing were developed to be used under the high temperature steam, severe condition for an innovative small reactor. About the driving motor, we manufactured the driving motor available for high temperature steam and carried out performance test under room temperature atmosphere to confirm the electric characteristic and coolability of the driving coil. With these test results and the past test results under high temperature water, we analyzed and evaluated the electric performance and coolability of the driving coil under high temperature steam. Concerning bearing, we manufactured the test pieces using some candidate material for material characteristic test and carried out the rolling wear test under high temperature steam to select the material. Consequently, we confirmed that performance of the driving coil for the advanced type driving motor, is enough to be used under high temperature steam. And, we evaluated the performance of the bearing and selected the material of the bearing, which can be used under high temperature steam. From these results, we have obtained the prospect that the INV-CRDM can be used for an innovative small reactor under steam atmosphere could be developed. (author)

  1. Development of in-vessel type control rod drive mechanism for a innovative small reactor (Contract research)

    Energy Technology Data Exchange (ETDEWEB)

    Yoritsune, Tsutomu; Ishida, Toshihisa [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    Although the control rod drive mechanism of an existing large scale light water reactor is generally installed outside the reactor vessel, an in-vessel type control rod drive mechanism (INV-CRDM) is installed inside the reactor vessel. The INV-CRDM contributes to compactness and simplicity of the reactor system, and it can eliminate the possibility of a rod ejection accident. Therefore, INV-CRDM is an important technology adopted in an innovative small reactor. Japan Atomic Energy Research Institute (JAERI) has developed this type of CRDM driven by an electric motor, which can work under high temperature and high pressure water for the advanced marine reactor. On the basis of this research result, a driving motor coil and a bearing were developed to be used under the high temperature steam, severe condition for an innovative small reactor. About the driving motor, we manufactured the driving motor available for high temperature steam and carried out performance test under room temperature atmosphere to confirm the electric characteristic and coolability of the driving coil. With these test results and the past test results under high temperature water, we analyzed and evaluated the electric performance and coolability of the driving coil under high temperature steam. Concerning bearing, we manufactured the test pieces using some candidate material for material characteristic test and carried out the rolling wear test under high temperature steam to select the material. Consequently, we confirmed that performance of the driving coil for the advanced type driving motor, is enough to be used under high temperature steam. And, we evaluated the performance of the bearing and selected the material of the bearing, which can be used under high temperature steam. From these results, we have obtained the prospect that the INV-CRDM can be used for an innovative small reactor under steam atmosphere could be developed. (author)

  2. Habitat Technology Research at DLR

    OpenAIRE

    Quantius, Dominik; Schubert, Daniel; Maiwald, Volker; Hauslage, Jens; Bornemann, Gerhild; Waßer, Kai; Hill, Jürgen; Henn, Norbert; Ruyters, Hans-Günter; Braun, Markus

    2013-01-01

    For long duration space missions a closed-loop system which can re-use of materials is mandatory. Also on Earth there are harsh environments or overpopulated areas where a sustainable handling of given goods is indispensable. Addressing these challenges the German Aerospace Center (DLR) conducts research in various fields of habitat technology development, which will be illustrated within this paper. There are various complementary topics, such as coordination and funding of building blocks f...

  3. Status of Fuel Development and Manufacturing for Space Nuclear Reactors at BWX Technologies

    Science.gov (United States)

    Carmack, W. J.; Husser, D. L.; Mohr, T. C.; Richardson, W. C.

    2004-02-01

    New advanced nuclear space propulsion systems will soon seek a high temperature, stable fuel form. BWX Technologies Inc (BWXT) has a long history of fuel manufacturing. UO2, UCO, and UCx have been fabricated at BWXT for various US and international programs. Recent efforts at BWXT have focused on establishing the manufacturing techniques and analysis capabilities needed to provide a high quality, high power, compact nuclear reactor for use in space nuclear powered missions. To support the production of a space nuclear reactor, uranium nitride has recently been manufactured by BWXT. In addition, analytical chemistry and analysis techniques have been developed to provide verification and qualification of the uranium nitride production process. The fabrication of a space nuclear reactor will require the ability to place an unclad fuel form into a clad structure for assembly into a reactor core configuration. To this end, BWX Technologies has reestablished its capability for machining, GTA welding, and EB welding of refractory metals. Specifically, BWX Technologies has demonstrated GTA welding of niobium flat plate and EB welding of niobium and Nb-1Zr tubing. In performing these demonstration activities, BWX Technologies has established the necessary infrastructure to manufacture UO2, UCx, or UNx fuel, components, and complete reactor assemblies in support of space nuclear programs.

  4. Flow through reactors for organic chemistry: directly electrically heated tubular mini reactors as an enabling technology for organic synthesis

    Directory of Open Access Journals (Sweden)

    Ulrich Kunz

    2009-11-01

    Full Text Available Until recently traditional heating in organic chemistry has been done with oil heating baths or using electric heat exchangers. With the advent of microwave equipment, heating by microwaves was rapidly introduced as standard method in organic chemistry laboratories, mainly because of the convenient possibility to operate at high temperature accompanied by accelerated reaction rates. In the present contribution we discuss the method of heating small, continuously operated reactors by passing electric current directly through the reactor wall as an enabling technology in organic chemistry. The benefit of this method is that the heat is generated directly inside the reactor wall. By this means high heating rates comparable to microwave ovens can be reached but at much lower cost for the equipment. A tool for the comparison of microwave heating and traditional heating is provided. As an example kinetic data for the acid catalyzed hydrolysis of methyl formate were measured using this heating concept. The reaction is not only a suitable model but also one of industrial importance since this is the main production process for formic acid.

  5. Flow through reactors for organic chemistry: directly electrically heated tubular mini reactors as an enabling technology for organic synthesis.

    Science.gov (United States)

    Kunz, Ulrich; Turek, Thomas

    2009-11-30

    Until recently traditional heating in organic chemistry has been done with oil heating baths or using electric heat exchangers. With the advent of microwave equipment, heating by microwaves was rapidly introduced as standard method in organic chemistry laboratories, mainly because of the convenient possibility to operate at high temperature accompanied by accelerated reaction rates. In the present contribution we discuss the method of heating small, continuously operated reactors by passing electric current directly through the reactor wall as an enabling technology in organic chemistry. The benefit of this method is that the heat is generated directly inside the reactor wall. By this means high heating rates comparable to microwave ovens can be reached but at much lower cost for the equipment. A tool for the comparison of microwave heating and traditional heating is provided. As an example kinetic data for the acid catalyzed hydrolysis of methyl formate were measured using this heating concept. The reaction is not only a suitable model but also one of industrial importance since this is the main production process for formic acid.

  6. Flow through reactors for organic chemistry: directly electrically heated tubular mini reactors as an enabling technology for organic synthesis

    Science.gov (United States)

    Turek, Thomas

    2009-01-01

    Summary Until recently traditional heating in organic chemistry has been done with oil heating baths or using electric heat exchangers. With the advent of microwave equipment, heating by microwaves was rapidly introduced as standard method in organic chemistry laboratories, mainly because of the convenient possibility to operate at high temperature accompanied by accelerated reaction rates. In the present contribution we discuss the method of heating small, continuously operated reactors by passing electric current directly through the reactor wall as an enabling technology in organic chemistry. The benefit of this method is that the heat is generated directly inside the reactor wall. By this means high heating rates comparable to microwave ovens can be reached but at much lower cost for the equipment. A tool for the comparison of microwave heating and traditional heating is provided. As an example kinetic data for the acid catalyzed hydrolysis of methyl formate were measured using this heating concept. The reaction is not only a suitable model but also one of industrial importance since this is the main production process for formic acid. PMID:20300506

  7. Thermal-Hydraulic Research Review and Cooperation Outcome for Light Water Reactor Fuel

    Energy Technology Data Exchange (ETDEWEB)

    In, Wang Kee; Shin, Chang Hwan; Lee, Chan; Chun, Tae Hyun; Oh, Dong Seok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, Chi Young [Pukyong Nat’l Univ., Busan (Korea, Republic of)

    2016-12-15

    The fuel assembly for pressurized water reactor (PWR) consists of fuel rod bundle, spacer grid and bottom/top end fittings. The cooling water in high pressure and temperature is introduced in lower plenum of reactor core and directed to upper plenum through the subchannel which is formed between the fuel rods. The main thermalhydraulic performance parameters for the PWR fuel are pressure drop and critical heat flux in normal operating condition, and quenching time in accident condition. The Korea Atomic Energy Research Institute (KAERI) has been developing an advanced PWR fuel, dual-cooled annular fuel and accident tolerant fuel for the enhancement of fuel performance and the localization. For the key thermal-hydraulic technology development of PWR fuel, the KAERI LWR fuel team has conducted the experiments for pressure drop, turbulent flow mixing and heat transfer, critical heat flux(CHF) and quenching. The computational fluid dynamics (CFD) analysis was also performed to predict flow and heat transfer in fuel assembly including the spent fuel assembly in dry cask for interim repository. In addition, the research cooperation with university and nuclear fuel company was also carried out to develop a basic thermalhydraulic technology and the commercialization.

  8. Innovative technology transfer of nondestructive evaluation research

    Science.gov (United States)

    Brian Brashaw; Robert J. Ross; Xiping Wang

    2008-01-01

    Technology transfer is often an afterthought for many nondestructive evaluation (NDE) researchers. Effective technology transfer should be considered during the planning and execution of research projects. This paper outlines strategies for using technology transfer in NDE research and presents a wide variety of technology transfer methods used by a cooperative...

  9. Research in Antenna Technology, Radar Technology and Electromagnetic Scattering Phenomena

    Science.gov (United States)

    2015-04-06

    AFRL-OSR-VA-TR-2015-0081 Research in Antenna Technology John Schindler ARCON CORP Final Report 04/06/2015 DISTRIBUTION A: Distribution approved for...a group of six researchers in the fields of electromagnetics, radar and antenna technology. Research was conducted during this reporting period in...Standard Form 298 (Rev. 8-98) Prescribed by ANSI Std. Z39.18 Research in Antenna technology, Radar Technology and Electromagnetic Scattering Phenomena

  10. Groundwater Monitoring Plan for the Reactor Technology Complex Operable Unit 2-13

    Energy Technology Data Exchange (ETDEWEB)

    Richard P. Wells

    2007-03-23

    This Groundwater Monitoring Plan describes the objectives, activities, and assessments that will be performed to support the on-going groundwater monitoring requirements at the Reactor Technology Complex, formerly the Test Reactor Area (TRA). The requirements for groundwater monitoring were stipulated in the Final Record of Decision for Test Reactor Area, Operable Unit 2-13, signed in December 1997. The monitoring requirements were modified by the First Five-Year Review Report for the Test Reactor Area, Operable Unit 2-13, at the Idaho National Engineering and Environmental Laboratory to focus on those contaminants of concern that warrant continued surveillance, including chromium, tritium, strontium-90, and cobalt-60. Based upon recommendations provided in the Annual Groundwater Monitoring Status Report for 2006, the groundwater monitoring frequency was reduced to annually from twice a year.

  11. 75 FR 57080 - In the Matter of Aerotest Operations, Inc. (Aerotest Radiography and Research Reactor); Order...

    Science.gov (United States)

    2010-09-17

    ... COMMISSION In the Matter of Aerotest Operations, Inc. (Aerotest Radiography and Research Reactor); Order... Aerotest Operations, Inc., (Aerotest, the licensee) is the holder of Facility Operating License No. R-98 which authorizes the possession, use, and operation of the Aerotest Radiography and Research Reactor...

  12. 75 FR 39985 - In the Matter of Aerotest Operations, Inc. (Aerotest Radiography and Research Reactor); Order...

    Science.gov (United States)

    2010-07-13

    ... COMMISSION In the Matter of Aerotest Operations, Inc. (Aerotest Radiography and Research Reactor); Order Approving Indirect Transfer of Facility Operating License and Conforming Amendment I. Aerotest Operations..., use and operation of the Aerotest Radiography and Research Reactor (ARRR) located in San Ramon...

  13. 78 FR 46618 - Order Prohibiting Operation of Aerotest Radiography and Research Reactor

    Science.gov (United States)

    2013-08-01

    ... COMMISSION Order Prohibiting Operation of Aerotest Radiography and Research Reactor I. Aerotest Operations... Licensing of Production and Utilization Facilities.'' The license authorizes the operation of the Aerotest Radiography and Research Reactor (ARRR) in accordance with the conditions specified therein. The ARRR is...

  14. Water Reactor Safety Research Division. Quarterly progress report, April 1-June 30, 1980

    Energy Technology Data Exchange (ETDEWEB)

    Abuaf, N.; Levine, M.M.; Saha, P.; van Rooyen, D.

    1980-08-01

    The Water Reactor Safety Research Programs quarterly report describes current activities and technical progress in the programs at Brookhaven National Laboratory sponsored by the USNRC Division of Reactor Safety Research. The projects reported each quarter are the following: LWR Thermal Hydraulic Development, Advanced Code Evlauation, TRAC Code Assessment, and Stress Corrosion Cracking of PWR Steam Generator Tubing.

  15. Implementation of a management system for operating organizations of research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kibrit, Eduardo, E-mail: kibrit@ctmsp.mar.mil.b [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), Sao Paulo, SP (Brazil); Aquino, Afonso Rodrigues de; Zouain, Desiree Moraes, E-mail: araquino@ipen.b, E-mail: dmzouain@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    This paper presents the requirements established by an IAEA draft technical document for the implementation of a management system for operating organisations of research reactors. The following aspects will be discussed: structure of IAEA draft technical document, management system requirements, processes common to all research reactors, aspects for the implementation of the management system, and a formula for grading the management system requirements. (author)

  16. Implementation of a management system for operating organizations of research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kibrit, Eduardo, E-mail: kibrit@ctmsp.mar.mil.b [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), Sao Paulo, SP (Brazil); Aquino, Afonso Rodrigues de; Zouain, Desiree Moraes, E-mail: araquino@ipen.b, E-mail: dmzouain@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    This paper presents the requirements established by an IAEA draft technical document for the implementation of a management system for operating organisations of research reactors. The following aspects will be discussed: structure of IAEA draft technical document, management system requirements, processes common to all research reactors, aspects for the implementation of the management system, and a formula for grading the management system requirements. (author)

  17. Hydrogen Technology Research at SRNL

    Energy Technology Data Exchange (ETDEWEB)

    Danko, E.

    2011-02-13

    The Savannah River National Laboratory (SRNL) is a U.S. Department of Energy research and development laboratory located at the Savannah River Site (SRS) near Aiken, South Carolina. SRNL has over 50 years of experience in developing and applying hydrogen technology, both through its national defense activities as well as through its recent activities with the DOE Hydrogen Programs. The hydrogen technical staff at SRNL comprises over 90 scientists, engineers and technologists. SRNL has ongoing R&D initiatives in a variety of hydrogen storage areas, including metal hydrides, complex hydrides, chemical hydrides and carbon nanotubes. SRNL has over 25 years of experience in metal hydrides and solid-state hydrogen storage research, development and demonstration. As part of its defense mission at SRS, SRNL developed, designed, demonstrated and provides ongoing technical support for the largest hydrogen processing facility in the world based on the integrated use of metal hydrides for hydrogen storage, separation, and compression. The SRNL has been active in teaming with academic and industrial partners to advance hydrogen technology. A primary focus of SRNL's R&D has been hydrogen storage using metal and complex hydrides. SRNL and its Hydrogen Technology Research Laboratory have been very successful in leveraging their defense infrastructure, capabilities and investments to help solve this country's energy problems. SRNL has participated in projects to convert public transit and utility vehicles for operation using hydrogen fuel. Two major projects include the H2Fuel Bus and an Industrial Fuel Cell Vehicle (IFCV) also known as the GATOR{trademark}. Both of these projects were funded by DOE and cost shared by industry. These are discussed further in Section 3.0, Demonstration Projects. In addition to metal hydrides technology, the SRNL Hydrogen group has done extensive R&D in other hydrogen technologies, including membrane filters for H2 separation, doped carbon

  18. Advanced reactor material research requirements; Necesidades de investigacion en materiales para reactores avanzados

    Energy Technology Data Exchange (ETDEWEB)

    Greene, C. A.; Muscara, J.; Srinivasan, M.

    2003-07-01

    The metal and graphite components used in high temperature gas-cooled reactors (HTGR) may suffer physical-chemical alterations, irradiation damage and mechanical alterations. Their failure may call the security of these reactors into question by affecting the integrity of the pressure control system, core geometry or its cooling, among other aspects. This article analyses the work currently being done in the matter by the US Nuclear Regulatory Commission. (Author)

  19. Reactor technology. Progress report, July-September 1980

    Energy Technology Data Exchange (ETDEWEB)

    Breslow, M. (ed.)

    1980-12-01

    Progress in the Space Power Advanced Reactor (SPAR) Program includes indications that revision of the BeO reflector configuration can reduce system weight. Observed boiling limit restrictions on the performance of the annular-wick core heat pipe have accelerated transition to the development of the target-design arterial heat pipe. Successful bends of core heat pipes have been made with sodium as the mandrel material. With the phasing out of the GCFR Program, work on the Low Power Safety Experiments Program is now concentrated on completion of the third 37-rod Full Length Subgroup test. In the Reactor Safety/Structural Analysis area, effort on the Category I Structures Program is toward developing an experimental test plan focusing on a specific structural design. Buckling experiments on thin-walled cylindrical shells with circular cutouts are reported. Results of a three-dimensional analysis of thermal stresses in the Fort St. Vrain core support block are presented. Materials investigations and operation of a molybdenum-core SiC heat pipe are reported. Entrainment limits for gravity-assisted heat pipes and heat pipe configurations for application to energy conservation are being investigated. The new solution critical assembly, SHEBA, was completed. Godiva IV was temporarily relocated at TA-15. Influence of scattered radiations in the test vault on InRad measurements was determined from detector scans of the vault produced by /sup 252/Cf neutron and /sup 137/Cs gamma sources.

  20. Moving bed biofilm reactor technology: process applications, design, and performance.

    Science.gov (United States)

    McQuarrie, James P; Boltz, Joshua P

    2011-06-01

    The moving bed biofilm reactor (MBBR) can operate as a 2- (anoxic) or 3-(aerobic) phase system with buoyant free-moving plastic biofilm carriers. These systems can be used for municipal and industrial wastewater treatment, aquaculture, potable water denitrification, and, in roughing, secondary, tertiary, and sidestream applications. The system includes a submerged biofilm reactor and liquid-solids separation unit. The MBBR process benefits include the following: (1) capacity to meet treatment objectives similar to activated sludge systems with respect to carbon-oxidation and nitrogen removal, but requires a smaller tank volume than a clarifier-coupled activated sludge system; (2) biomass retention is clarifier-independent and solids loading to the liquid-solids separation unit is reduced significantly when compared with activated sludge systems; (3) the MBBR is a continuous-flow process that does not require a special operational cycle for biofilm thickness, L(F), control (e.g., biologically active filter backwashing); and (4) liquid-solids separation can be achieved with a variety of processes, including conventional and compact high-rate processes. Information related to system design is fragmented and poorly documented. This paper seeks to address this issue by summarizing state-of-the art MBBR design procedures and providing the reader with an overview of some commercially available systems and their components.

  1. Under-sodium viewing technology for improvement of fast-reactor safeguards

    Energy Technology Data Exchange (ETDEWEB)

    Beddingfield, David H [Los Alamos National Laboratory; Gerhart, Jeremy J [Los Alamos National Laboratory; Kawakubo, Yoko [JAEA

    2009-01-01

    The current safeguards approach for fast reactors relies exclusively on maintenance of continuity of knowledge to track the movement of fuel assemblies through these facilities. The remote handling of fuel assemblies, the visual opacity of the liquid metal coolant. and the chemical reactivity of sodium all combine and result in significant limitations on the available options to verify fuel assembly identification numbers or the integrity of these assemblies. These limitations also serve to frustrate attempts to restore the continuity-of-knowledge in instances where the information is under a variety of scenarios. The technology of ultrasonic under-sodium viewing offers new options to the safeguards community for recovering continuity-of-knowledge and applying more traditional item accountancy to fast reactor facilities. We have performed a literature review to investigate the development of under-sodium viewing technologies. In this paper we will summarize our findings and report the state of development of this technology and we will present possible applications to the fast reactor system to improve the existing safeguards approach at these reactors and in future fast reactors.

  2. Teaching Sodium Fast Reactor Technology and Operation for the Present and Future Generations of SFR Users

    OpenAIRE

    Christian, Latge; Rodriguez, Gilles; Baque, Francois; Leclerc, Arnaud; Martin, Laurent; Vray, Bernard; Romanetti, Pascale

    2011-01-01

    International audience; This paper provides a description of the education and training activities related to sodium fast reactors, carried out respectively in the French Sodium and Liquid Metal School (ESML) created in 1975 and located in France (at the CEA Cadarache Research Centre), in the Fast Reactor Operation and Safety School (FROSS) created in 2005 at the Phenix plant, and in the Institut National des Sciences et Techniques Nucle'aires (INSTN). It presents their recent developments an...

  3. Status of reduced enrichment programs for research reactors in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Kanda, Keiji; Nishihara, Hedeaki [Kyoto Univ., Osaka (Japan); Shirai, Eiji; Oyamada, Rokuro; Sanokawa, Konomo [Japan Atomic Energy Research Institute, Tokyo (Japan)

    1997-08-01

    The reduced enrichment programs for the JRR-2, JRR-3, JRR-4 and JMTR of Japan Atomic Energy Research Institute (JAERI), and the KUR of Kyoto University Research Reactor Institute (KURRI) have been partially completed and are mostly still in progress under the Joint Study Programs with Argonne National Laboratory (ANL). The JMTR and JRR-2 have been already converted to use MEU aluminide fuels in 1986 and 1987, respectively. The operation of the upgraded JRR-3(JRR-3M) has started in March 1990 with the LEU aluminide fuels. Since May 1992, the two elements have been inserted in the KUR. The safety review application for the full core conversion to use LEU silicide in the JMTR was approved in February 1992 and the conversion has been done in January 1994. The Japanese Government approved a cancellation of the KUHFR Project in February 1991, and in April 1994 the U.S. Government gave an approval to utilize HEU in the KUR instead of the KUHFR. Therefore, the KUR will be operated with HEU fuel until 2001. Since March 1994, Kyoto University is continuing negotiation with UKAEA Dounreay on spent fuel reprocessing and blending down of recovered uranium, in addition to that with USDOE.

  4. Irradiation Tests Supporting LEU Conversion of Very High Power Research Reactors in the US

    Energy Technology Data Exchange (ETDEWEB)

    Woolstenhulme, N. E.; Cole, J. I.; Glagolenko, I.; Holdaway, K. K.; Housley, G. K.; Rabin, B. H.

    2016-10-01

    The US fuel development team is developing a high density uranium-molybdenum alloy monolithic fuel to enable conversion of five high-power research reactors. Previous irradiation tests have demonstrated promising behavior for this fuel design. A series of future irradiation tests will enable selection of final fuel fabrication process and provide data to qualify the fuel at moderately-high power conditions for use in three of these five reactors. The remaining two reactors, namely the Advanced Test Reactor and High Flux Isotope Reactor, require additional irradiation tests to develop and demonstrate the fuel’s performance with even higher power conditions, complex design features, and other unique conditions. This paper reviews the program’s current irradiation testing plans for these moderately-high irradiation conditions and presents conceptual testing strategies to illustrate how subsequent irradiation tests will build upon this initial data package to enable conversion of these two very-high power research reactors.

  5. Progress of Research on Demonstration Fast Reactor Main Pipe Material

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

    The main characteristics of the sodium pipe system in demonstration fast reactor are high-temperature, thin-wall and big-caliber, which is different from the high-pressure and thick-wall of the pressurized water reactor system, and the system is long-term

  6. Modeling of operating history of the research nuclear reactor

    Science.gov (United States)

    Naymushin, A.; Chertkov, Yu; Shchurovskaya, M.; Anikin, M.; Lebedev, I.

    2016-06-01

    The results of simulation of the IRT-T reactor operation history from 2012 to 2014 are presented. Calculations are performed using continuous energy Monte Carlo code MCU-PTR. Comparison is made between calculation and experimental data for the critical reactor.

  7. Researches on a reactor core in heavy ion inertial fusion

    CERN Document Server

    Kondo, S; Iinuma, T; Kubo, K; Kato, H; Kawata, S; Ogoyski, A I

    2016-01-01

    In this paper a study on a fusion reactor core is presented in heavy ion inertial fusion (HIF), including the heavy ion beam (HIB) transport in a fusion reactor, a HIB interaction with a background gas, reactor cavity gas dynamics, the reactor gas backflow to the beam lines, and a HIB fusion reactor design. The HIB has remarkable preferable features to release the fusion energy in inertial fusion: in particle accelerators HIBs are generated with a high driver efficiency of ~30-40%, and the HIB ions deposit their energy inside of materials. Therefore, a requirement for the fusion target energy gain is relatively low, that would be ~50 to operate a HIF fusion reactor with a standard energy output of 1GW of electricity. In a fusion reactor the HIB charge neutralization is needed for a ballistic HIB transport. Multiple mechanical shutters would be installed at each HIB port at the reactor wall to stop the blast waves and the chamber gas backflow, so that the accelerator final elements would be protected from the ...

  8. Research and development on next generation reactor (phase I)

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong Kyoon; Chang, Moon Heuy; Hwang, Yung Dong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)] [and others

    1994-10-01

    The objective of the study is to improve the volume of nuclear power plant which adopts passive safety system concept. The passive safety system reactor is characterized by excellent safety and reliability. But the volume of NSSS (Nuclear Steam Supply System) of the passive safety system reactor is so small that it should be upgraded for commercial operation. For volume upgrade, detailed analyses are performed as follows; core design, hydraulics, design and mechnical structures, and safety analysis. In addition to above analysis, some investigations must be supplied as follows: power density vs. DNB margin decrease, outlet temperature vs. EPRI-URD, additional tests for upgraded reactor, dynamic analysis of mechanical vibration according to expanded reactor vessel and expanded in-core structures, and Merit loss of passive safety system reactor according to design margin decrease. (Author).

  9. Design and R&D Progress of China Lead-Based Reactor for ADS Research Facility

    Directory of Open Access Journals (Sweden)

    Yican Wu

    2016-03-01

    Full Text Available In 2011, the Chinese Academy of Sciences launched an engineering project to develop an accelerator-driven subcritical system (ADS for nuclear waste transmutation. The China Lead-based Reactor (CLEAR, proposed by the Institute of Nuclear Energy Safety Technology, was selected as the reference reactor for ADS development, as well as for the technology development of the Generation IV lead-cooled fast reactor. The conceptual design of CLEAR-I with 10 MW thermal power has been completed. KYLIN series lead-bismuth eutectic experimental loops have been constructed to investigate the technologies of the coolant, key components, structural materials, fuel assembly, operation, and control. In order to validate and test the key components and integrated operating technology of the lead-based reactor, the lead alloy-cooled non-nuclear reactor CLEAR-S, the lead-based zero-power nuclear reactor CLEAR-0, and the lead-based virtual reactor CLEAR-V are under realization.

  10. Reactor

    Science.gov (United States)

    Evans, Robert M.

    1976-10-05

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

  11. The application of research reactor Maria for analysis of thorium use in nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Chwaszczewski, S.; Andrzejewski, K.; Myslek-Laurikainen, B.; Pytel, B.; Szczurek, J. [Dep. Thorium Project, Institute of Atomic Energy POLATOM, 05-400 Otwock-Swierk (Poland); Polkowska-Motrenko, H. [Institute of Nuclear Chemistry and Technology, ul.Dorodna 16 03-195 Warszawa (Poland)

    2010-07-01

    The MARIA reactor, pool-type light-water cooled and beryllium moderated nuclear research reactor was used to evaluate the {sup 233}U breeding during the experimental irradiation of the thorium samples. The level of impurities concentrations was determined using ICP-MS method. The associated development of computer programs for analysis of application of thorium in EPR reactor consist of PC version of CORD-2/GNOMER system are presented. (authors)

  12. Research about reactor operator's personability characteristics and performance

    Energy Technology Data Exchange (ETDEWEB)

    Wei Li; He Xuhong; Zhao Bingquan [Tsinghua Univ., Institute of Nuclear Energy Technology, Beijing (China)

    2003-03-01

    To predict and evaluate the reactor operator's performance by personality characteristics is an important part of reactor operator safety assessment. Using related psychological theory combined with the Chinese operator's fact and considering the effect of environmental factors to personality analysis, paper does the research about the about the relationships between reactor operator's performance and personality characteristics, and offers the reference for operator's selection, using and performance in the future. (author)

  13. Research Agenda in Cloud Technologies

    CERN Document Server

    Sriram, Ilango

    2010-01-01

    Cloud computing is the latest effort in delivering computing resources as a service. It represents a shift away from computing as a product that is purchased, to computing as a service that is delivered to consumers over the internet from large-scale data centres - or "clouds". Whilst cloud computing is gaining growing popularity in the IT industry, academia appeared to be lagging behind the rapid developments in this field. This paper is the first systematic review of peer-reviewed academic research published in this field, and aims to provide an overview of the swiftly developing advances in the technical foundations of cloud computing and their research efforts. Structured along the technical aspects on the cloud agenda, we discuss lessons from related technologies; advances in the introduction of protocols, interfaces, and standards; techniques for modelling and building clouds; and new use-cases arising through cloud computing.

  14. Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research and Development by the Light Water Reactor Sustainability Program

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Cyrus M [ORNL; Nanstad, Randy K [ORNL; Clayton, Dwight A [ORNL; Matlack, Katie [Georgia Institute of Technology; Ramuhalli, Pradeep [Pacific Northwest National Laboratory (PNNL); Light, Glenn [Southwest Research Institute, San Antonio

    2012-09-01

    The Department of Energy s (DOE) Light Water Reactor Sustainability (LWRS) Program is a five year effort which works to develop the fundamental scientific basis to understand, predict, and measure changes in materials and systems, structure, and components as they age in environments associated with continued long-term operations of existing commercial nuclear power reactors. This year, the Materials Aging and Degradation (MAaD) Pathway of this program has placed emphasis on emerging Non-Destructive Evaluation (NDE) methods which support these objectives. DOE funded Research and Development (R&D) on emerging NDE techniques to support commercial nuclear reactor sustainability is expected to begin next year. This summer, the MAaD Pathway invited subject matter experts to participate in a series of workshops which developed the basis for the research plan of these DOE R&D NDE activities. This document presents the results of one of these workshops which are the DOE LWRS NDE R&D Roadmap for Reactor Pressure Vessels (RPV). These workshops made a substantial effort to coordinate the DOE NDE R&D with that already underway or planned by the Electric Power Research Institute (EPRI) and the Nuclear Regulatory Commission (NRC) through their representation at these workshops.

  15. QUARTERLY PROGRESS REPORT JANUARY, FEBRUARY, MARCH, 1967 REACTOR FUELS AND MATERIALS DEVELOPMENT PROGRAMS FOR FUELS AND MATERIALS BRANCH OF USAEC DEVISION OF REACTOR DEVELOPMENT AND TECHNOLOGY

    Energy Technology Data Exchange (ETDEWEB)

    Albaugh, F. W.; Bush, S. H.; Cadwell, J. J.; de Halas, D. R.; Worlton, D. C.

    1967-06-01

    Work is reported in the areas of: fast fuels oxides and nitrides; nuclear ceramics; nuclear graphite; basic swelling studies; irradiation damage to reactor metals; ATR gas loop operation and maintenance; metallic fuels; nondestructive testing research; and fast reactor dosimetry and damage analysis.

  16. Current status of neutron activation analysis in HANARO Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Yong Sam; Moon, Jong Hwa; Sohn, Jae Min [Korea Atomic Energy Research Institute, Daejeon (Korea)

    2003-03-01

    The facilities for neutron activation analysis in the HANARO (Hi-flux Advanced Neutron Application Research Reactor) are described and the main applications of NAA (Neutron Activation Analysis) are reviewed. The sample irradiation tube, automatic and manual pneumatic transfer system were installed at three irradiation holes of HANARO at the end of 1995. The performance of the NAA facility was examined to identify the characteristics of the tube transfer system, irradiation sites and custom-made polyethylene irradiation capsule. The available thermal neutron fluxes at irradiation sites are in the range of 3 x 10{sup 13} - 1 x 10{sup 14} n/cm{sup 2}{center_dot}s and cadmium ratios are in 15 - 250. For an automatic sample changer for gamma-ray counting, a domestic product was designed and manufactured. An integrated computer program (Labview) to analyse the content was developed. In 2001, PGNAA (Prompt Gamma Neutron Activation Analysis) facility has been installed using a diffracted neutron beam of ST1. NAA has been applied in the trace component analysis of nuclear, geological, biological, environmental and high purity materials, and various polymers for research and development. The improvement of analytical procedures and establishment of an analytical quality control and assurance system were studied. Applied research and development for the environment, industry and human health by NAA and its standardization were carried out. For the application of the KOLAS (Korea Laboratory Accreditation Scheme), evaluation of measurement uncertainty and proficiency testing of reference materials were performed. Also to verify the reliability and to validate analytical results, intercomparison studies between laboratories were carried out. (author)

  17. Research on the usage of a deep sea fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Otsubo, Akira; Kowata, Yasuki [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1997-09-01

    Many new types of fast reactors have been studied in PNC. A deep sea fast reactor has the highest realization probability of the reactors studied because its development is desired by many specialists of oceanography, meteorology, deep sea bottom oil field, seismology and so on and because the development does not cost big budget and few technical problems remain to be solved. This report explains the outline and the usage of the reactor of 40 kWe and 200 to 400 kWe. The reactor can be used as a power source at an unmanned base for long term climate prediction and the earth science and an oil production base in a deep sea region. On the other hand, it is used for heat and electric power supply to a laboratory in the polar region. In future, it will be used in the space. At the present time, a large FBR development plan does not proceed successfully and a realization goal time of FBR has gone later and later. We think that it is the most important to develop the reactor as fast as possible and to plant a fast reactor technique in our present society. (author)

  18. Development of MMIS design technology for integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Koo, In Soo; Park, H. Y.; Lee, C. K. and others

    1999-03-01

    The objective of this study is to establish the concept design of the SMART MMIS through the setup of a structure and design concept of MMIS, the development of design methodology, and essential technologies for MMIS considering computer based digital technologies and human factors. Main activities of this study divided into three categories such as the development of conceptual design and requirement of the SMART I and C, the development of conceptual design and requirement of the SMART MMI, and essential technology for the MMIS. The results of main activities that have been performed during the study are as follows; 1) The review of licensing requirements and operating experience of existing plants. 2) The establishment of a system design concept about development strategy, basic functions and requirements, and applicable technology through the adoption of digital technology and human factorsengineering. 3) The accomplishment of function, structure, and design requirement of each MMIS through the performance of conceptual design on system structure and actuation logic using function analysis. It is necessary to develop and apply new technologies for the MMIS design compatible with the design basis of SMART. Considered essential technologies for the MMIS in this study are 1) the development of a signal validation algorithm and sensor diagnosis technique 2) the development of a soft controller for operator's action 3) the establishment of a data communication structure for MMIS architecture 4) the development of an information optimization basis and automatic start-up algorithm 5) the development of sensor reduction requirements 6) the development of a qualification method of digital equipment for applied digital technology and commercial grade items. (author)

  19. Jordan's First Research Reactor Project: Driving Forces, Present Status and the Way Ahead

    Energy Technology Data Exchange (ETDEWEB)

    Xoubi, Ned, E-mail: Ned@Xoubi.co [Jordan Atomic Energy Commission (JAEC), P.O.Box 70, Shafa Badran, 11934 Amman (Jordan)

    2011-07-01

    In a gigantic step towards establishing Jordan's nuclear power program, Jordan Atomic Energy Commission (JAEC) is building the first nuclear research and test reactor in the Kingdom. The new reactor will serve as the focal point for Jordan Center for Nuclear Research (JCNR), a comprehensive state of the art nuclear center not only for Jordan but for the whole region, the center will include in addition to the reactor a radioisotopes production plant, a nuclear fuel fabrication plant, a cold neutron source (CNS), a radioactive waste treatment facility, and education and training center. The JRTR reactor is the only research reactor new build worldwide in 2010, it is a 5 MW light water open pool multipurpose reactor, The reactor core is composed of 18 fuel assemblies, MTR plate type, with 19.75% enriched uranium silicide (U{sub 3}Si{sub 2}) in an aluminum matrix. It is reflected on all sides by beryllium and graphite blocks. Reactor power is upgradable to 10 MW with a maximum thermal flux of 1.45x10{sup 14} cm{sup -2}s{sup -1}. The reactor reactivity is controlled by four Hafnium Control Absorber Rods (CAR). Jordan Center for Nuclear Research is located in Ramtha city, it is owned by Jordan Atomic Energy Commission (JAEC), and is contracted to Korea Atomic Energy Research Institute (KAERI) and Daewoo E and C. The JCNR project is a 56 months EPC fixed price contract for the design engineering, construction, and commissioning the JCNR reactor, and other nuclear facilities. The project presents many challenges for both the owner and the contractor, being the first nuclear reactor for Jordan, and the first nuclear export for Korea. The driving forces, present status and the way ahead will be presented in this paper. (author)

  20. A neutron tomography facility at a low power research reactor

    CERN Document Server

    Körner, S; Von Tobel, P; Rauch, H

    2001-01-01

    Neutron radiography (NR) provides a very efficient tool in the field of non-destructive testing as well as for many applications in fundamental research. A neutron beam penetrating a specimen is attenuated by the sample material and detected by a two-dimensional (2D) imaging device. The image contains information about materials and structure inside the sample because neutrons are attenuated according to the basic law of radiation attenuation. Contrary to X-rays, neutrons can be attenuated by some light materials, as for example, hydrogen and boron, but penetrate many heavy materials. Therefore, NR can yield important information not obtainable by more traditional methods. Nevertheless, there are many aspects of structure, both quantitative and qualitative, that are not accessible from 2D transmission images. Hence, there is an interest in three-dimensional neutron imaging. At the 250 kW TRIGA Mark II reactor of the Atominstitut in Austria a neutron tomography facility has been installed. The neutron flux at ...

  1. Characterization of Novel Calorimeters in the Annular Core Research Reactor *

    Directory of Open Access Journals (Sweden)

    Hehr Brian D.

    2016-01-01

    Full Text Available A series of pulsed irradiation experiments have been performed in the central cavity of Sandia National Laboratories' Annular Core Research Reactor (ACRR to characterize the responses of a set of elemental calorimeter materials including Si, Zr, Sn, Ta, W, and Bi. Of particular interest was the perturbing effect of the calorimeter itself on the ambient radiation field – a potential concern in dosimetry applications. By placing the calorimeter package into a neutron-thermalizing lead/polyethylene (LP bucket and irradiating both with and without a cadmium wrapper, it was demonstrated that prompt capture gammas generated inside the calorimeters can be a significant contributor to the measured dose in the active disc region. An MCNP model of the experimental setup was shown to replicate measured dose responses to within 10%. The internal (n,γ contribution was found to constitute as much as 50% of the response inside the LP bucket and up to 20% inside the nominal (unmodified cavity environment, with Ta and W exhibiting the largest enhancement due to their sizable (n,γ cross sections. Capture reactions in non-disc components of the calorimeter were estimated to be responsible for up to a few percent of the measured response.

  2. Current status, research progress and future plan of Kartini research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sardjono, Y.; Syarip; Tjiptono, T.W. [Yogyakarta Nuclear Research Center, Batan (Indonesia)

    1999-10-01

    The current status, research progress and future plan of the Kartini Research Reactor (KRR) is presented. The measurements of axial burn-up distributions for each fuel element by gamma scanning techniques, core axial power distribution display, fuel management for safeguards purpose as well as some research progress activities i.e.; utilization of beamport for: neutron radiography, application neutron activation analysis and history record of KRR power operations is also presented. The KRR is 100 kW pool water reactor type which uses natural circulation and provided by: five beamports in which one of them already coupled with natural uranium subcritical assembly, two thermalizing columns in which one of them is prepared for developing Boron Neutron Capture Therapy (BNCT), two rabbit systems utilized for special analysis uranium ore by delayed neutron counting techniques, one center timbre and 40 irradiation rack (lazy susan) for neutron activation analysis. The KRR was constructed as a second research reactor in Indonesia with special purpose for training and education, high safety margin with involve in high negative temperature coefficient which achieved its first criticality on January 25, 1979. The maximum power level on first criticality is 50 kW and since August 1981 up to now is operating 100 kW. Base on the KRR design limit, it is planned to increase the power level up to 250 kW in the future plan. The preliminary activities such as Non Destructive Testing (NDT) for some reactor components especially water tank and thermal column should be done before decided to increase power level. (author)

  3. Technology Development Roadmap for the Advanced High Temperature Reactor Secondary Heat Exchanger

    Energy Technology Data Exchange (ETDEWEB)

    P. Sabharwall; M. McCllar; A. Siahpush; D. Clark; M. Patterson; J. Collins

    2012-09-01

    This Technology Development Roadmap (TDRM) presents the path forward for deploying large-scale molten salt secondary heat exchangers (MS-SHX) and recognizing the benefits of using molten salt as the heat transport medium for advanced high temperature reactors (AHTR). This TDRM will aid in the development and selection of the required heat exchanger for: power production (the first anticipated process heat application), hydrogen production, steam methane reforming, methanol to gasoline production, or ammonia production. This TDRM (a) establishes the current state of molten salt SHX technology readiness, (b) defines a path forward that systematically and effectively tests this technology to overcome areas of uncertainty, (c) demonstrates the achievement of an appropriate level of maturity prior to construction and plant operation, and (d) identifies issues and prioritizes future work for maturing the state of SHX technology. This study discusses the results of a preliminary design analysis of the SHX and explains the evaluation and selection methodology. An important engineering challenge will be to prevent the molten salt from freezing during normal and off-normal operations because of its high melting temperature (390°C for KF ZrF4). The efficient transfer of energy for industrial applications depends on the ability to incorporate cost-effective heat exchangers between the nuclear heat transport system and industrial process heat transport system. The need for efficiency, compactness, and safety challenge the capabilities of existing heat exchanger technology. The description of potential heat exchanger configurations or designs (such as printed circuit, spiral or helical coiled, ceramic, plate and fin, and plate type) were covered in an earlier report (Sabharwall et al. 2011). Significant future work, much of which is suggested in this report, is needed before the benefits and full potential of the AHTR can be realized. The execution of this TDRM will focuses

  4. Numerical Simulation of Flow Field in Flow-guide Tank of China Advanced Research Reactor

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    The flow-guide tank of China advanced research reactor (CARR) is located at the top of the reactor vessel and connected with the inlet coolant pipe. It acts as a reactor inlet coolant distributor and plays an important role in reducing the flow-induced vibration of the internal components of the reactor core. Several designs of the flow-guide tank have been proposed, however, the final design option has to be made after detailed investigation of the velocity profile within the flow-guide tank for each configuration.

  5. Burnup measurements on spent fuel elements of the RP-10 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Vela Mora, Mariano; Gallardo Padilla, Alberto; Palomino, Jose Luis Castro, E-mail: mvela@ipen.gob.p [Instituto Peruano de Energia Nuclear (IPEN/Peru), Lima (Peru). Grupo de Calculo, Analisis y Seguridad de Reactores; Terremoto, Luis Antonio Albiac, E-mail: laaterre@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    This work describes the measurement, using nondestructive gamma-ray spectroscopy, of the average burnup attained by Material Testing Reactor (MTR) fuel elements irradiated in the RP-10 research reactor. Measurements were performed at the reactor storage pool area using {sup 137}Cs as the only burnup monitor, even for spent fuel elements with cooling times much shorter than two years. The experimental apparatus was previously calibrated in efficiency to obtain absolute average burnup values, which were compared against corresponding ones furnished by reactor physics calculations. The mean deviation between both values amounts to 6%. (author)

  6. Nuclear energy was the way of the future; 50 anniversary of the research reactor

    NARCIS (Netherlands)

    Wassink, J.

    2013-01-01

    It was the hidden jewel of TU Delft, according to the employees of the nuclear reactor. Others protested against it and insisted that it be eliminated. Following a major mid-life crisis, the Delft research reactor is now in better shape than ever before.

  7. Highest average burnups achieved by MTR fuel elements of the IEA-R1 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Damy, Margaret A.; Terremoto, Luis A.A.; Silva, Jose E.R.; Silva, Antonio Teixeira e; Castanheira, Myrthes; Teodoro, Celso A. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil). Centro de Engenharia Nuclear (CEN)]. E-mail: madamy@ipen.br

    2007-07-01

    Different nuclear fuels were employed in the manufacture of plate type at IPEN , usually designated as Material Testing Reactor (MTR) fuel elements. These fuel elements were used at the IEA-R1 research reactor. This work describes the main characteristics of these nuclear fuels, emphasizing the highest average burn up achieved by these fuel elements. (author)

  8. Status report on the Small Secure Transportable Autonomous Reactor (SSTAR) /Lead-cooled Fast Reactor (LFR) and supporting research and development.

    Energy Technology Data Exchange (ETDEWEB)

    Sienicki, J. J.; Moisseytsev, A.; Yang, W. S.; Wade, D. C.; Nikiforova, A.; Hanania, P.; Ryu, H. J.; Kulesza, K. P.; Kim, S. J.; Halsey, W. G.; Smith, C. F.; Brown, N. W.; Greenspan, E.; de Caro, M.; Li, N.; Hosemann, P.; Zhang, J.; Yu, H.; Nuclear Engineering Division; LLNL; LANL; Massachusetts Inst. of Tech.; Ecole des Mines de Paris; Oregon State Univ.; Univ.of California at Berkley

    2008-06-23

    This report provides an update on development of a pre-conceptual design for the Small Secure Transportable Autonomous Reactor (SSTAR) Lead-Cooled Fast Reactor (LFR) plant concept and supporting research and development activities. SSTAR is a small, 20 MWe (45 MWt), natural circulation, fast reactor plant for international deployment concept incorporating proliferation resistance for deployment in non-fuel cycle states and developing nations, fissile self-sufficiency for efficient utilization of uranium resources, autonomous load following making it suitable for small or immature grid applications, and a high degree of passive safety further supporting deployment in developing nations. In FY 2006, improvements have been made at ANL to the pre-conceptual design of both the reactor system and the energy converter which incorporates a supercritical carbon dioxide Brayton cycle providing higher plant efficiency (44 %) and improved economic competitiveness. The supercritical CO2 Brayton cycle technology is also applicable to Sodium-Cooled Fast Reactors providing the same benefits. One key accomplishment has been the development of a control strategy for automatic control of the supercritical CO2 Brayton cycle in principle enabling autonomous load following over the full power range between nominal and essentially zero power. Under autonomous load following operation, the reactor core power adjusts itself to equal the heat removal from the reactor system to the power converter through the large reactivity feedback of the fast spectrum core without the need for motion of control rods, while the automatic control of the power converter matches the heat removal from the reactor to the grid load. The report includes early calculations for an international benchmarking problem for a LBE-cooled, nitride-fueled fast reactor core organized by the IAEA as part of a Coordinated Research Project on Small Reactors without Onsite Refueling; the calculations use the same neutronics

  9. Research Supporting Satellite Communications Technology

    Science.gov (United States)

    Horan Stephen; Lyman, Raphael

    2005-01-01

    This report describes the second year of research effort under the grant Research Supporting Satellite Communications Technology. The research program consists of two major projects: Fault Tolerant Link Establishment and the design of an Auto-Configurable Receiver. The Fault Tolerant Link Establishment protocol is being developed to assist the designers of satellite clusters to manage the inter-satellite communications. During this second year, the basic protocol design was validated with an extensive testing program. After this testing was completed, a channel error model was added to the protocol to permit the effects of channel errors to be measured. This error generation was used to test the effects of channel errors on Heartbeat and Token message passing. The C-language source code for the protocol modules was delivered to Goddard Space Flight Center for integration with the GSFC testbed. The need for a receiver autoconfiguration capability arises when a satellite-to-ground transmission is interrupted due to an unexpected event, the satellite transponder may reset to an unknown state and begin transmitting in a new mode. During Year 2, we completed testing of these algorithms when noise-induced bit errors were introduced. We also developed and tested an algorithm for estimating the data rate, assuming an NRZ-formatted signal corrupted with additive white Gaussian noise, and we took initial steps in integrating both algorithms into the SDR test bed at GSFC.

  10. Membrane pumping technology for helium and hydrogen isotope separation in the fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Pistunovich, V.I. [Kurchatov Inst., Moscow (Russian Federation). NFI RRC; Pigarov, A.Yu. [Kurchatov Inst., Moscow (Russian Federation). NFI RRC; Busnyuk, A.O. [Bonch-Bruyevich University, St. Petersburg (Russian Federation); Livshits, A.I. [Bonch-Bruyevich University, St. Petersburg (Russian Federation); Notkin, M.E. [Bonch-Bruyevich University, St. Petersburg (Russian Federation); Samartsev, A.A. [Bonch-Bruyevich University, St. Petersburg (Russian Federation); Borisenko, K.L. [Efremov Institute, St. Petersburg (Russian Federation); Darmogray, V.V. [Efremov Institute, St. Petersburg (Russian Federation); Ershov, B.D. [Efremov Institute, St. Petersburg (Russian Federation); Filippova, L.V. [Efremov Institute, St. Petersburg (Russian Federation); Mudugin, B.G. [Efremov Institute, St. Petersburg (Russian Federation); Odintsov, V.N. [Efremov Institute, St. Petersburg (Russian Federation); Saksagansky, G.L. [Efremov Institute, St. Petersburg (Russian Federation); Serebrennikov, D.V. [Efremov Institute, St. Petersburg (Russian Federation)

    1995-03-01

    A gas pumping system for ITER, improved by implementation of superpermeable membranes for selective hydrogen isotope exhaust, is considered. A study of the pumping capability of a niobium membrane for a hydrogen-helium mixture has been performed.Monte Carlo simulations of gas behaviour for the experimental facility and fusion reactor have been done.The scheme of the ITER pumping system with the membranes and membrane pumping technology was considered. The conceptual study the membrane pump for the ITER was done. This work gives good prospects for the membrane pumping use in ITER to reduce the total inventory of tritium necessary for reactor operation. (orig.).

  11. Under-Sodium Viewing: A Review of Ultrasonic Imaging Technology for Liquid Metal Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Griffin, Jeffrey W.; Peters, Timothy J.; Posakony, Gerald J.; Chien, Hual-Te; Bond, Leonard J.; Denslow, Kayte M.; Sheen, Shuh-Haw; Raptis, Paul

    2009-03-27

    This current report is a summary of information obtained in the "Information Capture" task of the U.S. DOE-funded "Under Sodium Viewing (USV) Project." The goal of the multi-year USV project is to design, build, and demonstrate a state-of-the-art prototype ultrasonic viewing system tailored for periodic reactor core in-service monitoring and maintenance inspections. The study seeks to optimize system parameters, improve performance, and re-establish this key technology area which will be required to support any new U.S. liquid-metal cooled fast reactors.

  12. Perspectives for Fluidized Bed Nuclear Reactor Technology using Rotating Fluidized Beds in a Static Geometry

    Science.gov (United States)

    Broqueville, Axel De; Wilde, Juray De

    The new concept of a rotating fluidized bed in a static geometry opens perspectives for fluidized bed nuclear reactor technology and is experimentally and numerically investigated. With conventional fluidized bed technology, the maximum attainable power is rather limited and maximum at a certain fluidization gas flow rate. Using a rotating fluidized bed in a static geometry, the fluidization gas drives both the centrifugal force and the counteracting radial gas-solid drag force in a similar way. This allows operating the reactor at any chosen sufficiently high solids loading over a much wider fluidization gas flow rate range and in particular at much higher fluidization gas flow rates than with conventional fluidized bed reactor technology, offering increased flexibility with respect to cooling via the fluidization gas. Furthermore, the centrifugal force can be a multiple of earth gravity, allowing radial gas-solid slip velocities much higher than in conventional fluidized beds. The latter result in gas-solid heat transfer coefficients one or multiple orders of magnitude higher than in conventional fluidized beds. The combination of dense operation and high fluidization gas flow rates allows process intensification and a more compact reactor design.

  13. Regulatory Risk Reduction for Advanced Reactor Technologies – FY2016 Status and Work Plan Summary

    Energy Technology Data Exchange (ETDEWEB)

    Moe, Wayne Leland [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-08-01

    Millions of public and private sector dollars have been invested over recent decades to realize greater efficiency, reliability, and the inherent and passive safety offered by advanced nuclear reactor technologies. However, a major challenge in experiencing those benefits resides in the existing U.S. regulatory framework. This framework governs all commercial nuclear plant construction, operations, and safety issues and is highly large light water reactor (LWR) technology centric. The framework must be modernized to effectively deal with non-LWR advanced designs if those designs are to become part of the U.S energy supply. The U.S. Department of Energy’s (DOE) Advanced Reactor Technologies (ART) Regulatory Risk Reduction (RRR) initiative, managed by the Regulatory Affairs Department at the Idaho National Laboratory (INL), is establishing a capability that can systematically retire extraneous licensing risks associated with regulatory framework incompatibilities. This capability proposes to rely heavily on the perspectives of the affected regulated community (i.e., commercial advanced reactor designers/vendors and prospective owner/operators) yet remain tuned to assuring public safety and acceptability by regulators responsible for license issuance. The extent to which broad industry perspectives are being incorporated into the proposed framework makes this initiative unique and of potential benefit to all future domestic non-LWR applicants

  14. Status of neutron beam utilization at the Dalat nuclear research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Dien, Nguyen Nhi; Hai, Nguyen Canh [Nuclear Research Institute, Dalat (Viet Nam)

    2003-03-01

    The 500-kW Dalat nuclear research reactor was reconstructed from the USA-made 250-kW TRIGA Mark II reactor. After completion of renovation and upgrading, the reactor has been operating at its nominal power since 1984. The reactor is used mainly for radioisotope production, neutron activation analysis, neutron beam researches and reactor physics study. In the framework of the reconstruction and renovation project of the 1982-1984 period, the reactor core, the control and instrumentation system, the primary and secondary cooling systems, as well as other associated systems were newly designed and installed by the former Soviet Union. Some structures of the reactor, such as the reactor aluminum tank, the graphite reflector, the thermal column, horizontal beam tubes and the radiation concrete shielding have been remained from the previous TRIGA reactor. As a typical configuration of the TRIGA reactor, there are four neutron beam ports, including three radial and one tangential. Besides, there is a large thermal column. Until now only two-neutron beam ports and the thermal column have been utilized. Effective utilization of horizontal experimental channels is one of the important research objectives at the Dalat reactor. The research program on effective utilization of these experimental channels was conducted from 1984. For this purpose, investigations on physical characteristics of the reactor, neutron spectra and fluxes at these channels, safety conditions in their exploitation, etc. have been carried out. The neutron beams, however, have been used only since 1988. The filtered thermal neutron beams at the tangential channel have been extracted using a single crystal silicon filter and mainly used for prompt gamma neutron activation analysis (PGNAA), neutron radiography (NR) and transmission experiments (TE). The filtered quasi-monoenergetic keV neutron beams using neutron filters at the piercing channel have been used for nuclear data measurements, study on

  15. Rapid Determination of Optimal Conditions in a Continuous Flow Reactor Using Process Analytical Technology

    Directory of Open Access Journals (Sweden)

    Michael F. Roberto

    2013-12-01

    Full Text Available Continuous flow reactors (CFRs are an emerging technology that offer several advantages over traditional batch synthesis methods, including more efficient mixing schemes, rapid heat transfer, and increased user safety. Of particular interest to the specialty chemical and pharmaceutical manufacturing industries is the significantly improved reliability and product reproducibility over time. CFR reproducibility can be attributed to the reactors achieving and maintaining a steady state once all physical and chemical conditions have stabilized. This work describes the implementation of a smart CFR with univariate physical and multivariate chemical monitoring that allows for rapid determination of steady state, requiring less than one minute. Additionally, the use of process analytical technology further enabled a significant reduction in the time and cost associated with offline validation methods. The technology implemented for this study is chemistry and hardware agnostic, making this approach a viable means of optimizing the conditions of any CFR.

  16. Recent contributions to fusion reactor design and technology development

    Energy Technology Data Exchange (ETDEWEB)

    1979-11-01

    The report contains a collection of 16 recent fusion technology papers on the STARFIRE Project, the study of alternate fusion fuel cycles, a maintainability study, magnet safety, neutral beam power supplies and pulsed superconducting magnets and energy transfer. This collection of papers contains contributions for Argonne National Laboratory, McDonnell Douglas Astronautics Company, General Atomic Company, The Ralph M. Parsons Company, the University of Illinois, and the University of Wisconsin. Separate abstracts are presented for each paper. (MOW)

  17. Conceptual Nuclear Design of a 20 MW Multipurpose Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Chul Gyo; Kim, Hak Sung; Park, Cheol [KAERI, Daejeon (Korea, Republic of); Nghiem, Huynh Ton; Vinh, Le Vinh; Dang, Vo Doan Hai [Dalat Nuclear Research Reactor, Hanoi (Viet Nam)

    2007-08-15

    A conceptual nuclear design of a 20 MW multi-purpose research reactor for Vietnam has been jointly done by the KAERI and the DNRI (VAEC). The AHR reference core in this report is a right water cooled and a heavy water reflected open-tank-in-pool type multipurpose research reactor with 20 MW. The rod type fuel of a dispersed U{sub 3}Si{sub 2}-Al with a density of 4.0 gU/cc is used as a fuel. The core consists of fourteen 36-element assemblies, four 18-element assemblies and has three in-core irradiation sites. The reflector tank filled with heavy water surrounds the core and provides rooms for various irradiation holes. Major analyses have been done for the relevant nuclear design parameters such as the neutron flux and power distributions, reactivity coefficients, control rod worths, etc. For the analysis, the MCNP, MVP, and HELIOS codes were used by KAERI and DNRI (VAEC). The results by MCNP (KAERI) and MVP (DNRI) showed good agreements and can be summarized as followings. For a clean, unperturbed core condition such that the fuels are all fresh and there are no irradiation holes in the reflector region, the fast neutron flux (E{sub n}{>=}1.0 MeV) reaches 1.47x10{sup 14} n/cm{sup 2}s and the maximum thermal neutron flux (E{sub n}{<=}0.625 eV) reaches 4.43x10{sup 14} n/cm{sup 2}s in the core region. In the reflector region, the thermal neutron peak occurs about 28 cm far from the core center and the maximum thermal neutron flux is estimated to be 4.09x10{sup 14} n/cm{sup 2}s. For the analysis of the equilibrium cycle core, the irradiation facilities in the reflector region were considered. The cycle length was estimated as 38 days long with a refueling scheme of replacing three 36-element fuel assemblies or replacing two 36-element and one 18-element fuel assemblies. The excess reactivity at a BOC was 103.4 mk, and 24.6 mk at a minimum was reserved at an EOC. The assembly average discharge burnup was 54.6% of initial U-235 loading. For the proposed fuel management

  18. Research and Development Methodology for Practical Use of Accident Tolerant Fuel in Light Water Reactors

    Directory of Open Access Journals (Sweden)

    Masaki Kurata

    2016-02-01

    Full Text Available Research and development (R&D methodology for the practical use of accident tolerant fuel (ATF in commercial light water reactors is discussed in the present review. The identification and quantification of the R&D-metrics and the attribute of candidate ATF-concepts, recognition of the gap between the present R&D status and the targeted practical use, prioritization of the R&D, and technology screening schemes are important for achieving a common understanding on technology screening process among stakeholders in the near term and in developing an efficient R&D track toward practical use. Technology readiness levels and attribute guides are considered to be proper indices for these evaluations. In the midterm, the selected ATF-concepts will be developed toward the technology readiness level-5, at which stage the performance of the prototype fuel rods and the practicality of industrial scale fuel manufacturing will be verified and validated. Regarding the screened-out concepts, which are recognized to have attractive potentials, the fundamental R&D should be continued in the midterm to find ways of addressing showstoppers.

  19. Cutting Technology for Decommissioning of the Reactor Pressure Vessels in Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Kwan Seong; Kim, Geun Ho; Moon, Jei Kwon; Choi, Byung Seon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-05-15

    Lots of nuclear power plants have been decommissioned during the last 2 decades. An essential part of this work is the dismantling of the Reactor Pressure Vessel and its Internals. For this purpose a wide variety of different cutting technologies have been developed, adapted and applied. A detailed introduction to Plasma Arc cutting, Contact Arc Metal cutting and Abrasive Water Suspension Jet cutting is given, as it turned out that these cutting technologies are particularly suitable for these type of segmentation work. A comparison of these technologies including gaseous emissions, cutting power, manipulator requirements as well as selected design approaches are given. Process limits as well as actual limits of application are presented

  20. Reactor pressure vessel integrity research at the Oak Ridge National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Corwin, W.R.; Pennell, W.E.; Pace, J.V.

    1995-12-31

    Maintaining the integrity of the reactor pressure vessel (RPV) in a light-water-cooled nuclear power plant is crucial in preventing and controlling severe accidents that have the potential for major contamination release. The RPV is the only key safety-related component of the plant for which a duplicate or redundant backup system does not exist. It is therefore imperative to understand and be able to predict the integrity inherent in the RPV. For this reason, the U.S. Nuclear Regulatory Commission has established the related research programs at ORNL described herein to provide for the development and confirmation of the methods used for: (1) establishing the irradiation exposure conditions within the RPV in the Embrittlement Data Base and Dosimetry Evaluation Program, (2) assessing the effects of irradiation on the RPV materials in the Heavy-Section Steel Irradiation Program, and (3) developing overall structural and fracture analyses of RPVs in the Heavy-Section Steel Technology Program.

  1. High-temperature gas-cooled reactor technology development program. Annual progress report for period ending December 31, 1982

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.; Rittenhouse, P.L.; Bartine, D.E.; Sanders, J.P.

    1983-06-01

    During 1982 the High-Temperature Gas-Cooled Reactor (HTGR) Technology Program at Oak Ridge National Laboratory (ORNL) continued to develop experimental data required for the design and licensing of cogeneration HTGRs. The program involves fuels and materials development (including metals, graphite, ceramic, and concrete materials), HTGR chemistry studies, structural component development and testing, reactor physics and shielding studies, performance testing of the reactor core support structure, and HTGR application and evaluation studies.

  2. High-temperature gas-cooled reactor technology development program. Annual progress report for period ending December 31, 1982

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.; Rittenhouse, P.L.; Bartine, D.E.; Sanders, J.P.

    1983-06-01

    During 1982 the High-Temperature Gas-Cooled Reactor (HTGR) Technology Program at Oak Ridge National Laboratory (ORNL) continued to develop experimental data required for the design and licensing of cogeneration HTGRs. The program involves fuels and materials development (including metals, graphite, ceramic, and concrete materials), HTGR chemistry studies, structural component development and testing, reactor physics and shielding studies, performance testing of the reactor core support structure, and HTGR application and evaluation studies.

  3. Present status of neutron beam facilities at the research reactor, HANARO, and its future prospect

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chang-Hee; Kang, Young-Hwan; Kuk, Il-Hiun [Korea Atomic Energy Research Institute, Taejon (Korea)

    2001-03-01

    Korea has been operating its new research reactor, HANARO, since its first criticality in 1995. It is an open-tank-in-pool type reactor using LEU fuel with thermal neutron flux of 2 x 10{sup 14} nominally at the nose in the D{sub 2}O reflector having 7 horizontal beam ports and a provision of vertical hole for cold neutron source installation. KAERI has pursued an extensive instrument development program since 1992 by the support of the nuclear long-term development program of the government and there are now 4 working instruments. A high resolution powder diffractometer and a neutron radiography facility has been operational since late 1997 and 1996, respectively. A four-circle diffractometer has been fully working since mid 1999 and a small angle neutron spectrometer is just under commissioning phase. With the development of linear position sensitive detector with delay-line readout electronics, we have developed a residual stress instrument as an optional machine to the HRPD for last two years. Around early 1998 informal users program started with friendly users and it became a formal users support program by the ministry of science and technology. Short description for peer group formation and users activities is given. (author)

  4. Proceedings of the 1997 workshop on the utilization of research reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-10-01

    The 1997 Workshop on the Utilization of Research Reactors, which is the sixth Workshop on the theme of research reactor utilization was held in Bandung, Indonesia from November 6 to 13. This Workshop was executed based on the agreement in the Eighth International conference for Nuclear Cooperation in Asia (ICNCA) held in Tokyo, March 1997. The whole Workshop consists of the preceding Sub-workshop carried out the demonstration experiment of Radioisotope Production, and the Workshop on the theme of three fields (Neutron Scattering, Radioisotope production, Safe Operation and Maintenance of Research Reactor). The total number of participants for the workshop was about 100 people from 8 countries, i.e. China, Indonesia, Korea, Malaysia, Philippine, Thailand, Vietnam and Japan. It consists of the papers for Sub-workshop, Neutron Scattering, Radioisotope Production, Safe Operation and Maintenance of research reactor, and summary reports. The 53 of the presented papers are indexed individually. (J.P.N.)

  5. Review of the nuclear reactor thermal hydraulic research in ocean motions

    Energy Technology Data Exchange (ETDEWEB)

    Yan, B.H., E-mail: yanbh3@mail.sysu.edu.cn

    2017-03-15

    The research and development of small modular reactor in floating platform has been strongly supported by Chinese government and enterprises. Due to the effect of ocean waves, the thermal hydraulic behavior and safety characteristics of floating reactor are different from that of land-based reactor. Many scholars including the author have published their research and results in open literatures. Much of these literatures are valuable but there are also some contradictory conclusions. In this wok, the nuclear reactor thermal hydraulic research in ocean motions was systematically summarized. Valuable results and experimental data were analyzed and classified. Inherent mechanism for controversial issues in different experiments was explained. Necessary work needed in the future was suggested. Through this work, we attempt to find as many valuable results as possible for the designing and subsequent research.

  6. McCARD for Neutronics Design and Analysis of Research Reactor Cores

    Science.gov (United States)

    Shim, Hyung Jin; Park, Ho Jin; Kwon, Soonwoo; Seo, Geon Ho; Hyo Kim, Chang

    2014-06-01

    McCARD is a Monte Carlo (MC) neutron-photon transport simulation code developed exclusively for the neutronics design and analysis of nuclear reactor cores. McCARD is equipped with the hierarchical modeling and scripting functions, the CAD-based geometry processing module, the adjoint-weighted kinetics parameter and source multiplication factor estimation modules as well as the burnup analysis capability for the neutronics design and analysis of both research and power reactor cores. This paper highlights applicability of McCARD for the research reactor core neutronics analysis, as demonstrated for Kyoto University Critical Assembly, HANARO, and YALINA.

  7. Refractory metals fabrication technology as applied to fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    1976-07-01

    Activities are reported in research programs on inspection and fabricion of refractory metals and alloys including those of Mo, Nb, Ta, and V. Progress is summarized in sections on blanking, edge preparation, machining, forming, joining, cleaning, thermal processing, and coating. (JRD)

  8. Preliminary Conceptual Design and Development of Core Technology of Very High Temperature Gas-Cooled Reactor Hydrogen Production

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Jong Hwa; Kang, H. S.; Gil, C. S. and others

    2006-05-15

    For the nuclear hydrogen production system, the VHTR technology and the IS cycle technology are being developed. A comparative evaluation on the block type reactor and the pebble type reactor is performed to decide a proper nuclear hydrogen production reactor. 100MWt prismatic type reactor is tentatively decided and its safety characteristics are roughly investigated. Computation codes of nuclear design, thermo-fluid design, safety-performance analysis are developed and verified. Also, the development of a risk informed design technology is started. Experiments for metallic materials and graphites are carried out for the selection of materials of VHTR components. Diverse materials for process heat exchanger are studied in various corrosive environments. Pyrolytic carbon and SiC coating technology is developed and fuel manufacturing technology is basically established. Computer program is developed to evaluate the performance of coated particle fuels.

  9. Multipurpose epithermal neutron beam on new research station at MARIA research reactor in Swierk-Poland

    Energy Technology Data Exchange (ETDEWEB)

    Gryzinski, M.A.; Maciak, M. [National Centre for Nuclear Research, Andrzeja Soltana 7, 05-400 Otwock-Swierk (Poland)

    2015-07-01

    MARIA reactor is an open-pool research reactor what gives the chance to install uranium fission converter on the periphery of the core. It could be installed far enough not to induce reactivity of the core but close enough to produce high flux of fast neutrons. Special design of the converter is now under construction. It is planned to set the research stand based on such uranium converter in the near future: in 2015 MARIA reactor infrastructure should be ready (preparation started in 2013), in 2016 the neutron beam starts and in 2017 opening the stand for material and biological research or for medical training concerning BNCT. Unused for many years, horizontal channel number H2 at MARIA research rector in Poland, is going to be prepared as a part of unique stand. The characteristics of the neutron beam will be significant advantage of the facility. High flux of neutrons at the level of 2x10{sup 9} cm{sup -2}s{sup -1} will be obtainable by uranium neutron converter located 90 cm far from the reactor core fuel elements (still inside reactor core basket between so called core reflectors). Due to reaction of core neutrons with converter U{sub 3}Si{sub 2} material it will produce high flux of fast neutrons. After conversion neutrons will be collimated and moderated in the channel by special set of filters and moderators. At the end of H2 channel i.e. at the entrance to the research room neutron energy will be in the epithermal energy range with neutron intensity at least at the level required for BNCT (2x10{sup 9} cm{sup -2}s{sup -1}). For other purposes density of the neutron flux could be smaller. The possibility to change type and amount of installed filters/moderators which enables getting different properties of the beam (neutron energy spectrum, neutron-gamma ratio and beam profile and shape) is taken into account. H2 channel is located in separate room which is adjacent to two other empty rooms under the preparation for research laboratories (200 m2). It is

  10. FY2012 Engineering Research & Technology Report

    Energy Technology Data Exchange (ETDEWEB)

    Lane, Monya

    2014-07-22

    This report documents engineering research, development, and technology advancements performed by LLNL during fiscal year 2012 in the following areas: computational engineering, engineering information systems, micro/nano-devices and structures, and measurement technologies.

  11. Core calculations for the upgrading of the IEA-R1 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Santos, Adimir dos; Perrotta, Jose A.; Bastos, Jose Luis F.; Yamaguchi, Mitsuo; Umbehaun, Pedro E. [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil)]. E-mail: asantos@net.ipen.br; perrotta@net.ipen.br; mitsuo@net.ipen.br

    1998-07-01

    The IEA-R1 Research Reactor is a multipurpose reactor. It has been used for basic and applied research in the nuclear area, training and radioisotopes production since 1957. In 1995, the Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP) took the decision to modernize and upgrade the power from 2 to 5 MW and increase the operational cycle. This work presents the design requirements and the calculations effectuated to reach this goal. (author)

  12. Technology Development Program for an Advanced Potassium Rankine Power Conversion System Compatible with Several Space Reactor Designs

    Energy Technology Data Exchange (ETDEWEB)

    Yoder, G.L.

    2005-10-03

    This report documents the work performed during the first phase of the National Aeronautics and Space Administration (NASA), National Research Announcement (NRA) Technology Development Program for an Advanced Potassium Rankine Power Conversion System Compatible with Several Space Reactor Designs. The document includes an optimization of both 100-kW{sub e} and 250-kW{sub e} (at the propulsion unit) Rankine cycle power conversion systems. In order to perform the mass optimization of these systems, several parametric evaluations of different design options were investigated. These options included feed and reheat, vapor superheat levels entering the turbine, three different material types, and multiple heat rejection system designs. The overall masses of these Nb-1%Zr systems are approximately 3100 kg and 6300 kg for the 100- kW{sub e} and 250-kW{sub e} systems, respectively, each with two totally redundant power conversion units, including the mass of the single reactor and shield. Initial conceptual designs for each of the components were developed in order to estimate component masses. In addition, an overall system concept was presented that was designed to fit within the launch envelope of a heavy lift vehicle. A technology development plan is presented in the report that describes the major efforts that are required to reach a technology readiness level of 6. A 10-year development plan was proposed.

  13. Critical technologies research: Opportunities for DOE

    Energy Technology Data Exchange (ETDEWEB)

    1992-12-01

    Recent studies have identified a number of critical technologies that are essential to the nation's defense, economic competitiveness, energy independence, and betterment of public health. The National Critical Technologies Panel (NCTP) has identified the following critical technology areas: Aeronautics and Surface Transportation; Biotechnology and Life Sciences; Energy and Environment; Information and Communications; Manufacturing; and Materials. Sponsored by the Department of Energy's Office of Energy Research (OER), the Critical Technologies Research Workshop was held in May 1992. Approximately 100 scientists, engineers, and managers from the national laboratories, industry, academia, and govemment participated. The objective of the Berkeley Workshop was to advance the role of the DOE multiprogram energy laboratories in critical technologies research by describing, defining, and illustrating research areas, opportunities, resources, and key decisions necessary to achieve national research goals. An agenda was developed that looked at DOE's capabilities and options for research in critical technologies and provided a forum for industry, academia, govemment, and the national laboratories to address: Critical technology research needs; existing research activities and resources; capabilities of the national laboratories; and opportunities for national laboratories, industries, and universities. The Workshop included plenary sessions in which presentations by technology and policy leaders set the context for further inquiry into critical technology issues and research opportunities. Separate sessions then focused on each of the following major areas of technology: Advanced materials; biotechnology and life sciences; energy and environment; information and communication; and manufacturing and transportation.

  14. Critical technologies research: Opportunities for DOE

    Energy Technology Data Exchange (ETDEWEB)

    1992-12-01

    Recent studies have identified a number of critical technologies that are essential to the nation`s defense, economic competitiveness, energy independence, and betterment of public health. The National Critical Technologies Panel (NCTP) has identified the following critical technology areas: Aeronautics and Surface Transportation; Biotechnology and Life Sciences; Energy and Environment; Information and Communications; Manufacturing; and Materials. Sponsored by the Department of Energy`s Office of Energy Research (OER), the Critical Technologies Research Workshop was held in May 1992. Approximately 100 scientists, engineers, and managers from the national laboratories, industry, academia, and govemment participated. The objective of the Berkeley Workshop was to advance the role of the DOE multiprogram energy laboratories in critical technologies research by describing, defining, and illustrating research areas, opportunities, resources, and key decisions necessary to achieve national research goals. An agenda was developed that looked at DOE`s capabilities and options for research in critical technologies and provided a forum for industry, academia, govemment, and the national laboratories to address: Critical technology research needs; existing research activities and resources; capabilities of the national laboratories; and opportunities for national laboratories, industries, and universities. The Workshop included plenary sessions in which presentations by technology and policy leaders set the context for further inquiry into critical technology issues and research opportunities. Separate sessions then focused on each of the following major areas of technology: Advanced materials; biotechnology and life sciences; energy and environment; information and communication; and manufacturing and transportation.

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

    Science.gov (United States)

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

    2017-01-01

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

  16. Advanced Reactor Licensing: Experience with Digital I&C Technology in Evolutionary Plants

    Energy Technology Data Exchange (ETDEWEB)

    Wood, RT

    2004-09-27

    This report presents the findings from a study of experience with digital instrumentation and controls (I&C) technology in evolutionary nuclear power plants. In particular, this study evaluated regulatory approaches employed by the international nuclear power community for licensing advanced l&C systems and identified lessons learned. The report (1) gives an overview of the modern l&C technologies employed at numerous evolutionary nuclear power plants, (2) identifies performance experience derived from those applications, (3) discusses regulatory processes employed and issues that have arisen, (4) captures lessons learned from performance and regulatory experience, (5) suggests anticipated issues that may arise from international near-term deployment of reactor concepts, and (6) offers conclusions and recommendations for potential activities to support advanced reactor licensing in the United States.

  17. Electrically Heated Testing of the Kilowatt Reactor Using Stirling Technology (KRUSTY) Experiment Using a Depleted Uranium Core

    Science.gov (United States)

    Briggs, Maxwell H.; Gibson, Marc A.; Sanzi, James

    2017-01-01

    The Kilopower project aims to develop and demonstrate scalable fission-based power technology for systems capable of delivering 110 kW of electric power with a specific power ranging from 2.5 - 6.5 Wkg. This technology could enable high power science missions or could be used to provide surface power for manned missions to the Moon or Mars. NASA has partnered with the Department of Energys National Nuclear Security Administration, Los Alamos National Labs, and Y-12 National Security Complex to develop and test a prototypic reactor and power system using existing facilities and infrastructure. This technology demonstration, referred to as the Kilowatt Reactor Using Stirling TechnologY (KRUSTY), will undergo nuclear ground testing in the summer of 2017 at the Nevada Test Site. The 1 kWe variation of the Kilopower system was chosen for the KRUSTY demonstration. The concept for the 1 kWe flight system consist of a 4 kWt highly enriched Uranium-Molybdenum reactor operating at 800 degrees Celsius coupled to sodium heat pipes. The heat pipes deliver heat to the hot ends of eight 125 W Stirling convertors producing a net electrical output of 1 kW. Waste heat is rejected using titanium-water heat pipes coupled to carbon composite radiator panels. The KRUSTY test, based on this design, uses a prototypic highly enriched uranium-molybdenum core coupled to prototypic sodium heat pipes. The heat pipes transfer heat to two Advanced Stirling Convertors (ASC-E2s) and six thermal simulators, which simulate the thermal draw of full scale power conversion units. Thermal simulators and Stirling engines are gas cooled. The most recent project milestone was the completion of non-nuclear system level testing using an electrically heated depleted uranium (non-fissioning) reactor core simulator. System level testing at the Glenn Research Center (GRC) has validated performance predictions and has demonstrated system level operation and control in a test configuration that replicates the one

  18. The Role of Technology in SLA Research

    Science.gov (United States)

    Chun, Dorothy M.

    2016-01-01

    In this review article for the 20th Anniversary Issue, I look back at research from the last two decades on the role of computer technology in understanding and facilitating second language acquisition (SLA) and forward to what future research might investigate. To be discussed are both how technology has been used to conduct research on SLA…

  19. The Role of Technology in SLA Research

    Science.gov (United States)

    Chun, Dorothy M.

    2016-01-01

    In this review article for the 20th Anniversary Issue, I look back at research from the last two decades on the role of computer technology in understanding and facilitating second language acquisition (SLA) and forward to what future research might investigate. To be discussed are both how technology has been used to conduct research on SLA…

  20. Two conceptual designs of helical fusion reactor FFHR-d1A based on ITER technologies and challenging ideas

    Science.gov (United States)

    Sagara, A.; Miyazawa, J.; Tamura, H.; Tanaka, T.; Goto, T.; Yanagi, N.; Sakamoto, R.; Masuzaki, S.; Ohtani, H.; The FFHR Design Group

    2017-08-01

    The Fusion Engineering Research Project (FERP) at the National Institute for Fusion Science (NIFS) is conducting conceptual design activities for the LHD-type helical fusion reactor FFHR-d1A. This paper newly defines two design options, ‘basic’ and ‘challenging.’ Conservative technologies, including those that will be demonstrated in ITER, are chosen in the basic option in which two helical coils are made of continuously wound cable-in-conduit superconductors of Nb3Sn strands, the divertor is composed of water-cooled tungsten monoblocks, and the blanket is composed of water-cooled ceramic breeders. In contrast, new ideas that would possibly be beneficial for making the reactor design more attractive are boldly included in the challenging option in which the helical coils are wound by connecting high-temperature REBCO superconductors using mechanical joints, the divertor is composed of a shower of molten tin jets, and the blanket is composed of molten salt FLiNaBe including Ti powers to increase hydrogen solubility. The main targets of the challenging option are early construction and easy maintenance of a large and three-dimensionally complicated helical structure, high thermal efficiency, and, in particular, realistic feasibility of the helical reactor.

  1. Research reactor systems for the stable and efficient supply of RI

    Energy Technology Data Exchange (ETDEWEB)

    Lim, In Choel; Oh, Sooy Oul; Lee, Choong Sung; Jun, Byung Jin [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2009-04-15

    The supply of medical isotopes has relied on the major four research reactors over the world and an unavailability of some of these reactors caused a problem in the stable supply of medical isotopes, especially {sup 9}9{sup M}o. There are several ways to produce {sup 9}9{sup M}o but is is believed that the use of a research reactor is the most efficient way. There are two ways to produce {sup 9}9{sup M}o in a research reactor; they are the separation of {sup 9}9{sup M}o from the fission product and the use of neutron capture reaction of {sup 9}8{sup M}o. For the former, various ways are available depending on the target morphology and the enrichment of uranium in the target. The efficiency of the neutron capture method depends on the available neutron flux, the enrichment of {sup 9}8{sup M}o in the target and the efficiency of the adsorption column. Besides these nuclear engineering aspects, other issues affect the use of the research reactor and they include the following; the on power loading of the target, the methods to reduce the cost for the production of RI in research reactors, the logistics between the producer and the consumer, and the coalition of research reactors. In addition, the producers of RI products or the distributors should become the prosumers in the production of sources. The stable and efficient supply of medical isotopes is believed to depend on all these factors and the future options on the use of a research reactor in Korea for the medical isotope supply should consider these.

  2. Training courses on neutron detection systems on the ISIS research reactor: on-site and through internet training

    Energy Technology Data Exchange (ETDEWEB)

    Lescop, B.; Badeau, G.; Ivanovic, S.; Foulon, F. [National Institute for Nuclear science and Technology French Atomic Energy and Alternative Energies Commission (CEA), Saclay Research Center, 91191 Gif-sur-Yvette (France)

    2015-07-01

    Today, ISIS research reactor is an essential tool for Education and Training programs organized by the National Institute for Nuclear Science and Technology (INSTN) from CEA. In the field of nuclear instrumentation, the INSTN offers both, theoretical courses and training courses on the use of neutron detection systems taking advantage of the ISIS research reactor for the supply of a wide range of neutron fluxes. This paper describes the content of the training carried out on the use of neutron detectors and detection systems, on-site or remote. The ISIS reactor is a 700 kW open core pool type reactor. The facility is very flexible since neutron detectors can be inserted into the core or its vicinity, and be used at different levels of power according to the needs of the course. Neutron fluxes, typically ranging from 1 to 10{sup 12} n/cm{sup 2}.s, can be obtained for the characterisation of the neutron detectors and detection systems. For the monitoring of the neutron density at low level of power, the Instrumentation and Control (I and C) system of the reactor is equipped with two detection systems, named BN1 and BN2. Each way contains a fission chamber, type CFUL01, connected to an electronic system type SIREX.The system works in pulse mode and exhibits two outputs: the counting rate and the doubling time. For the high level of power, the I and C is equipped with two detection systems HN1 and HN2.Each way contain a boron ionization chamber (type CC52) connected to an electronics system type SIREX. The system works in current mode and has two outputs: the current and the doubling time. For each mode, the trainees can observe and measure the signal at the different stages of the electronic system, with an oscilloscope. They can understand the role of each component of the detection system: detector, cable and each electronic block. The limitation of the detection modes and their operating range can be established from the measured signal. The trainees can also

  3. Design and installation of a hot water layer system at the Tehran research reactor

    Directory of Open Access Journals (Sweden)

    Mirmohammadi Sayedeh Leila

    2013-01-01

    Full Text Available A hot water layer system (HWLS is a novel system for reducing radioactivity under research reactor containment. This system is particularly useful in pool-type research reactors or other light water reactors with an open pool surface. The main purpose of a HWLS is to provide more protection for operators and reactor personnel against undesired doses due to the radio- activity of the primary loop. This radioactivity originates mainly from the induced radioactivity contained within the cooling water or probable minute leaks of fuel elements. More importantly, the bothersome radioactivity is progressively proportional to reactor power and, thus, the HWLS is a partial solution for mitigating such problems when power upgrading is planned. Following a series of tests and checks for different parameters, a HWLS has been built and put into operation at the Tehran research reactor in 2009. It underwent a series of comprehensive tests for a period of 6 months. Within this time-frame, it was realized that the HWLS could provide a better protection for reactor personnel against prevailing radiation under containment. The system is especially suitable in cases of abnormality, e. g. the spread of fission products due to fuel failure, because it prevents the mixing of pollutants developed deep in the pool with the upper layer and thus mitigates widespread leakage of radioactivity.

  4. Technology, relationships, and problems: a research synthesis.

    Science.gov (United States)

    Hertlein, Katherine M; Webster, Megan

    2008-10-01

    The advances in technology alter the ways we interact with each other. For some, the use of technology can facilitate a relationship; for others, technology can complicate aspects of a relationship. The purpose of this research synthesis is to summarize current research exploring the ways in which technology impacts relationships negatively. Eight studies were reviewed across the following areas: preoperational definitions, sample, methodology, control of extraneous variables, causal influence, generalizability, validity of statistical findings, and conclusions. Implications for authors, researchers, and therapists working with couples and families struggling with technology issues are discussed.

  5. On the implementation of new technology modules for fusion reactor systems codes

    Energy Technology Data Exchange (ETDEWEB)

    Franza, F., E-mail: fabrizio.franza@kit.edu [Institute of Neutron Physics and Reactor Technology, Karlsruhe Institute of Technology (KIT), Eggenstein-Leopoldshafen, 76344 (Germany); Boccaccini, L.V.; Fisher, U. [Institute of Neutron Physics and Reactor Technology, Karlsruhe Institute of Technology (KIT), Eggenstein-Leopoldshafen, 76344 (Germany); Gade, P.V.; Heller, R. [Institute for Technical Physics, Karlsruhe Institute of Technology (KIT), Eggenstein-Leopoldshafen, 76344 (Germany)

    2015-10-15

    Highlights: • At KIT a new technology modules for systems code are under development. • A new algorithm for the definition of the main reactor's components is defined. • A new blanket model based on 1D neutronics analysis is described. • A new TF coil stress model based on 3D electromagnetic analysis is described. • The models were successfully benchmarked against more detailed models. - Abstract: In the frame of the pre-conceptual design of the next generation fusion power plant (DEMO), systems codes are being used from nearly 20 years. In such computational tools the main reactor components (e.g. plasma, blanket, magnets, etc.) are integrated in a unique computational algorithm and simulated by means of rather simplified mathematical models (e.g. steady state and zero dimensional models). The systems code tries to identify the main design parameters (e.g. major radius, net electrical power, toroidal field) and to make the reactor's requirements and constraints to be simultaneously accomplished. In fusion applications, requirements and constraints can be either of physics or technology kind. Concerning the latest category, at Karlsruhe Institute of Technology a new modelling activity has been recently launched aiming to develop improved models focusing on the main technology areas, such as neutronics, thermal-hydraulics, electromagnetics, structural mechanics, fuel cycle and vacuum systems. These activities started by developing: (1) a geometry model for the definition of poloidal profiles for the main reactors components, (2) a blanket model based on neutronics analyses and (3) a toroidal field coil model based on electromagnetic analysis, firstly focusing on the stresses calculations. The objective of this paper is therefore to give a short outline of these models.

  6. Water chemistry management of research reactor in JAERI

    Energy Technology Data Exchange (ETDEWEB)

    Yoshijima, Tetsuo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-10-01

    The JRR-3M cooling system consists of four systems, namely; (1) primary cooling system, (2) heavy water cooling system, (3) helium system and (4) secondary cooling system. The heavy water is used for reflector and pressurized with helium gas. Water chemistry management of the JRR-3M cooling systems is one of the important subject for the safety operation. The main objects are to prevent the corrosion of cooling system and fuel elements, to suppress the plant radiation build-up and to minimize the generation of radioactive waste. All measured values were within the limits of specifications and JRR-3M reactor was operated with safety in 1996. Spent fuels of JRR-3M reactor are stored in the spent fuel pool. This pool water has been analyzed to prevent corrosion of aluminum cladding of spent fuels. Water chemistry of spent fuel pool water is applied to the prevention of corrosion of aluminum alloys including fuel cladding. The JRR-2 reactor was eternally stopped in December 1996 and is now under decommissioning. The JRR-2 reactor is composed of heavy water tank, fuel guide tube and horizontal experimental hole. These are constructed of aluminum alloy and biological shield and upper shield are constructed of concrete. Three types of corrosion of aluminum alloy were observed in the JRR-2. The Alkaline corrosion of aluminum tube occurred in 1972 because of the mechanical damage of the aluminum fuel guide tube which is used for fuel handling. Modification of the reactor top shield was started in 1974 and completed in 1975. (author)

  7. CO2 conversion by plasma technology: insights from modeling the plasma chemistry and plasma reactor design

    Science.gov (United States)

    Bogaerts, A.; Berthelot, A.; Heijkers, S.; Kolev, St.; Snoeckx, R.; Sun, S.; Trenchev, G.; Van Laer, K.; Wang, W.

    2017-06-01

    In recent years there has been growing interest in the use of plasma technology for CO2 conversion. To improve this application, a good insight into the underlying mechanisms is of great importance. This can be obtained from modeling the detailed plasma chemistry in order to understand the chemical reaction pathways leading to CO2 conversion (either in pure form or mixed with another gas). Moreover, in practice, several plasma reactor types are being investigated for CO2 conversion, so in addition it is essential to be able to model these reactor geometries so that their design can be improved, and the most energy efficient CO2 conversion can be achieved. Modeling the detailed plasma chemistry of CO2 conversion in complex reactors is, however, very time-consuming. This problem can be overcome by using a combination of two different types of model: 0D chemical reaction kinetics models are very suitable for describing the detailed plasma chemistry, while the characteristic features of different reactor geometries can be studied by 2D or 3D fluid models. In the first instance the latter can be developed in argon or helium with a simple chemistry to limit the calculation time; however, the ultimate aim is to implement the more complex CO2 chemistry in these models. In the present paper, examples will be given of both the 0D plasma chemistry models and the 2D and 3D fluid models for the most common plasma reactors used for CO2 conversion in order to emphasize the complementarity of both approaches. Furthermore, based on the modeling insights, the paper discusses the possibilities and limitations of plasma-based CO2 conversion in different types of plasma reactors, as well as what is needed to make further progress in this field.

  8. Production and release rate of (37)Ar from the UT TRIGA Mark-II research reactor.

    Science.gov (United States)

    Johnson, Christine; Biegalski, Steven R; Artnak, Edward J; Moll, Ethan; Haas, Derek A; Lowrey, Justin D; Aalseth, Craig E; Seifert, Allen; Mace, Emily K; Woods, Vincent T; Humble, Paul

    2017-02-01

    Air samples were taken at various locations around The University of Texas at Austin's TRIGA Mark II research reactor and analyzed to determine the concentrations of (37)Ar, (41)Ar, and (133)Xe present. The measured ratio of (37)Ar/(41)Ar and historical records of (41)Ar releases were then utilized to estimate an annual average release rate of (37)Ar from the reactor facility. Using the calculated release rate, atmospheric transport modeling was performed in order to determine the potential impact of research reactor operations on nearby treaty verification activities. Results suggest that small research reactors (∼1 MWt) do not release (37)Ar in concentrations measurable by currently proposed OSI detection equipment.

  9. ITHNA.SYS: An Integrated Thermal Hydraulic and Neutronic Analyzer SYStem for NUR research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mazidi, S., E-mail: samirmazidi@gmail.com [Division Physique et Applications Nucléaires, Centre de Recherche Nucléaire de Draria (CRND), BP 43 Sebala, Draria, Alger (Algeria); Meftah, B., E-mail: b_meftah@yahoo.com [Division Physique et Applications Nucléaires, Centre de Recherche Nucléaire de Draria (CRND), BP 43 Sebala, Draria, Alger (Algeria); Belgaid, M., E-mail: belgaidm@yahoo.com [Faculté de Physique, Université Houari Boumediene, USTHB, BP 31, Bab Ezzouar, Alger (Algeria); Letaim, F., E-mail: fletaim@yahoo.fr [Faculté des Sciences et Technologies, Université d’El-oued, PO Box 789, El-oued (Algeria); Halilou, A., E-mail: hal_rane@yahoo.fr [Division Réacteur NUR, Centre de Recherche Nucléaire de Draria, BP 43 Sebala, Draria, Alger (Algeria)

    2015-08-15

    Highlights: • We develop a neutronic and thermal hydraulic MTR reactor analyzer. • The analyzer allows a rapid determination of the reactor core parameters. • Some NUR reactor parameters have been analyzed. - Abstract: This paper introduces the Integrated Thermal Hydraulic and Neutronic Analyzer SYStem (ITHNA.SYS) that has been developed for the Algerian research reactor NUR. It is used both as an operating aid tool and as a core physics engineering analysis tool. The system embeds three modules of the MTR-PC software package developed by INVAP SE: the cell calculation code WIMSD, the core calculation code CITVAP and the program TERMIC for thermal hydraulic analysis of a material testing reactor (MTR) core in forced convection. ITHNA.SYS operates both in on-line and off-line modes. In the on-line mode, the system is linked, via the computer parallel port, to the data acquisition console of the reactor control room and allows a real time monitoring of major physical and safety parameters of the NUR core. PC-based ITHNA.SYS provides a viable and convenient way of using an accumulated and often complex reactor physics stock of knowledge and frees the user from the intricacy of adequate reactor core modeling. This guaranties an accurate, though rapid, determination of a variety of neutronic and thermal hydraulic parameters of importance for the operation and safety analysis of the NUR research reactor. Instead of the several hours usually required, the processing time for the determination of such parameters is now reduced to few seconds. Validation of the system was performed with respect to experimental measurements and to calculations using reference codes. ITHNA.SYS can be easily adapted to accommodate other kinds of MTR reactors.

  10. Bioactive Foamed Emulsion Reactor: A New Technology for Biotreatment of Airborne Volatile Organic Compound

    Directory of Open Access Journals (Sweden)

    F. Ghorbani Shahna

    2010-04-01

    Full Text Available Introduction & Objective: Biological treatment is a new established technology for the air pollutants. This technology can be an alternative for physical and chemical treatment methods. Among bioreators, the Bioactive Foamed Emulsion Reactor (BFER is a new alternative that has not the problems of the conventional ones. In this reactor bed clogging in the conventional bioreactor was resolved by bioactive foam as a substitute of packing bed. The pollutant absorption has been increased using biocompatible organic phase in liquid .This reactor can be used for higher inlet toluene concentration. The objective of this study was designing and optimizing the operational parameters of BFER for toluene treatment.Materials & Methods: In the first step of this experimental-analytic study, the toluene degradation microorganisms were identified, extracted and concentrated for injection to bioreactor. Then the effect of several parameters such as Kind and concentration of organic phase,and residence time oxygen content on bioreactor performance were studied and the optimum conditions were selected for continuous operation. The continuous operation of bioreactor was monitored at the optimum conditions.Results: Experimental results showed that the residence time of 15s, oxygen content of 40%, and the 4 % (v/v n-hexadecane as organic phase were the optimum conditions. The average elimination capacity (EC and removal efficiency of bioreactor were 231.68 g/m3h and 88.44% respectively for the inlet concentration about of 1 g/m3. The statistical developed model predicted that the maximum EC of this reactor could reach to 426.21 g/m3h.Conclusion: Since the elimination capacity of this reactor is several times more than the other bioreactors, it has the potential to be applied instead of biofilters and biotrickling filters.

  11. Operating Modes Of Chemical Reactors Of Polymerization

    Directory of Open Access Journals (Sweden)

    Meruyert Berdieva

    2012-05-01

    Full Text Available In the work the issues of stable technological modes of operation of main devices of producing polysterol reactors have been researched as well as modes of stable operation of a chemical reactor have been presented, which enables to create optimum mode parameters of polymerization process, to prevent emergency situations of chemical reactor operation in industrial conditions.

  12. Building Technologies Research and Integration Center (BTRIC)

    Data.gov (United States)

    Federal Laboratory Consortium — The Building Technologies Research and Integration Center (BTRIC), in the Energy and Transportation Science Division (ETSD) of Oak Ridge National Laboratory (ORNL),...

  13. Strategic Plan for Light Water Reactor Research and Development

    Energy Technology Data Exchange (ETDEWEB)

    None

    2004-02-01

    The purpose of this strategic plan is to establish a framework that will allow the Department of Energy (DOE) and the nuclear power industry to jointly plan the nuclear energy research and development (R&D) agenda important to achieving the Nation's energy goals. This strategic plan has been developed to focus on only those R&D areas that will benefit from a coordinated government/industry effort. Specifically, this plan focuses on safely sustaining and expanding the electricity output from currently operating nuclear power plants and expanding nuclear capacity through the deployment of new plants. By focusing on R&D that addresses the needs of both current and future nuclear plants, DOE and industry will be able to take advantage of the synergism between these two technology areas, thus improving coordination, enhancing efficiency, and further leveraging public and private sector resources. By working together under the framework of this strategic plan, DOE and the nuclear industry reinforce their joint commitment to the future use of nuclear power and the National Energy Policy's goal of expanding its use in the United States. The undersigned believe that a public-private partnership approach is the most efficient and effective way to develop and transfer new technologies to the marketplace to achieve this goal. This Strategic Plan is intended to be a living document that will be updated annually.

  14. Kilowatt Reactor Using Stirling TechnologY (KRUSTY) Demonstration. CEDT Phase 1 Preliminary Design Documentation

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez, Rene Gerardo [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Hutchinson, Jesson D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Mcclure, Patrick Ray [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Myers, William L. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-08-20

    The intent of the integral experiment request IER 299 (called KiloPower by NASA) is to assemble and evaluate the operational performance of a compact reactor configuration that closely resembles the flight unit to be used by NASA to execute a deep space exploration mission. The reactor design will include heat pipes coupled to Stirling engines to demonstrate how one can generate electricity when extracting energy from a “nuclear generated” heat source. This series of experiments is a larger scale follow up to the DUFF series of experiments1,2 that were performed using the Flat-Top assembly.

  15. Quantum technology: from research to application

    Science.gov (United States)

    Schleich, Wolfgang P.; Ranade, Kedar S.; Anton, Christian; Arndt, Markus; Aspelmeyer, Markus; Bayer, Manfred; Berg, Gunnar; Calarco, Tommaso; Fuchs, Harald; Giacobino, Elisabeth; Grassl, Markus; Hänggi, Peter; Heckl, Wolfgang M.; Hertel, Ingolf-Volker; Huelga, Susana; Jelezko, Fedor; Keimer, Bernhard; Kotthaus, Jörg P.; Leuchs, Gerd; Lütkenhaus, Norbert; Maurer, Ueli; Pfau, Tilman; Plenio, Martin B.; Rasel, Ernst Maria; Renn, Ortwin; Silberhorn, Christine; Schiedmayer, Jörg; Schmitt-Landsiedel, Doris; Schönhammer, Kurt; Ustinov, Alexey; Walther, Philip; Weinfurter, Harald; Welzl, Emo; Wiesendanger, Roland; Wolf, Stefan; Zeilinger, Anton; Zoller, Peter

    2016-05-01

    The term quantum physics refers to the phenomena and characteristics of atomic and subatomic systems which cannot be explained by classical physics. Quantum physics has had a long tradition in Germany, going back nearly 100 years. Quantum physics is the foundation of many modern technologies. The first generation of quantum technology provides the basis for key areas such as semiconductor and laser technology. The "new" quantum technology, based on influencing individual quantum systems, has been the subject of research for about the last 20 years. Quantum technology has great economic potential due to its extensive research programs conducted in specialized quantum technology centres throughout the world. To be a viable and active participant in the economic potential of this field, the research infrastructure in Germany should be improved to facilitate more investigations in quantum technology research.

  16. Report of the second joint Research Committee for Fusion Reactor and Materials. July 12, 2002, Tokyo, Japan

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2003-05-01

    Joint research committees in purpose of the discussion on DEMO blanket in view point of the both of reactor technology and materials were held by the Research Committee for Fusion Reactor and Fusion Materials. The joint research committee was held in Tokyo on July 12, 2002. In the committee, the present status of development of solid and liquid breeding blanket, the present status of development of reduced activation structure materials, and IFMIF (International Fusion Materials Irradiation Facility) program were discussed based on the discussions of the development programs of the blanket and materials at the first joint research committee. As a result, it was confirmed that high electric efficiency with 41% would be obtained in the solid breeding blanket system, that neutron radiation data of reduced activation ferritic steel was obtained by HFIR collaboration, and that KEP (key element technology phase) of IFMIF would be finished at the end of 2002 and the data base for the next step, i.e. EVEDA (engineering validation/engineering design activity) was obtained. In addition, the present status of ITER CTA, which was a transient phase for the construction, and the outline of ITER Fast Track, which was an accelerated plan for the performance of the power plants, were reported. This report consists of the summary of the discussion and the viewgraphs which were used at the second joint research committee, and these are very useful for the researchers of the fusion area in Japan. (author)

  17. Experimental facility for development of high-temperature reactor technology: instrumentation needs and challenges

    OpenAIRE

    Sabharwall Piyush; O’Brien James E.; Yoon SuJong; Sun Xiaodong

    2015-01-01

    A high-temperature, multi-fluid, multi-loop test facility is under development at the Idaho National Laboratory for support of thermal hydraulic materials, and system integration research for high-temperature reactors. The experimental facility includes a high-temperature helium loop, a liquid salt loop, and a hot water/steam loop. The three loops will be thermally coupled through an intermediate heat exchanger (IHX) and a secondary heat exchanger (SHX). Research topics to be addressed includ...

  18. Analysis on Electromagnetic Characteristics of Research Reactor Control Rod Drive Mechanism for Thrust Force Improvement

    Energy Technology Data Exchange (ETDEWEB)

    Huh, Hyung; Choi, Myoung Hwan; Yu, Je Yong; Cho, Yeong Garp; Kim, Jong In [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    The control rod drive mechanism (CRDM) is the part of reactor regulating system (RRS), which is located in the reactor pool top or the room below the reactor pool. The function of the CRDM is to insert, withdraw or maintain neutron absorbing material (control rod) at any required position within the reactor core, in order to the reactivity of the core. There are so many kinds of CRDM, such as magnetic-jack type, hydraulic type, rack and pinion type, chain type and linear or rotary step motor and so on. As a part of a new project, we are investigating the movable coil electromagnetic drive mechanism (MCEDM) which is new scheme for the reactor control rod adopted by China Advanced Research Reactor (CARR). To have a better knowledge of the electromagnetic and magnetic characteristics, numerical models of MCEDM are proposed. Especially in order to achieve improved thrust force, numerical magnetic field calculations for various kinds of magnetic and electromagnetic configuration have been performed. As a result, we present the improved design of MCEDM for research reactor

  19. DISMANTLING OF THE UPPER RPV COMPONENTS OF THE KARLSRUHE MULTI-PURPOSE RESEARCH REACTOR (MZFR), GERMANY

    Energy Technology Data Exchange (ETDEWEB)

    Prechtl, E.; Suessdorf, W.

    2003-02-27

    The Multi-purpose Research Reactor was a pressurized-water reactor cooled and moderated with heavy water. It was built from 1961 to 1966 and went critical for the first time on 29 September 1965. After nineteen years of successful operation, the reactor was de-activated on 3 May 1984. The reactor had a thermal output of 200 MW and an electrical output of 50 MW. The MZFR not only served to supply electrical power, but also as a test bed for: - research into various materials for reactor building (e. g. zirkaloy), - the manufacturing and operating industry to gain experience in erection and operation, - training scientific and technical reactor staff, and - power supply (first nuclear combined-heat-and-power system, 1979-1984). The experience gained in operating the MZFR was very helpful for the development and operation of power reactors. At first, safe containment and enclosure of the plant was planned, but then it was decided to dismantle the plant completely, step by step, in view o f the clear advantages of this approach. The decommissioning concept for the complete elimination of the plant down to a green-field site provides for eight steps. A separate decommissioning license is required for each step. As part of the dismantling, about 72,000 Mg [metric tons] of concrete and 7,200 Mg of metal (400 Mg RPV) must be removed. About 700 Mg of concrete (500 Mg biological shield) and 1300 Mg of metal must be classified as radioactive waste.

  20. Development of essential system technologies for advanced reactor - Development of natural circulation analysis code for integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Goon Cherl; Park, Ik Gyu; Kim, Jae Hak; Lee, Sang Min; Kim, Tae Wan [Seoul National University, Seoul (Korea)

    1999-04-01

    The objective of this study is to understand the natural circulation characteristics of integral type reactors and to develope the natural circulation analysis code for integral type reactors. This study is focused on the asymmetric 3-dimensional flow during natural circulation such as 1/4 steam generator section isolation and the inclination of the reactor systems. Natural circulation experiments were done using small-scale facilities of integral reactor SMART (System-Integrated Modular Advanced ReacTor). CFX4 code was used to investigate the flow patterns and thermal mixing phenomena in upper pressure header and downcomer. Differences between normal operation of all steam generators and the 1/4 section isolation conditions were observed and the results were used as the data 1/4 section isolation conditions were observed and the results were used as the data for RETRAN-03/INT code validation. RETRAN-03 code was modified for the development of natural circulation analysis code for integral type reactors, which was development of natural circulation analysis code for integral type reactors, which was named as RETRAN-03/INT. 3-dimensional analysis models for asymmetric flow in integral type reactors were developed using vector momentum equations in RETRAN-03. Analysis results using RETRAN-03/INT were compared with experimental and CFX4 analysis results and showed good agreements. The natural circulation characteristics obtained in this study will provide the important and fundamental design features for the future small and medium integral reactors. (author). 29 refs., 75 figs., 18 tabs.

  1. Development of essential system technologies for advanced reactor - Development of natural circulation analysis code for integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Goon Cherl; Park, Ik Gyu; Kim, Jae Hak; Lee, Sang Min; Kim, Tae Wan [Seoul National University, Seoul (Korea)

    1999-04-01

    The objective of this study is to understand the natural circulation characteristics of integral type reactors and to develope the natural circulation analysis code for integral type reactors. This study is focused on the asymmetric 3-dimensional flow during natural circulation such as 1/4 steam generator section isolation and the inclination of the reactor systems. Natural circulation experiments were done using small-scale facilities of integral reactor SMART (System-Integrated Modular Advanced ReacTor). CFX4 code was used to investigate the flow patterns and thermal mixing phenomena in upper pressure header and downcomer. Differences between normal operation of all steam generators and the 1/4 section isolation conditions were observed and the results were used as the data 1/4 section isolation conditions were observed and the results were used as the data for RETRAN-03/INT code validation. RETRAN-03 code was modified for the development of natural circulation analysis code for integral type reactors, which was development of natural circulation analysis code for integral type reactors, which was named as RETRAN-03/INT. 3-dimensional analysis models for asymmetric flow in integral type reactors were developed using vector momentum equations in RETRAN-03. Analysis results using RETRAN-03/INT were compared with experimental and CFX4 analysis results and showed good agreements. The natural circulation characteristics obtained in this study will provide the important and fundamental design features for the future small and medium integral reactors. (author). 29 refs., 75 figs., 18 tabs.

  2. Research needs of the new accelerator technologies

    Energy Technology Data Exchange (ETDEWEB)

    Sessler, A.M.

    1982-08-01

    A review is given of some of the new accelerator technologies with a special eye to the requirements which they generate for research and development. Some remarks are made concerning the organizational needs of accelerator research.

  3. Control rod reactivity worth determination of a typical MTR research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rizwan, M.; Raza, S.S.; Khan, R. [Pakistan Institute of Engineering and Applied Sciences (PIEAS), Islamabad (Pakistan). Dept. of Nuclear Engineering

    2015-10-15

    The safe and reliable utilization of research reactor demands the possible accurate information of control rod (CR) worths. The criticality positions of the control rods changes with time due to build up fission products. It is therefore important to determine the reactivity worth of control rods. The aim of this article is to estimate the reactivity worth of controls rods in the equilibrium core of a Materials Testing Reactor (MTR). A deterministic model of the reactor core was developed and confirmed against the reference results of excess reactivity, shutdown margin and combined control rod reactivity worth using the combination of WIMS/D4 and CITATION computer codes.

  4. Fuel-coolant interaction (FCI) phenomena in reactor safety. Current understanding and future research needs

    Energy Technology Data Exchange (ETDEWEB)

    Speis, T.P. [Maryland Univ., College Park, MD (United States); Basu, S.

    1998-01-01

    This paper gives an account of the current understanding of fuel-coolant interaction (FCI) phenomena in the context of reactor safety. With increased emphasis on accident management and with emerging in-vessel core melt retention strategies for advanced light water reactor (ALWR) designs, recent interest in FCI has broadened to include an evaluation of potential threats to the integrity of reactor vessel lower head and ex-vessel structural support, as well as the role of FCI in debris quenching and coolability. The current understanding of FCI with regard to these issues is discussed, and future research needs to address the issues from a risk perspective are identified. (author)

  5. Fuel burnup calculation of a research reactor plate element

    Energy Technology Data Exchange (ETDEWEB)

    Santos, Nadia Rodrigues dos; Lima, Zelmo Rodrigues de; Moreira, Maria de Lourdes, E-mail: nadiasam@gmail.com, E-mail: zrlima@ien.gov.br, E-mail: malu@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2013-07-01

    This work consists in simulating the burnup of two different plate type fuel elements, where one is the benchmark MTR of the IAEA, which is made of an alloy of uranium and aluminum, while the other belonging to a typical multipurpose reactor is composed of an alloy of uranium and silicon. The simulation is performed using the WIMSD-5B computer code, which makes use of deterministic methods for solving neutron transport. In developing this task, fuel element equivalent cells were calculated representing each of the reactors to obtain the initial concentrations of each isotope constituent element of the fuel cell and the thicknesses corresponding to each region of the cell, since this information is part of the input data. The compared values of the k∞ showed a similar behavior for the case of the MTR calculated with the WIMSD-5B and EPRI-CELL codes. Relating the graphs of the concentrations in the burnup of both reactors, there are aspects very similar to each isotope selected. The application WIMSD-5B code to calculate isotopic concentrations and burnup of the fuel element, proved to be satisfactory for the fulfillment of the objective of this work. (author)

  6. Event management in research reactors; Gestion de eventos en reactores de investigacion

    Energy Technology Data Exchange (ETDEWEB)

    Perrin, C.D. [Coordinador Reactores de Investigacion y Conjuntos Criticos, Autoridad Regulatoria Nuclear (Argentina)]. e-mail: cperrin@sede.arn.gov.ar

    2006-07-01

    In the Radiological and Nuclear Safety field, the Nuclear Regulatory Authority of Argentina controls the activities of three investigation reactors and three critical groups, by means of evaluations, audits and inspections, in order to assure the execution of the requirements settled down in the Licenses of the facilities, in the regulatory standards and in the documentation of mandatory character in general. In this work one of the key strategies developed by the ARN to promote an appropriate level of radiological and nuclear safety, based on the control of the administration of the abnormal events that its could happen in the facilities is described. The established specific regulatory requirements in this respect and the activities developed in the entities operators are presented. (Author)

  7. MITR-III: Upgrade and relicensing studies for the MIT Research Reactor. Second annual report

    Energy Technology Data Exchange (ETDEWEB)

    Trosman, H.G. [ed.; Lanning, D.D.; Harling, O.K.

    1994-08-01

    The current operating license of the MIT research reactor will expire on May 7, 1996 or possibly a few years later if the US Nuclear Regulatory Commission agrees that the license period can start with the date of initial reactor operation. Driven by the imminent expiration of the operating license, a team of nuclear engineering staff and students have begun a study of the future options for the MIT Research Reactor. These options have included the range from a major rebuilding of the reactor to its decommissioning. This document reports the results of a two year intensive activity which has been supported by a $148,000 grant from the USDOE contract Number DEFG0293ER75859, approximately $100,000 of internal MIT funds and Nuclear Engineering Department graduate student fellowships as well as assistance from international visiting scientists and engineers.

  8. Photon spectrum behind biological shielding of the LVR-15 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Klupak, V.; Viererbl, L.; Lahodova, Z.; Marek, M.; Vins, M. [Research Centre Rez Ltd., Husinec-Rez 130 (Czech Republic)

    2011-07-01

    The LVR-15 reactor is a light water research reactor situated at the Research Centre Rez, near Prague. It operates as a multipurpose facility with a maximum thermal power of 10 MW. The reactor core usually contains from 28 to 32 fuel assemblies with a total mass of {sup 235}U of about 5 kg. Emitted radiation from the fuel caused by fission is shielded by moderating water, a steel reactor vessel, and heavy concrete. This paper deals with measurement and analysis of the gamma spectrum near the outer surface of the concrete wall, behind biological shielding, mainly in the 3- to 10-MeV energy range. A portable HPGe detector with a portable multichannel analyzer was used to measure gamma spectra. The origin of energy lines in gamma detector spectra was identified. (authors)

  9. 基于SOPC的反应堆棒位信息监测技术研究%Research on Reactor Control Rod Position Indication Information Monitoring Technology Based on SOPC

    Institute of Scientific and Technical Information of China (English)

    郑晓; 蔡晨; 孙宇; 刘明星

    2013-01-01

    A new monitoring technology for control rod position is developed by utilizing the FPGA (Field-Programmable Gate Array) platform and using the SOPC (System On Programmable Chip) technology.In this SOPC system,Nios Ⅱ CPU,VGA (Video Graphics Array) display controller and CAN (Controller Area Network) bus controller are integrated in one FPGA chip.Thus the SOPC hardware platform with comprehensive functionality is constructed.Based on this SOPC platform,the real time data of control rod position indication system can be vividly displayed on the LCD and stored in the external nonvolatile RAM as history records.Therefore,the operator can obtain the overall operation status of control rod position indication system quickly and conveniently.The developed prototype has proved the feasibility of this technology.%运用可编程片上系统(SOPC)技术,以现场可编程门阵列(FPGA)为数字平台,研究一种反应堆棒位信息监测技术.该监测技术将Nios Ⅱ处理器、视频图形阵列(VGA)显示控制器、控制器局域网络(CAN)总线控制器等集成在一片FPGA芯片中,构建完成一个具有丰富功能的SOPC硬件平台.采用该硬件平台,可将棒位系统的运行状态信息实时、直观地进行数字化显示,同时以日志的形式存储在外部存储设备中,能够全方位地监测棒位系统的运行状态,使操纵人员能更方便、快捷地掌握整个棒位系统的运行情况.研制完成的原理样机验证了该技术的可行性.

  10. Instrumentation, Controls, and Human-Machine Interface Technology Development Roadmap in Support of Grid Appropriate Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Holcomb, David Eugene [ORNL; Upadhyaya, Belle R. [University of Tennessee, Knoxville (UTK); Kisner, Roger A [ORNL; O' Hara, John [Brookhaven National Laboratory (BNL); Quinn, Edward L. [Longenecker & Associates; Miller, Don W. [Ohio State University

    2009-01-01

    Grid Appropriate Reactors (GARs) are a component of the U.S. Department of Energy s (DOE s) Global Nuclear Energy Partnership (GNEP) program. GARs have smaller output power (<~600 MWe), than those intended for deployment on large, tightly coupled grids. This smaller size is important in avoiding grid destabilization, which can result from having a large fraction of a grid s electrical generation supplied by a single source. GARs are envisioned to be deployed worldwide often in locations without extensive nuclear power experience. DOE recently sponsored the creation of an Instrumentation, Controls, and Human-Machine Interface (ICHMI) technology development roadmap emphasizing the specific characteristics of GARs [1]. This roadmapping effort builds upon and focuses the recently developed, more general nuclear energy ICHMI technology development roadmap [2]. The combination of the smaller plant size, smaller grids, and deployment in locations without extensive prior nuclear power experience presents particular infrastructure, regulation, design, operational, and safeguards challenges for effective GAR deployment. ICHMI technologies are central to efficient GAR operation and as such are a dimension of each of these challenges. Further, while the particular ICHMI technologies to be developed would be useful at larger power plants, they are not high-priority development items at the larger plants. For example, grid transient resilience would be a useful feature for any reactor/grid combination and indeed would have limited some recent blackout events. However, most large reactors have limited passive cooling features. Large plants with active safety response features will likely preserve trip preferential grid transient response. This contrasts sharply with GARs featuring passive shutdown cooling, which can safely support grid stability during large grid transients. ICHMI technologies ranging from alternative control algorithms to simplified human-interface system

  11. Development of technology-neutral safety requirements for the regulation of future nuclear power reactors: Back to basics

    Energy Technology Data Exchange (ETDEWEB)

    Tronea, Madalina, E-mail: madalina.tronea@gmail.co [Faculty of Physics, University of Bucharest (Romania)

    2011-03-15

    This paper explores the current trends as regards the development of technology-neutral safety requirements to be used in the regulation of future nuclear power reactors and the role of the quantitative safety goals in the design of reactor safety systems. The use of the recommendations of the International Commission on Radiological Protection (ICRP) on protection against potential exposure could form the basis of a technology-neutral framework for safety requirements on new reactor designs and could contribute to international harmonisation of nuclear safety assessment practices as part of the licensing processes for future nuclear power plants.

  12. Radiation protection tasks on the Kiev research reactor WWR-M

    Directory of Open Access Journals (Sweden)

    Lobach Yuri N.

    2009-01-01

    Full Text Available Both the description of and the operational experience with the radiation protection system at the research reactor WWR-M are presented. The list of the factors regarding the radiation hazards during the reactor routine operation is given and the main activities on the radiation safety provision are established. The statistical information for the staff exposure, the radioactive aerosol releases and the external radiation monitoring is shown. The preliminary considerations on the system upgrading for the decommissioning are presented.

  13. Oil heat technology research and development

    Energy Technology Data Exchange (ETDEWEB)

    Kweller, E.R. [Department of Energy, Washington, DC (United States); McDonald, R.J. [Brookhaven National Lab., Upton, NY (United States)

    1995-04-01

    The purpose of this United States Department of Energy (DOE)/Brookhaven National Laboratory (BNL) program is to develop a technology base for advancing the state-of-the-art related to oilfired combustion equipment. The major thrust is through technology based research that will seek new knowledge leading to improved designs and equipment optimization. The Combustion Equipment space Conditioning Technology program currently deals exclusively with residential and small commercial building oil heat technology.

  14. Safety Issues at the DOE Test and Research Reactors. A Report to the U.S. Department of Energy.

    Science.gov (United States)

    National Academy of Sciences - National Research Council, Washington, DC. Commission on Physical Sciences, Mathematics, and Resources.

    This report provides an assessment of safety issues at the Department of Energy (DOE) test and research reactors. Part A identifies six safety issues of the reactors. These issues include the safety design philosophy, the conduct of safety reviews, the performance of probabilistic risk assessments, the reliance on reactor operators, the fragmented…

  15. Disposal of irradiated fuel elements from German research reactors. Status and outlook

    Energy Technology Data Exchange (ETDEWEB)

    Thamm, G. [Central Research Reactor and Nuclear Operations Division, Research Centre Juelich, Forschungszentrum Juelich GmbH, Juelich (Germany)

    1999-07-01

    There will be a quantity of highly radioactive spent nuclear fuel (snf) from German research reactors amounting to about 9.1 t by the end of the next decade, which has to be disposed of. About 4.1 t of this quantity are intended to be returned to the USA. The remaining approximately 5 t can be loaded into approximately 30 CASTOR-2 casks and will be stored in a central German dry interim store for about 30 to 50 years (first step of the domestic disposal concept). Of course, snf arising from the operation of research reactors beyond 2010 has to be disposed of in the same way (3 MTR-2 casks every two years for BER-II and FRM-II). It is expected that snf from the zero-power facilities probably will be recycled for reusing the uranium. Due to the amendment of the German Atomic Energy Act intended by the new Federal German Government, the interim dry storage of snf from power reactors in central storage facilities like Ahaus or Gorleben will be stopped and the power reactors have to store snf at their own sites. Although the amendment only concerns nuclear power reactors, it could not be excluded that snf from research reactors, too, cannot be stored at Ahaus or Gorleben at present. (author)

  16. Using Videotape Technology: Innovations in Behavioral Research.

    Science.gov (United States)

    Niebuhr, Robert E.; And Others

    1981-01-01

    Reviews present and potential uses of videotape technology in behavioral research. Emphasis is placed on research methodology and the value of incorporating videotape in current research practices, including behavior observation studies, in modeling research, and in perceptual attribution investigations. Concludes with an analysis of advantages…

  17. Korea-China Optical Technology Research Centre

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Cheol Jung; Cha, H. K.; Rhee, Y. J. (and others)

    2007-04-15

    The main objectives of this project are to develop cooperative channel by personnel exchanges between industrial, educational and research partners of Korea and China on the fields of optical technologies which are the basis of optical industry and being spot-lighted as new industry of 21th century, and to raise the class of Korean optical technology up to world class by utilization of Chinese large facilities through the cooperative research between the optical technology institutions of both sides. To attain the goals mentioned above, we carried out the cooperative researches between the Korean and Chinese optical technology institutions in the following 7 fields; ? research cooperation between KAERI-SITP for the quantum structured far-IR sensor technology - research cooperation for the generation of femtosecond nuclear fusion induced neutrons - research cooperation between KAERI-AIOFM for laser environment analysis and remote sensing technology - research cooperation between KAERI-SIOM for advanced diode-pumped laser technology - cooperative research related on linear and nonlinear magneto-optical properties of advanced magnetic quantum structures - design of pico-second PW high power laser system and its simulation and - cooperative research related on the femto-second laser-plasma interaction physics.

  18. High Efficiency Nuclear Power Plants Using Liquid Fluoride Thorium Reactor Technology

    Science.gov (United States)

    Juhasz, Albert J.; Rarick, Richard A.; Rangarajan, Rajmohan

    2009-01-01

    An overall system analysis approach is used to propose potential conceptual designs of advanced terrestrial nuclear power plants based on Oak Ridge National Laboratory (ORNL) Molten Salt Reactor (MSR) experience and utilizing Closed Cycle Gas Turbine (CCGT) thermal-to-electric energy conversion technology. In particular conceptual designs for an advanced 1 GWe power plant with turbine reheat and compressor intercooling at a 950 K turbine inlet temperature (TIT), as well as near term 100 MWe demonstration plants with TITs of 950 and 1200 K are presented. Power plant performance data were obtained for TITs ranging from 650 to 1300 K by use of a Closed Brayton Cycle (CBC) systems code which considered the interaction between major sub-systems, including the Liquid Fluoride Thorium Reactor (LFTR), heat source and heat sink heat exchangers, turbo-generator machinery, and an electric power generation and transmission system. Optional off-shore submarine installation of the power plant is a major consideration.

  19. High temperature gas-cooled reactor (HTGR) graphite pebble fuel: Review of technologies for reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Mcwilliams, A. J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-09-08

    This report reviews literature on reprocessing high temperature gas-cooled reactor graphite fuel components. A basic review of the various fuel components used in the pebble bed type reactors is provided along with a survey of synthesis methods for the fabrication of the fuel components. Several disposal options are considered for the graphite pebble fuel elements including the storage of intact pebbles, volume reduction by separating the graphite from fuel kernels, and complete processing of the pebbles for waste storage. Existing methods for graphite removal are presented and generally consist of mechanical separation techniques such as crushing and grinding chemical techniques through the use of acid digestion and oxidation. Potential methods for reprocessing the graphite pebbles include improvements to existing methods and novel technologies that have not previously been investigated for nuclear graphite waste applications. The best overall method will be dependent on the desired final waste form and needs to factor in the technical efficiency, political concerns, cost, and implementation.

  20. Air leakage analysis of research reactor HANARO building in typhoon condition for the nuclear emergency preparedness

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Goany Up; Lee, Hae Cho; Kim, Bong Seok; Kim, Jong Soo; Choi, Pyung Kyu [Dept. of Emergency Preparedness, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-12-15

    To find out the leak characteristic of research reactor 'HANARO' building in a typhoon condition MELCOR code which normally is used to simulate severe accident behavior in a nuclear power plant was used to simulate the leak rate of air and fission products from reactor hall after the shutdown of the ventilation system of HANARO reactor building. For the simulation, HANARO building was designed by MELCOR code and typhoon condition passed through Daejeon in 2012 was applied. It was found that the leak rate is 0.1%·day{sup -1} of air, 0.004%·day{sup -1} of noble gas and 3.7×10{sup -5}%·day{sup -1} of aerosol during typhoon passing. The air leak rate of 0.1%·day can be converted into 1.36 m{sup 3}·hr{sup -1} , but the design leak rate in HANARO safety analysis report was considered as 600 m3·hr{sup -1} under the condition of 20 m·sec{sup -1} wind speed outside of the building by typhoon. Most of fission products during the maximum hypothesis accident at HANARO reactor will be contained in the reactor hall, so the direct radiation by remained fission products in the reactor hall will be the most important factor in designing emergency preparedness for HANARO reactor.

  1. Regulatory Technology Development Plan Sodium Fast Reactor. Mechanistic Source Term Development

    Energy Technology Data Exchange (ETDEWEB)

    Grabaskas, David S. [Argonne National Lab. (ANL), Argonne, IL (United States); Brunett, Acacia Joann [Argonne National Lab. (ANL), Argonne, IL (United States); Bucknor, Matthew D. [Argonne National Lab. (ANL), Argonne, IL (United States); Sienicki, James J. [Argonne National Lab. (ANL), Argonne, IL (United States); Sofu, Tanju [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-02-28

    Construction and operation of a nuclear power installation in the U.S. requires licensing by the U.S. Nuclear Regulatory Commission (NRC). A vital part of this licensing process and integrated safety assessment entails the analysis of a source term (or source terms) that represents the release of radionuclides during normal operation and accident sequences. Historically, nuclear plant source term analyses have utilized deterministic, bounding assessments of the radionuclides released to the environment. Significant advancements in technical capabilities and the knowledge state have enabled the development of more realistic analyses such that a mechanistic source term (MST) assessment is now expected to be a requirement of advanced reactor licensing. This report focuses on the state of development of an MST for a sodium fast reactor (SFR), with the intent of aiding in the process of MST definition by qualitatively identifying and characterizing the major sources and transport processes of radionuclides. Due to common design characteristics among current U.S. SFR vendor designs, a metal-fuel, pool-type SFR has been selected as the reference design for this work, with all phenomenological discussions geared toward this specific reactor configuration. This works also aims to identify the key gaps and uncertainties in the current knowledge state that must be addressed for SFR MST development. It is anticipated that this knowledge state assessment can enable the coordination of technology and analysis tool development discussions such that any knowledge gaps may be addressed. Sources of radionuclides considered in this report include releases originating both in-vessel and ex-vessel, including in-core fuel, primary sodium and cover gas cleanup systems, and spent fuel movement and handling. Transport phenomena affecting various release groups are identified and qualitatively discussed, including fuel pin and primary coolant retention, and behavior in the cover gas and

  2. New technologies in gastrointestinal research

    Institute of Scientific and Technical Information of China (English)

    Asbjφrn Mohr Drewes; Hans Gregersen

    2009-01-01

    This issue presents different new techniques aiming to increase our understanding of the gastrointestinal system and to improve treatment. The technologies cover selected methods to evoke and assess gut pain, new methods for imaging and physiological measurements, histochemistry, pharmacological modelling etc. There is no doubt that the methods will revolutionize the diagnostic approach in near future.

  3. Numerical Research on Hybrid Fuel Locking Device for Upward Flow Core-Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Huh, Hyung; Cho, Yeong-Garp; Yoo, Yeon-Sik; Ryu, Jeong-Soo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The assembly must be held firmly against these forces, but cannot be permanently attached to the support stand because periodic refueling of the reactor requires removal or relocation of each assembly. There are so many kinds of fuel locking device, but they are operated manually. As a part of a new project, we have investigated a hybrid fuel locking device (HFLD) for research reactor which is operated automatically. Prior method of holding down the fuel assembly includes a hybrid zero electromagnet consisting of an electromagnet and a permanent magnet. The role of an electromagnet is converged to zero power for overcoming the lifting power of a permanent magnet by controlling the coil current. At this time, a HFLD is an unlocking state. On the contrary, it is locking state that only a permanent magnet works when the power of an electromagnet is off. The results of a FEM in this work lead to the following conclusions: (1) It is possible that an electromagnet is converged to zero power for overcoming the lifting power of a permanent magnet by remote controlling the coil current. (2) At this time, it is able to detect remotely using proximity sensor whether a HFLD is latched or not.

  4. Activity report on the utilization of research reactors. Japanese Fiscal Year, 1998

    Energy Technology Data Exchange (ETDEWEB)

    Toyoda, Masayuki; Koyama, Yoshimi [eds.] [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2000-03-01

    This is the second issue of the activity report on the utilization of research reactors in the fields of neutron beam experiments, neutron activation analysis, radioisotope production, etc., performed during Japanese Fiscal Year 1998 (April 1, 1998 - March 31, 1999). All reports in this volume were described by users from JAERI and also users from the other organizations, i.e., universities, national research institutes and private companies, who have utilized our research reactor utilization facilities for the purpose of the above studies. (author)

  5. Activity report on the utilization of research reactors. Japanese Fiscal Year, 1999

    Energy Technology Data Exchange (ETDEWEB)

    Toyoda, Masayuki [ed.] [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-03-01

    This is the second issue of the activity report on the utilization of research reactors in the fields of neutron beam experiments, neutron activation analysis, radioisotope production, etc., performed during Japanese Fiscal Year 1999 (April 1, 1999 - March 31, 2000). All reports in this volume were described by users from JAERI and also users from the other organizations, i.e., universities, national research institutes and private companies, who have utilized our research reactor utilization facilities for the purpose of the above studies. (author)

  6. Energy Technology Division research summary -- 1994

    Energy Technology Data Exchange (ETDEWEB)

    1994-09-01

    Research funded primarily by the NRC is directed toward assessing the roles of cyclic fatigue, intergranular stress corrosion cracking, and irradiation-assisted stress corrosion cracking on failures in light water reactor (LWR) piping systems, pressure vessels, and various core components. In support of the fast reactor program, the Division has responsibility for fuel-performance modeling and irradiation testing. The Division has major responsibilities in several design areas of the proposed International Thermonuclear Experimental Reactor (ITER). The Division supports the DOE in ensuring safe shipment of nuclear materials by providing extensive review of the Safety Analysis Reports for Packaging (SARPs). Finally, in the nuclear area they are investigating the safe disposal of spent fuel and waste. In work funded by DOE`s Energy Efficiency and Renewable Energy, the high-temperature superconductivity program continues to be a major focal point for industrial interactions. Coatings and lubricants developed in the division`s Tribology Section are intended for use in transportation systems of the future. Continuous fiber ceramic composites are being developed for high-performance heat engines. Nondestructive testing techniques are being developed to evaluate fiber distribution and to detect flaws. A wide variety of coatings for corrosion protection of metal alloys are being studied. These can increase lifetimes significant in a wide variety of coal combustion and gasification environments.

  7. Current Status of the Transmutation Reactor Technology and Preliminary Evaluation of Transmutation Performance of the KALIMER Core

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Ser Gi; Sim, Yoon Sub; Kim, Yeong Il; Kim, Young Gyum; Lee, Byung Woon; Song, Hoon; Lee, Ki Bog; Jang, Jin Wook; Lee, Dong Uk

    2005-08-15

    Recently the most countries using the nuclear power plants for electricity generation have been faced with the problem of the preparation of the repository for the disposition of the nuclear waste generated from LWR. It was well-known that the issues related with long term risk of the radioactive wastes for the future generations are due only to 1% of the total waste. This small fraction of 1% consists of transuranic (TRU) nuclides such as Pu, Np, Am, Cm and the long lived fission products such as Tc and I. For the transuranic (TRU) nuclides, their half lives range from several years to several hundred thousands years and hence their radioactive toxicity can be lasted over very long time period. This has made the change of the rule of the fast spectrum reactor from the economical use of uranium resource through breeding to the reduction of the nuclear waste through the transmutation. The purpose of this study is to obtain the basic knowledge on the nuclear transmutation technology and to suggest the technical solution ways for the future technology development and enhancement through a survey of the state-of-art of the international research on the nuclear transmutation. The increase of the transmutation rate requires the reduction of the breeding ratio. In fact, the transmutation rate is determined by the breeding ratio. The reduction of the breeding ratio can be achieved by reducing the U-238 content in fuel or increasing the neutron leakage through core boundary or absorbing the neutrons by using some absorbers. However, the reduction of the U-238 content results in the degradation of the fuel Doppler coefficient that is one of the most important safety-related parameters and the reduction of the effective delayed neutron fraction that is related with the controllability of the reactor core. Also, the increase of the transmutation rate can lead to the increase of the coolant void reactivity worth unless some ways to reduce the coolant void reactivity are not

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-04-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

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

  10. Progress in the decommissioning planning for the Kiev’s research reactor WWR-M

    Directory of Open Access Journals (Sweden)

    Lobach Yuri N.

    2010-01-01

    Full Text Available The Kiev’s research reactor WWR-M has been in operation for more than 50 years and its further operation is planned for no less than 8-10 years. The acting nuclear legislation of Ukraine demands from the operator to perform the decommissioning planning during the reactor operation stage as early as possible. Recently, the Decommissioning Program has been approved by the regulatory body. The Program is based on the plans for the further use of the reactor site and foresees the strategy of immediate dismantling. The Program covers the whole de- commissioning process and represents the main guiding document during the whole decommissioning period, which determines and substantiates the principal technical and organizational activities on the preparation and implementation of the reactor decommissioning, the consequence of the decommissioning stages, the sequence of planned works and measures as well as the necessary conditions and infrastructure for the provision and safe implementation.

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

    Directory of Open Access Journals (Sweden)

    Wonkyeong Kim

    2016-04-01

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

  12. Technology meets research 60 years of CERN technology : selected highlights

    CERN Document Server

    Taylor, Thomas; Treille, Daniel; Wenninger, Horst

    2017-01-01

    "Big" science and advanced technology are known to cross-fertilize. This book emphasizes the interplay between particle physics and technology at CERN that has led to breakthroughs in both research and technology over the laboratory's first 60 years. The innovations, often the work of individuals or by small teams, are illustrated with highlights describing selected technologies from the domains of accelerators and detectors. The book also presents the framework and conditions prevailing at CERN that enabled spectacular advances in technology and contributed to propel the European organization into the league of leading research laboratories in the world. While the book is specifically aimed at providing information for the technically interested general public, more expert readers may also appreciate the broad variety of subjects presented. Ample references are given for those who wish to further explore a given topic.

  13. Design of a New Research Reactor: Preliminary Conceptual Design (3rd Year)

    Energy Technology Data Exchange (ETDEWEB)

    Park, Cheol; Lee, B. C.; Chae, H. T. and others

    2006-01-15

    A research reactor design is a kind of integral engineering project and a process to obtain a concrete shape through several years of concept development, conceptual design, basic design and detail design. So it requires close cooperation in various areas as well as lots of manpower and cost. The overall process at each stage may be said to be similar except for some stage-specific works. In 2005 as last year of a concept development stage, investigations on the various concepts of the fuel, reactor structure and systems which can meet the requirements established. The requirements for the process systems and I and C systems have also been embodied. The major tasks planned at the early of 2005 have been performed for each area of reactor design as follows: Establishment of the fuel and reactor core concept, and the core analysis, Preliminary thermal-hydraulic and safety analyses for the conceptual cores, Establishment and improvement of analysis system, Concept developments of the reactor structures and major systems, Test and test plan to verify the developed concepts, International cooperation to establish the foundations for exporting a research reactor.

  14. Proceedings of the US Nuclear Regulatory Commission twentieth water reactor safety information meeting; Volume 2, Severe accident research, Thermal hydraulics

    Energy Technology Data Exchange (ETDEWEB)

    Weiss, A.J. [comp.] [Brookhaven National Lab., Upton, NY (United States)

    1993-03-01

    This three-volume report contains papers presented at the Twentieth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 21--23, 1992. The papers describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included 10 different papers presented by researchersfrom CEC, China, Finland, France, Germany, Japan, Spain and Taiwan. Selected papers have been processed separately for inclusion in the Energy Science and Technology Database.

  15. Constituting Information Technology Research: The Experience of IT Researchers

    Science.gov (United States)

    Pham, Binh; Bruce, Christine; Stoodley, Ian

    2005-01-01

    The collective consciousness of effective groups of researchers is characterized by shared understandings of their research object or territory. In this study, we adopted a phenomenographic approach to investigate information technology (IT) research, and its objects and territories, as they are constituted in the experience of IT researchers.…

  16. End of the line for Harwell's Dido and Pluto research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Tom; Nicholson, K.

    1990-04-01

    After 34 years of continuous operation the Dido and Pluto research reactors were shutdown for the last time on the 31 March 1990. The history of their development and contributions to the UK nuclear programme, isotope production, support to industry and basic scientific research are described. (author).

  17. Decommissioning of the Astra research reactor: Review and status on July 2003

    Directory of Open Access Journals (Sweden)

    Meyer Franz

    2003-01-01

    Full Text Available The paper describes work on the decommissioning of the ASTRA research reactor at the Austrian Research Centers Seibersdorf. Organizational, planning, and dismantling work done until July 2003 including radiation protection and waste management procedures as well as the current status of the project are presented. Completion of the decommissioning activities is planned for 2006.

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

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sujin; Kim, Jeongeun; Kim, Hyeonil [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

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

  19. Bubble column reactor fluid dynamic study at pilot plant scale for residue and extra heavy crude oil upgrading technology

    Energy Technology Data Exchange (ETDEWEB)

    Sardella, R.; Medina, H. [Infrastructure and Upgrading Department PDVSA-Intevep (Venezuela); Zacarias, L.; Paiva, M. [Refining Department. PDVSA-Intevep (Venezuela)

    2011-07-01

    Bubble column reactors are used in several applications because of their simplicity and low cost; a new technology was developed to convert extra heavy crude oil into upgraded crude using a bubble column reactor. To design this kind of reactor, a lot of parameters like flow regime, gas hold up and dispersion coefficient have to be taken into account. This study aimed at determining the fluid dynamic behaviour of a bubble column working under Aquaconversion operating conditions. Experiments were undertaken on air-tap water and air-light oil systems under atmospheric conditions with various gas superficial velocities and liquid flowrates. Results showed that gas hold up increases with superficial gas velocity but is independent of liquid flowrate and that both systems tested work at the same flow regimes. This paper showed that under the experimental conditions used, this reactor tends to be a complete mixing reactor.

  20. A mechanism for proven technology foresight for emerging fast reactor designs and concepts

    Science.gov (United States)

    Anuar, Nuraslinda; Muhamad Pauzi, Anas

    2016-01-01

    The assessment of emerging nuclear fast reactor designs and concepts viability requires a combination of foresight methods. A mechanism that allows for the comparison and quantification of the possibility of being a proven technology in the future, β for the existing fast reactor designs and concepts is proposed as one of the quantitative foresight method. The methodology starts with the identification at the national or regional level, of the factors that would affect β. The factors are then categorized into several groups; economic, social and technology elements. Each of the elements is proposed to be mathematically modelled before all of the elemental models can be combined. Once the overall β model is obtained, the βmin is determined to benchmark the acceptance as a candidate design or concept. The β values for all the available designs and concepts are then determined and compared with the βmin, resulting in a list of candidate designs that possess the β value that is larger than the βmin. The proposed methodology can also be applied to purposes other than technological foresight.

  1. A mechanism for proven technology foresight for emerging fast reactor designs and concepts

    Energy Technology Data Exchange (ETDEWEB)

    Anuar, Nuraslinda, E-mail: nuraslinda@uniten.edu.my; Muhamad Pauzi, Anas, E-mail: anas@uniten.edu.my [College of Engineering, Universiti Tenaga Nasional, Jalan IKRAM-UNITEN, 43000 Kajang, Selangor (Malaysia)

    2016-01-22

    The assessment of emerging nuclear fast reactor designs and concepts viability requires a combination of foresight methods. A mechanism that allows for the comparison and quantification of the possibility of being a proven technology in the future, β for the existing fast reactor designs and concepts is proposed as one of the quantitative foresight method. The methodology starts with the identification at the national or regional level, of the factors that would affect β. The factors are then categorized into several groups; economic, social and technology elements. Each of the elements is proposed to be mathematically modelled before all of the elemental models can be combined. Once the overall β model is obtained, the β{sub min} is determined to benchmark the acceptance as a candidate design or concept. The β values for all the available designs and concepts are then determined and compared with the β{sub min}, resulting in a list of candidate designs that possess the β value that is larger than the β{sub min}. The proposed methodology can also be applied to purposes other than technological foresight.

  2. Proceedings of the 1998 workshop on the utilization of research reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-10-01

    The 1998 Workshop on the Utilization of Research Reactors, which is the seventh Workshop on the theme of research reactor utilization was held in Yogyakarta and Serpong, Indonesia from February 8 to 14. This Workshop was executed based on the agreement in the Ninth International Conference for Nuclear Cooperation in Asia (ICNCA) held in Tokyo, March 1998. The whole Workshop consists of the Workshop on the theme of following three fields, 1) Neutron Scattering, 2) Neutron Activation analysis and 3) Safe Operation and Maintenance of Research Reactor, and the Sub-workshop carried out the experiment of Neutron Activation analysis. The total number of participants for the workshop was about 100 people from 8 countries, i.e. Australia, China, Indonesia, Korea, Malaysia, Thailand, Vietnam and Japan. The 38 papers are indexed individually. (J.P.N.)

  3. Annual report of department of research reactor, 1995 (April 1, 1995 - March 31, 1996)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-12-01

    The Department of Research Reactor is responsible for the operation, maintenance, utilization and related R and D works of the research reactors including JRR-2, JRR-3M (new JRR-3) and JRR-4. This report describes the activities of our department in fiscal year of 1995 and it also includes some of the technical topics on the works mentioned above. As for the research reactors, we carried out the operation, maintenance, irradiation utilization, neutron beam experiments, technical management including fuels and water chemistry, radiation monitoring as related R and D works. The international cooperations between the developing countries and our department were also made concerning the operation, utilization and safety analysis for nuclear facilities. (author)

  4. Annual report of Department of Research Reactor, 1996. April 1, 1996 - March 31, 1997

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-03-01

    The Department of Research Reactor is responsible for the operation, maintenance, utilization and related R and D works of the research reactors including JRR-2, JRR-3M (new JRR-3) and JRR-4. This report describes the activities of our department in fiscal year of 1996 and it also includes some of the technical topics on the works mentioned above. As for the research reactors, we carried out the operation, maintenance, irradiation utilization, neutron beam experiments, technical management including fuels and water chemistry, radiation monitoring as related R and D works. The international cooperations between the developing countries and our department were also made concerning the operation, utilization and safety analysis for nuclear facilities. (author)

  5. Annual report of department of research reactors, 2001. April 1, 2001 - March 31, 2002

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2002-12-01

    The Department of Research Reactor is responsible for the operation, maintenance, utilization of the JRR-3 and the JRR-4 and for the related R and D. Besides RI production including its R and D are carried out. This report describes the activities of the department in fiscal year of 2001 and it also includes some of the technical topics on the works mentioned above. As for the research reactors, we carried out the operation, maintenance, the utilization of irradiation and neutron beam experiments, technical management including fuels and water chemistry, radiation monitoring as related R and D works. RI production and its R and D works were conducted as well. The international cooperations between the developing countries and the department were also made concerning the operation, utilization and safety analysis for research reactors. (author)

  6. Annual report of department of research reactor, 1999. April 1, 1999 - March 31, 2000

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-03-01

    The Department of Research Reactor is responsible for the operation, maintenance, utilization of the JRR-3M (new JRR-3) and the JRR-4 and for the related R and D. Besides the decommissioning of the JRR-2 and RI production including its R and D are carried out. This report describes the activities of the department in fiscal year of 1999 and it also includes some of the technical topics on the works mentioned above. As for the research reactors, we carried out the operation, maintenance, the utilization of irradiation and neutron beam experiments, technical management including fuels and water chemistry, radiation monitoring as related R and D works. RI production and its R and D works were conducted as well. The international cooperations between the developing countries and the department were also made concerning the operation, utilization and safety analysis for research reactors. (author)

  7. Overview of Stirling Technology Research at NASA Glenn Research Center

    Science.gov (United States)

    Wilson, Scott D.; Schifer, Nicholas A.; Williams, Zachary D.; Metscher, Jonathan F.

    2016-01-01

    Stirling Radioisotope Power Systems (RPSs) are under development to provide power on future space science missions where robotic spacecraft will orbit, fly by, land, or rove using less than a quarter of the plutonium the currently available RPS uses to produce about the same power. NASA Glenn Research Center's newly formulated Stirling Cycle Technology Development Project (SCTDP) continues development of Stirling-based systems and subsystems, which include a flight-like generator and related housing assembly, controller, and convertors. The project also develops less mature technologies under Stirling Technology Research, with a focus on demonstration in representative environments to increase the technology readiness level (TRL). Matured technologies are evaluated for selection in future generator designs. Stirling Technology Research tasks focus on a wide variety of objectives, including increasing temperature capability to enable new environments, reducing generator mass and/or size, improving reliability and system fault tolerance, and developing alternative designs. The task objectives and status are summarized.

  8. Computer science research and technology volume 3

    CERN Document Server

    Bauer, Janice P

    2011-01-01

    This book presents leading-edge research from across the globe in the field of computer science research, technology and applications. Each contribution has been carefully selected for inclusion based on the significance of the research to this fast-moving and diverse field. Some topics included are: network topology; agile programming; virtualization; and reconfigurable computing.

  9. The Need for Cyber-Informed Engineering Expertise for Nuclear Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, Robert Stephen [Idaho National Laboratory

    2015-12-01

    Engineering disciplines may not currently understand or fully embrace cyber security aspects as they apply towards analysis, design, operation, and maintenance of nuclear research reactors. Research reactors include a wide range of diverse co-located facilities and designs necessary to meet specific operational research objectives. Because of the nature of research reactors (reduced thermal energy and fission product inventory), hazards and risks may not have received the same scrutiny as normally associated with power reactors. Similarly, security may not have been emphasized either. However, the lack of sound cybersecurity defenses may lead to both safety and security impacts. Risk management methodologies may not contain the foundational assumptions required to address the intelligent adversary’s capabilities in malevolent cyber attacks. Although most research reactors are old and may not have the same digital footprint as newer facilities, any digital instrument and control function must be considered as a potential attack platform that can lead to sabotage or theft of nuclear material, especially for some research reactors that store highly enriched uranium. This paper will provide a discussion about the need for cyber-informed engineering practices that include the entire engineering lifecycle. Cyber-informed engineering as referenced in this paper is the inclusion of cybersecurity aspects into the engineering process. A discussion will consider several attributes of this process evaluating the long-term goal of developing additional cyber safety basis analysis and trust principles. With a culture of free information sharing exchanges, and potentially a lack of security expertise, new risk analysis and design methodologies need to be developed to address this rapidly evolving (cyber) threatscape.

  10. KSC Education Technology Research and Development Plan

    Science.gov (United States)

    Odell, Michael R. L.

    2003-01-01

    Educational technology is facilitating new approaches to teaching and learning science, technology, engineering, and mathematics (STEM) education. Cognitive research is beginning to inform educators about how students learn providing a basis for design of more effective learning environments incorporating technology. At the same time, access to computers, the Internet and other technology tools are becoming common features in K-20 classrooms. Encouraged by these developments, STEM educators are transforming traditional STEM education into active learning environments that hold the promise of enhancing learning. This document illustrates the use of technology in STEM education today, identifies possible areas of development, links this development to the NASA Strategic Plan, and makes recommendations for the Kennedy Space Center (KSC) Education Office for consideration in the research, development, and design of new educational technologies and applications.

  11. Pacific Northwest Laboratory Monthly Activities Report APRIL 1966 on AEC Division of Reactor Development and Technology

    Energy Technology Data Exchange (ETDEWEB)

    S. L. Fawcett

    1966-05-01

    This report has the following sections: Summary of Activities; Civilian Power Reactors; Applied and Reactor Physics; Reactor Fuels and Materials; Engineering Development; Plutonium Recycle Program; Advanced Systems; and Nuclear Safety.

  12. Pacific Northwest Laboratory Monthly Activities Report March 1966 On AEC Division of Reactor Development and Technology

    Energy Technology Data Exchange (ETDEWEB)

    S. L. Fawcett

    1966-04-01

    This report has the following sections: Summary of Activities; Civilian Power Reactors; Applied and Reactor Physics; Reactor Fuels and Materials; Engineering Development; Plutonium Recycle Program; Advanced Systems; and Nuclear Safety.

  13. Human reliability analysis of the Tehran research reactor using the SPAR-H method

    Directory of Open Access Journals (Sweden)

    Barati Ramin

    2012-01-01

    Full Text Available The purpose of this paper is to cover human reliability analysis of the Tehran research reactor using an appropriate method for the representation of human failure probabilities. In the present work, the technique for human error rate prediction and standardized plant analysis risk-human reliability methods have been utilized to quantify different categories of human errors, applied extensively to nuclear power plants. Human reliability analysis is, indeed, an integral and significant part of probabilistic safety analysis studies, without it probabilistic safety analysis would not be a systematic and complete representation of actual plant risks. In addition, possible human errors in research reactors constitute a significant part of the associated risk of such installations and including them in a probabilistic safety analysis for such facilities is a complicated issue. Standardized plant analysis risk-human can be used to address these concerns; it is a well-documented and systematic human reliability analysis system with tables for human performance choices prepared in consultation with experts in the domain. In this method, performance shaping factors are selected via tables, human action dependencies are accounted for, and the method is well designed for the intended use. In this study, in consultations with reactor operators, human errors are identified and adequate performance shaping factors are assigned to produce proper human failure probabilities. Our importance analysis has revealed that human action contained in the possibility of an external object falling on the reactor core are the most significant human errors concerning the Tehran research reactor to be considered in reactor emergency operating procedures and operator training programs aimed at improving reactor safety.

  14. Electronic imaging system for neutron radiography at a low power research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ferreira, F.J.O., E-mail: fferreira@ien.gov.b [Instituto de Engenharia Nuclear, Comissao Nacional de Energia Nuclear, Caixa Postal 68550, CEP 21945-970, Rio de Janeiro (Brazil); Silva, A.X.; Crispim, V.R. [PEN/COPPE-DNC/POLI CT, Universidade Federal do Rio de Janeiro, Ilha do Fundao, Caixa Postal 68509, 21945-970 Rio de Janeiro (Brazil)

    2010-08-15

    This paper describes an electronic imaging system for producing real time neutron radiography from a low power research reactor, which will allow inspections of samples with high efficiency, in terms of measuring time and result analysis. This system has been implanted because of its potential use in various scientific and industrial areas where neutron radiography with photographic film could not be applied. This real time system is installed in neutron radiography facility of Argonauta nuclear research reactor, at the Instituto de Engenharia Nuclear of the Comissao Nacional de Energia Nuclear, in Brazil. It is adequate to perform real time neutron radiography of static and dynamic events of samples.

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

    Energy Technology Data Exchange (ETDEWEB)

    1993-07-01

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

  16. Advanced Construction Technology Center Research

    Science.gov (United States)

    1993-03-19

    M.S., Ceramic Engineering, May 1989 RESEARCH ASSISTANTS: 1 J.M. Bukowski Ph.D., Ceramic Engineering, August 1993* Hung-Yuan Hsieh M.S., Civil...Parallel Processing (St. Charles , IL, August 1988) 204-211 (1988). 97 U LO, T.M. Data modeling of three-dimensional objects. M.S. thesis, J. Liu, advisor U

  17. Air Force Research Laboratory Technology Milestones 2008

    Science.gov (United States)

    2008-01-01

    develop a unique measurement platform employing tunable diode laser absorption spectroscopy ( TDLAS ). The TDLAS platform provides a novel approach to...conduct research in the exploration and development of fundamental hypersonic aerospace technologies. TDLAS experiments are scheduled for three of...team expects that the TDLAS measurement platform will achieve Technology Readiness Level 6 status (i.e., system/subsystem model or prototype

  18. Research and technology, fiscal year 1982

    Science.gov (United States)

    1982-01-01

    Advanced studies are reviewed. Atmospheric sciences, magnetospheric physics, solar physics, gravitational physics, astronomy, and materials processing in space comprise the research programs. Large space systems, propulsion technology, materials and processes, electrical/electronic systems, data bases/design criteria, and facilities development comprise the technology development activities.

  19. Science, Technology and Arts Research Journal

    African Journals Online (AJOL)

    Science, Technology and Arts Research Journal ... Determination of Nutritive Value and Mineral Elements of Some Species of Genus Memecylon Linn. ... towards Gender and Intimate Partner Violence Against Women in Eastern Ethiopia: A ...

  20. The Europlanet Research Infrastructure and Technology Foresight

    Science.gov (United States)

    Grande, M.; Europlanet Community

    2016-10-01

    The Europlanet 2020 Research Infrastructure is a project to integrate and support planetary science activities across Europe. The project is funded under the European Commission's Horizon 2020 programme. Technology Foresight is a key activity.

  1. Neutron Ghost Imaging Technology Research on CARR Reactor

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

    Ghost imaging is also known as quantum imaging. Different from the classical imaging, the neutron ghost imaging is based on the quantum mechanics properties of light field and its intrinsic parallel characteristic, and developed by new optical

  2. Radioisotope radiotherapy research and achievements at the University of Missouri Research Reactor

    Science.gov (United States)

    Ehrhardt, G. J.; Ketring, A. R.; Cutler, C. S.

    2003-01-01

    The University of Missouri Research Reactor (MURR) in collaboration with faculty in other departments at the University of Missouri has been involved in developing new means of internal radioisotopic therapy for cancer for many years. These efforts have centered on methods of targeting radioisotopes such as brachytherapy, embolisation of liver tumors with radioactive microspheres, small-molecule-labelled chelates for the treatment of bone cancer, and various means of radioimmunotherapy or labelled receptor agent targeting. This work has produced two radioactive agents, Sm-153 Quadramet™ and Y-90 TheraSphere™, which have U.S. Food and Drug Administration approval for the palliation of bone cancer pain and treatment of inoperable liver cancer, respectively. MURR has also pioneered development of other beta-emitting isotopes for internal radiotherapy such as Re-186, Re-188, Rh-105, Ho-166, Lu-177, and Pm-149, many of which are in research and clinical trials throughout the U.S. and the world. This important work has been made possible by the very high neutron flux available at MURR combined with MURR's outstanding reliability of operation and flexibility in meeting the needs of researchers and the radiopharmaceutical industry.

  3. Nanoscale Science, Engineering and Technology Research Directions

    Energy Technology Data Exchange (ETDEWEB)

    Lowndes, D. H.; Alivisatos, A. P.; Alper, M.; Averback, R. S.; Jacob Barhen, J.; Eastman, J. A.; Imre, D.; Lowndes, D. H.; McNulty, I.; Michalske, T. A.; Ho, K-M; Nozik, A. J.; Russell, T. P.; Valentin, R. A.; Welch, D. O.; Barhen, J.; Agnew, S. R.; Bellon, P.; Blair, J.; Boatner, L. A.; Braiman, Y.; Budai, J. D.; Crabtree, G. W.; Feldman, L. C.; Flynn, C. P.; Geohegan, D. B.; George, E. P.; Greenbaum, E.; Grigoropoulos, C.; Haynes, T. E.; Heberlein, J.; Hichman, J.; Holland, O. W.; Honda, S.; Horton, J. A.; Hu, M. Z.-C.; Jesson, D. E.; Joy, D. C.; Krauss, A.; Kwok, W.-K.; Larson, B. C.; Larson, D. J.; Likharev, K.; Liu, C. T.; Majumdar, A.; Maziasz, P. J.; Meldrum, A.; Miller, J. C.; Modine, F. A.; Pennycook, S. J.; Pharr, G. M.; Phillpot, S.; Price, D. L.; Protopopescu, V.; Poker, D. B.; Pui, D.; Ramsey, J. M.; Rao, N.; Reichl, L.; Roberto, J.; Saboungi, M-L; Simpson, M.; Strieffer, S.; Thundat, T.; Wambsganss, M.; Wendleken, J.; White, C. W.; Wilemski, G.; Withrow, S. P.; Wolf, D.; Zhu, J. H.; Zuhr, R. A.; Zunger, A.; Lowe, S.

    1999-01-01

    This report describes important future research directions in nanoscale science, engineering and technology. It was prepared in connection with an anticipated national research initiative on nanotechnology for the twenty-first century. The research directions described are not expected to be inclusive but illustrate the wide range of research opportunities and challenges that could be undertaken through the national laboratories and their major national scientific user facilities with the support of universities and industry.

  4. Proceedings of the 1999 workshop on the utilization of research reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-10-01

    The 1999 workshop on the utilization of reactors, which is the eighth workshop on the theme of research reactor utilization was held at JAERI Tokai and Mito Plaza Hotel, in Japan from November 25 to December 2. This workshop was executed based on the agreement in the Tenth International conference for Nuclear Cooperation in Asia (ICNCA) held in Tokyo, March 1999. The whole workshop consists of the workshop on the theme of following three fields, 1) neutron scattering, 2) radioisotope production and 3) safe operation and maintenance of research reactor, and the sub-workshop carried out the experiments of small angle neutron scattering. The total number of participants for the workshop was about 70 people from 9 countries, i.e. Australia, China, Indonesia, Korea, Malaysia, The Philippines, Thailand, Vietnam and Japan. The 37 of the presented papers are indexed individually. (J.P.N.)

  5. Characterization of the Annular Core Research Reactor (ACRR Neutron Radiography System Imaging Plane

    Directory of Open Access Journals (Sweden)

    Kaiser Krista

    2016-01-01

    Full Text Available The Annular Core Research Reactor (ACRR at Sandia National Laboratories (SNL is an epithermal pool-type research reactor licensed up to a thermal power of 2.4 MW. The ACRR facility has a neutron radiography facility that is used for imaging a wide range of items including reactor fuel and neutron generators. The ACRR neutron radiography system has four apertures (65:1, 125:1, 250:1, and 500:1 available to experimenters. The neutron flux and spectrum as well as the gamma dose rate were characterized at the imaging plane for the ACRR's neutron radiography system for the 65:1, 125:1 and 250:1 apertures.

  6. Initial verification and validation of RAZORBACK - A research reactor transient analysis code

    Energy Technology Data Exchange (ETDEWEB)

    Talley, Darren G. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States)

    2015-09-01

    This report describes the work and results of the initial verification and validation (V&V) of the beta release of the Razorback code. Razorback is a computer code designed to simulate the operation of a research reactor (such as the Annular Core Research Reactor (ACRR)) by a coupled numerical solution of the point reactor kinetics equations, the energy conservation equation for fuel element heat transfer, and the mass, momentum, and energy conservation equations for the water cooling of the fuel elements. This initial V&V effort was intended to confirm that the code work to-date shows good agreement between simulation and actual ACRR operations, indicating that the subsequent V&V effort for the official release of the code will be successful.

  7. Characterization of the Annular Core Research Reactor (ACRR) Neutron Radiography System Imaging Plane

    Science.gov (United States)

    Kaiser, Krista; Chantel Nowlen, K.; DePriest, K. Russell

    2016-02-01

    The Annular Core Research Reactor (ACRR) at Sandia National Laboratories (SNL) is an epithermal pool-type research reactor licensed up to a thermal power of 2.4 MW. The ACRR facility has a neutron radiography facility that is used for imaging a wide range of items including reactor fuel and neutron generators. The ACRR neutron radiography system has four apertures (65:1, 125:1, 250:1, and 500:1) available to experimenters. The neutron flux and spectrum as well as the gamma dose rate were characterized at the imaging plane for the ACRR's neutron radiography system for the 65:1, 125:1 and 250:1 apertures.

  8. Radiation dosimetry for NCT facilities at the Brookhaven Medical Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Holden, N.E.; Hu, J.P.; Greenberg, D.D.; Reciniello, R.N.

    1998-12-31

    Brookhaven Medical Research Reactor (BMRR) is a 3 mega-watt (MW) heterogeneous, tank-type, light water cooled and moderated, graphite reflected reactor, which was designed for medical and biological studies and became operational in 1959. Over time, the BMRR was modified to provide thermal and epithermal neutron beams suitable for research studies. NCT studies have been performed at both the epithermal neutron irradiation facility (ENIF) on the east side of the BMRR reactor core and the thermal neutron irradiation facility (TNIF) on the west side of the core. Neutron and gamma-ray dosimetry performed from 1994 to the present in both facilities are described and the results are presented and discussed.

  9. Use of research and test reactors for SPD development and calibration

    Energy Technology Data Exchange (ETDEWEB)

    LaFontaine, M.W.R. [Physics Solutions Inc., Kitchener, ON (Canada)

    2011-07-01

    Prior to using a research or test reactor for performance studies or calibration of self powered detectors, it is first necessary to fully characterize the reactor environment in the region to be utilized. This presentation details Characterization Experiments performed to quantify research/test reactor core/site parameters as they would apply for use with SPD applications. Methods will be described to: Determine the Westcott parameter, r (T {sub n}/T {sub o}) , for the region of interest; Characterize the neutron energy spectrum in terms of the cadmium absorption cut-off, i.e., consider neutrons of energy 5kT < 0.13 eV to be thermal neutrons, and neutrons of energy 5kT > 0.13 eV to be epithermal neutrons; Determine T {sub n}, the effective neutron temperature, in the region of interest; Determine the gamma flux in the region of interest; and, Establish SPD calibration standard detectors.

  10. Development of a Monolithic Research Reactor Fuel Type at Argonne National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Clark, C.R.; Briggs, R.J.

    2004-10-06

    The Reduced Enrichment for Research and Test Reactors (RERTR) program has been tasked with the conversion of research reactors from highly enriched to low-enriched uranium (LEU). To convert several high power reactors, monolithic fuel, a new fuel type, is being developed. This fuel type replaces the standard fuel dispersion with a fuel alloy foil, which allows for fuel densities far in excess of that found in dispersion fuel. The single-piece fuel foil also contains a significantly lower interface area between the fuel and the aluminum in the plate than the standard fuel type, limiting the amount of detrimental fuel-aluminum interaction that can occur. Implementation of monolithic fuel is dependant on the development of a suitable fabrication method as traditional roll-bonding techniques are inadequate.

  11. High-temperature gas-cooled reactor technology development program. Annual progress report for period ending December 31, 1980

    Energy Technology Data Exchange (ETDEWEB)

    1981-08-01

    Research activities are described concerning HTGR chemistry; fueled graphite development; prestressed concrete pressure vessel development; structural materials; HTGR graphite studies; HTR core evaluation; reactor physics; shielding; application and project assessments; and HTR Core Flow Test Loop studies.

  12. Determination of the optimal positions for installing gamma ray detection systems at Tehran Research Reactor

    Science.gov (United States)

    Sayyah, A.; Rahmani, F.; Khalafi, H.

    2015-09-01

    Dosimetric instruments must constantly monitor radiation dose levels in different areas of nuclear reactor. Tehran Research Reactor (TRR) has seven beam tubes for different research purposes. All the beam tubes extend from the reactor core to Beam Port Floor (BPF) of the reactor facility. During the reactor operation, the gamma rays exiting from each beam tube outlet produce a specific gamma dose rate field in the space of the BPF. To effectively monitor the gamma dose rates on the BPF, gamma ray detection systems must be installed in optimal positions. The selection of optimal positions is a compromise between two requirements. First, the installation positions must possess largest gamma dose rates and second, gamma ray detectors must not be saturated in these positions. In this study, calculations and experimental measurements have been carried out to identify the optimal positions of the gamma ray detection systems. Eight three dimensional models of the reactor core and related facilities corresponding to eight scenarios have been simulated using MCNPX Monte Carlo code to calculate the gamma dose equivalent rate field in the space of the BPF. These facilities are beam tubes, thermal column, pool, BPF space filled with air, facilities such as neutron radiography facility, neutron powder diffraction facility embedded in the beam tubes as well as biological shields inserted into the unused beam tubes. According to the analysis results of the combined gamma dose rate field, three positions on the north side and two positions on the south side of the BPF have been recognized as optimal positions for installing the gamma ray detection systems. To ensure the consistency of the simulation data, experimental measurements were conducted using TLDs (600 and 700) pairs during the reactor operation at 4.5 MW.

  13. Novel omics technologies in nutrition research.

    Science.gov (United States)

    Zhang, Xuewu; Yap, Yeeleng; Wei, Dong; Chen, Gu; Chen, Feng

    2008-01-01

    A key scientific objective of nutrition research is to determine the role of diet in metabolic regulation and to improve health. There are many new opportunities and challenges for the nutrition research in post-genome era. Novel omics technologies and bioinformatics tools offer enormous potential to investigate the complex relationship between nutrition and metabolism. An overview of omics technologies in nutrition research is presented in this paper, which focuses on recent applications of nutritranscriptomics, nutriproteomics and nutrimetabolomics in nutrition research. The potential limitations and future prospects are also discussed. The combination of various omics technologies, systems biology, will greatly facilitate the discovery of new biomarkers associated with specific nutrients or other dietary factors. It can be expected that the future omics-based human nutrition research can provide personalized dietary recommendations for disease prevention.

  14. Social technologies and socialization of research

    Directory of Open Access Journals (Sweden)

    Jos Leijten

    2009-09-01

    Full Text Available Whether we like it or not, and how many difficulties this may pose, scientific research and technology are becoming the “property” of everybody and increasingly will become subject of public guidance and political decision making. Socialization happens because what people think, want and do has become central to the development of science and technology. Socialization of research is simply happening because it is the development characteristic of a society in which knowledge is becoming the main driving force. And just like in agricultural or industrial societies in the past it leads to (re-invent the institutions and mechanisms which allow the knowledge society to function properly.This note will further explore the developments contributing to the socialization of research and their impact on research and research institutes. It will focus more on technologies than on science per se, because applications and usage will become the main drivers.

  15. Improvement of the reactivity computer for HANARO research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Min Jin; Park, S. J.; Jung, H. S.; Choi, Y. S.; Lee, K. H.; S