WorldWideScience

Sample records for general atomic standard reactor

  1. GASSAR-6 (General Atomic Standard Safety Analysis Report). Volume 1

    International Nuclear Information System (INIS)

    1975-01-01

    A standard nuclear steam system for a 3000 MW(t), 1160 MW(e) high temperature gas-cooled reactor (HTGR) nuclear power station is described. The HTGR operates on a uranium-235/thorium-232 cycle. Spherical fuel pellets are coated with multiple layers of pyrolytic carbon, bonded into rods, and encased in hexagonal graphite fuel elements. The core is nuclear-purity-grade, near isotropic graphite machined in hexagonal blocks, which serves as moderator and the heat transfer medium between the fuel and the coolant. Forced helium is the primary coolant and water is the secondary coolant. A prestressed concrete reactor vessel (PCRV) houses the reactor core, primary coolant system and portions of the secondary coolant system, steam generators and circulators. A continuous internal steel liner in the PCRV acts as the primary coolant boundary and sealing membrane. The control rods contain boron carbide in a graphite matrix sheathed in Incoloy 800 cans. Boron constitutes about 40 percent of the absorber material by volume. The control rod drives provide insertion and withdrawal rates consistent with the required reactivity changes for operational load fluctuations and reactor shutdown. Control rods have shim and trip capability. (U.S.)

  2. Atomic reactor thermal engineering

    International Nuclear Information System (INIS)

    Kim, Gwang Ryong

    1983-02-01

    This book starts the introduction of atomic reactor thermal engineering including atomic reaction, chemical reaction, nuclear reaction neutron energy and soon. It explains heat transfer, heat production in the atomic reactor, heat transfer of fuel element in atomic reactor, heat transfer and flow of cooler, thermal design of atomic reactor, design of thermodynamics of atomic reactor and various. This deals with the basic knowledge of thermal engineering for atomic reactor.

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

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  4. Standard technical specifications for General Electric boiling water reactors

    International Nuclear Information System (INIS)

    1979-08-01

    This Standard Technical Specification (STS) has been structured for the broadest possible use on General Electric plants currently being reviewed for an Operating License. Optional specifications are provided for those features and systems which may be included in individual plant designs but are not generic in their scope of application. This revision of the GE-STS does not typically include requirements which may be added or revised as a result of the NRC staff's further review of the Three Mile Island incident

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  6. Guide to General Atomic studies of hypothetical nuclear driven accidents for the Fort St. Vrain reactor

    International Nuclear Information System (INIS)

    Wei, T.; Tobias, M.

    1974-03-01

    The work of the General Atomic Company (GAC) in preparing those portions of the Final Safety Analysis Report for the Fort St. Vrain Reactor (FSV) having to do with hypothetical nuclear driven accidents has been reviewed and a guide to this literature has been prepared. The sources for this study are the Final Safety Analysis Report itself, the Quarterly and Monthly Progress Reports, Topical Reports, and Technical Specifications. The problems considered and the methods used are outlined. An appendix gives a systematic analysis which was used as a guide in organizing the references. (U.S.)

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

    International Nuclear Information System (INIS)

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

    1976-01-01

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

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

    International Nuclear Information System (INIS)

    Bottimore, R.R.

    1980-12-01

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

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  10. Atomic hydrogen reactor

    International Nuclear Information System (INIS)

    Massip de Turville, C.M.D.

    1982-01-01

    Methods are discussed of generating heat in an atomic hydrogen reactor which involve; the production of atomic hydrogen by an electrical discharge, the capture of nascent neutrons from atomic hydrogen in a number of surrounding steel alloy tubes having a high manganese content to produce 56 Mn, the irradiation of atomic hydrogen by the high energy antineutrinos from the beta decay of 56 Mn to yield nascent neutrons, and the removal of the heat generated by the capture of nascent neutrons by 55 Mn and the beta decay of 56 Mn. (U.K.)

  11. A General Small-Scale Reactor To Enable Standardization and Acceleration of Photocatalytic Reactions.

    Science.gov (United States)

    Le, Chi Chip; Wismer, Michael K; Shi, Zhi-Cai; Zhang, Rui; Conway, Donald V; Li, Guoqing; Vachal, Petr; Davies, Ian W; MacMillan, David W C

    2017-06-28

    Photocatalysis for organic synthesis has experienced an exponential growth in the past 10 years. However, the variety of experimental procedures that have been reported to perform photon-based catalyst excitation has hampered the establishment of general protocols to convert visible light into chemical energy. To address this issue, we have designed an integrated photoreactor for enhanced photon capture and catalyst excitation. Moreover, the evaluation of this new reactor in eight photocatalytic transformations that are widely employed in medicinal chemistry settings has confirmed significant performance advantages of this optimized design while enabling a standardized protocol.

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

    Science.gov (United States)

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

    2014-06-11

    Bismuth was evaluated as internal standard for Pb determination by line source flame atomic absorption spectrometry (LS FAAS), high-resolution continuum source flame atomic absorption spectrometry (HR-CS FAAS) and line source graphite furnace atomic absorption spectrometry (LS GFAAS). Analysis of samples containing different matrices indicated close relationship between Pb and Bi absorbances. Correlation coefficients of calibration curves built up by plotting A(Pb)/A(Bi)versus Pb concentration were higher than 0.9953 (FAAS) and higher than 0.9993 (GFAAS). Recoveries of Pb improved from 52-118% (without IS) to 97-109% (IS, LS FAAS); 74-231% (without IS) to 96-109% (IS, HR-CS FAAS); and 36-125% (without IS) to 96-110% (IS, LS GFAAS). The relative standard deviations (n=12) were reduced from 0.6-9.2% (without IS) to 0.3-4.3% (IS, LS FAAS); 0.7-7.7% (without IS) to 0.1-4.0% (IS, HR-CS FAAS); and 2.1-13% (without IS) to 0.4-5.9% (IS, LS GFAAS). Copyright © 2014 Elsevier B.V. All rights reserved.

  13. Processing of FRG high-temperature gas-cooled reactor fuel elements at General Atomic under the US/FRG cooperative agreement for spent fuel elements

    International Nuclear Information System (INIS)

    Holder, N.D.; Strand, J.B.; Schwarz, F.A.; Drake, R.N.

    1981-11-01

    The Federal Republic of Germany (FRG) and the United States (US) are cooperating on certain aspects of gas-cooled reactor technology under an umbrella agreement. Under the spent fuel treatment development section of the agreement, both FRG mixed uranium/ thorium and low-enriched uranium fuel spheres have been processed in the Department of Energy-sponsored cold pilot plant for high-temperature gas-cooled reactor (HTGR) fuel processing at General Atomic Company in San Diego, California. The FRG fuel spheres were crushed and burned to recover coated fuel particles suitable for further treatment for uranium recovery. Successful completion of the tests described in this paper demonstrated certain modifications to the US HTGR fuel burining process necessary for FRG fuel treatment. Results of the tests will be used in the design of a US/FRG joint prototype headend facility for HTGR fuel

  14. The general atomic strand winding machine

    International Nuclear Information System (INIS)

    Matt, P.

    1976-01-01

    In conjunction with the integrated development of their high temperature gas cooled reactors (HTGR), General Atomic of San Diego, USA, also developed a strand winding system for the horizontal prestressing of pressure vessels. The machine lay-out, its capabilities and the test program carried out in the laboratory and on a full scale pressure vessel model are described. (author)

  15. IAEA safety standards for research reactors

    International Nuclear Information System (INIS)

    Abou Yehia, H.

    2007-01-01

    The general structure of the IAEA Safety Standards and the process for their development and revision are briefly presented and discussed together with the progress achieved in the development of Safety Standards for research reactor. These documents provide the safety requirements and the key technical recommendations to achieve enhanced safety. They are intended for use by all organizations involved in safety of research reactors and developed in a way that allows them to be incorporated into national laws and regulations. The author reviews the safety standards for research reactors and details their specificities. There are 4 published safety standards: 1) Safety assessment of research reactors and preparation of the safety analysis report (35-G1), 2) Safety in the utilization and modification of research reactors (35-G2), 3) Commissioning of research reactors (NS-G-4.1), and 4) Maintenance, periodic testing and inspection of research reactors (NS-G-4.2). There 5 draft safety standards: 1) Operational limits and conditions and operating procedures for research reactors (DS261), 2) The operating organization and the recruitment, training and qualification of personnel for research reactors (DS325), 3) Radiation protection and radioactive waste management in the design and operation of research reactors (DS340), 4) Core management and fuel handling at research reactors (DS350), and 5) Grading the application of safety requirements for research reactors (DS351). There are 2 planned safety standards, one concerning the ageing management for research reactor and the second deals with the control and instrumentation of research reactors

  16. General Atomic's radioactive gas recovery system

    International Nuclear Information System (INIS)

    Mahn, J.A.; Perry, C.A.

    1975-01-01

    General Atomic Company has developed a Radioactive Gas Recovery System for the HTGR which separates, for purposes of retention, the radioactive components from the non-radioactive reactor plant waste gases. This provides the capability for reducing to an insignificant level the amount of radioactivity released from the gas waste system to the atmosphere--a most significant improvement in reducing total activity release to the environment. (U.S.)

  17. Reference design for the standard mirror hybrid reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bender, D.J.; Fink, J.H.; Galloway, T.R.; Kastenberg, W.E.; Lee, J.D.; Devoto, R.S.; Neef, W.S. Jr.; Schultz, K.R.; Culver, D.W.; Rao, S.B.; Rao, S.R.

    1978-05-22

    This report describes the results of a two-year study by Lawrence Livermore Laboratory and General Atomic Co. to develop a conceptual design for the standard (minimum-B) mirror hybrid reactor. The reactor parameters have been chosen to minimize the cost of producing nuclear fuel (/sup 239/Pu) for consumption in fission power reactors (light water reactors). The deuterium-tritium plasma produces approximately 400 MW of fusion power with a plasma Q of 0.64. The fast-fission blanket, which is fueled with depleted uranium and lithium, generates sufficient tritium to run the reactor, has a blanket energy multiplication of M = 10.4, and has a net fissile breeding ratio of Pu/n = 1.51. The reactor has a net electrical output of 600 MWe, a fissile production of 2000 kg of plutonium per year (at a capacity factor of 0.74), and a net plant efficiency of 0.18. The plasma-containment field is generated by a Yin-Yang magnet using NbTi superconductor, and the neutral beam system uses positive-ion acceleration with beam direct conversion. The spherical blanket is based on gas-cooled fast reactor technology. The fusion components, blanket, and primary heat-transfer loop components are all contained within a prestressed-concrete reactor vessel, which provides magnet restraint and supports the primary heat-transfer loop and the blanket.

  18. Reference design for the standard mirror hybrid reactor

    International Nuclear Information System (INIS)

    Bender, D.J.; Fink, J.H.; Galloway, T.R.; Kastenberg, W.E.; Lee, J.D.; Devoto, R.S.; Neef, W.S. Jr.; Schultz, K.R.; Culver, D.W.; Rao, S.B.; Rao, S.R.

    1978-01-01

    This report describes the results of a two-year study by Lawrence Livermore Laboratory and General Atomic Co. to develop a conceptual design for the standard (minimum-B) mirror hybrid reactor. The reactor parameters have been chosen to minimize the cost of producing nuclear fuel ( 239 Pu) for consumption in fission power reactors (light water reactors). The deuterium-tritium plasma produces approximately 400 MW of fusion power with a plasma Q of 0.64. The fast-fission blanket, which is fueled with depleted uranium and lithium, generates sufficient tritium to run the reactor, has a blanket energy multiplication of M = 10.4, and has a net fissile breeding ratio of Pu/n = 1.51. The reactor has a net electrical output of 600 MWe, a fissile production of 2000 kg of plutonium per year (at a capacity factor of 0.74), and a net plant efficiency of 0.18. The plasma-containment field is generated by a Yin-Yang magnet using NbTi superconductor, and the neutral beam system uses positive-ion acceleration with beam direct conversion. The spherical blanket is based on gas-cooled fast reactor technology. The fusion components, blanket, and primary heat-transfer loop components are all contained within a prestressed-concrete reactor vessel, which provides magnet restraint and supports the primary heat-transfer loop and the blanket

  19. The contract of design of an atomic reactor system in Yonggwang - 5, 6 nuclear power plant

    International Nuclear Information System (INIS)

    1995-05-01

    This is a contract of design of an atomic reactor system in Yonggwang 5, 6 nuclear power plant. It has the general contract condition. In the appendix, it indicates the detail regulations between two parties which are the coverage of division on the responsibility, schedule of the delivery, standard of the technology, guarantee, drawing and paper support of the Korea Electric Power Corporation, support of technology drill, test, regulations of code and standard and list of items and prices.

  20. Generalities about nuclear reactors

    International Nuclear Information System (INIS)

    Jaouen, C.; Beroux, P.

    2012-01-01

    From Zoe, the first nuclear reactor, till the current EPR, the French nuclear industry has always advanced by profiting from the feedback from dozens of years of experience and operations, in particular by drawing lessons from the most significant events in its history, such as the Fukushima accident. The new generations of reactors must improve safety and economic performance so that the industry maintain its legitimacy and its share in the production of electricity. This article draws the history of nuclear power in France, gives a brief description of the pressurized water reactor design, lists the technical features of the different versions of PWR that operate in France and compares them with other types of reactors. The feedback experience concerning safety, learnt from the major nuclear accidents Three Miles Island (1979), Chernobyl (1986) and Fukushima (2011) is also detailed. Today there are 26 third generation reactors being built in the world: 4 EPR (1 in Finland, 1 in France and 2 in China); 2 VVER-1200 in Russia, 8 AP-1000 (4 in China and 4 in the Usa), 8 APR-1400 (4 in Korea and 4 in UAE), and 4 ABWR (2 in Japan and 2 in Taiwan)

  1. TRIGA International - History of Training Research Isotope production General Atomics

    International Nuclear Information System (INIS)

    2008-01-01

    TRIGA conceived at GA in 1956 by a distinguished group of scientists including Edward Teller and Freeman Dyson. First TRIGA reactor Mk-1 was commissioned on 3 may 1958 at G.A. Characteristic feature of TRIGA reactors is inherent safety: Sitting can be confinement or conventional building. TRIGA reactors are the most prevalent in the world: 67 reactors in 24 countries. Steady state powers up to 14 MWt, pulsing up to 22,000 MWt. To enlarge the scope of its manufactured products, CERCA engaged in a Joint Venture with General Atomics, and in July 1995 a new Company was founded: TRIGA INTERNATIONAL SAS (50% GA, 50% CERCA; Head Office: Paris (France); Sales offices: GA San Diego (Ca, USA) and CERCA Lyon (France); Manufacturing plant: CERCA Romans. General Atomics ID: founded in 1955 at San Diego, California, by General Dynamics; status: Privately held corporation; owners: Neal and Linden Blue; business: High technology research, design, manufacturing, and production for industry and Government in the U.S. and overseas; locations: U.S., Germany, Japan, Australia, Thailand, Morocco; employees: 5,000. TRIGA's ID: CERCA is a subsidiary of AREVA, born in November 05, 1957. Activities: fuel manufacture for research reactor, equipment and components for high-energy physics, radioactive sources and reference sources; plants locations: Romans and Pierrelatte (France); total strength: 180. Since the last five years TRIGA has manufactured and delivered more than 800 fuel elements with a door to door service. TRIGA International has the experience to manufacture all types of TRIGA fuel: standard fuel elements, instrumented fuel elements, fuel followed control rods, geometry: 37.3 mm (1.47 in.), 35.8 mm (1.4 in), 13 mm (0.5 in), chemical Composition: U w% 8.5, 12, 20, 30 and 45 w/o, erbium and no erbium. TRIGA International is on INL's approved vendor list (ISO 9000/NQA) and is ready to meet any TRIGA fuel needs either in the US or worldwide

  2. Scram device for atomic reactors

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  3. Advanced fuel technologies at General Atomics

    International Nuclear Information System (INIS)

    Back, Christina A.

    2013-01-01

    General Atomics (GA) has made significant contributions since its founding in the 1950's to develop nuclear power for peaceful means. With the conception and construction of the TRIGA reactors and research on TRISO particles, GA has long recognised the importance of 'accident-tolerant' materials. Before the accident at Fukushima Daiichi, GA had already initiated work on silicon carbide (SiC) and SiC-related technologies for application in nuclear reactors. At that time, the work was initiated in support of the GA advanced gas-cooled fast reactor concept called the Energy Multiplier Module, EM2. This work continues, however, the reasons that make SiC materials attractive for fast reactor concepts also make them attractive for advanced light water reactors. These include superior performance over zircaloy for high-temperature strength, especially above 1500 deg. C, and significantly reduced hydrogen production in accident scenarios. The current focus on 'accident-tolerant' components is to develop cladding made of silicon carbide fiber and silicon carbide matrix, SiC-SiC composites. The goal for this work is to produce a cladding that provides strength and impermeability to meet reactor performance and safety requirements. To date, GA has examined the trade-offs between processing time and infiltration uniformity to reduce fabrication time, fabricated cylindrical prototypes, and refined material properties for fracture toughness, impermeability, and thermal conductivity. Generally, the GA programme is developing innovative fuel elements that employ both high density uranium-bearing fuels that enable longer lifetime with higher burn-up, and claddings that are more resistant to neutron damage. In addition to fabrication, significant effort is devoted to measuring the critical parameters, such as thermal conductivity, mechanical strength and component performance at reactor-relevant operational conditions, using a mix of commercial equipment

  4. The national standards program for research reactors

    International Nuclear Information System (INIS)

    Whittemore, W.L.

    1977-01-01

    In 1970 a standards committee called ANS-15 was established by the American Nuclear Society (ANS) to prepare appropriate standards for research reactors. In addition, ANS acts as Secretariat for a national standards committee N17 which is responsible to the American National Standards Institute (ANSI) for the national consensus efforts for standards related to research reactors. To date ANS-15 has completed or is working on 14 standards covering all aspects of the operation of research reactors. Of the 11 research reactor standards submitted to the ANSI N17 Committee since its inception, six have been issued as National standards, and the remaining are still in the process of review. (author)

  5. Atomic frequency-time-length standards

    International Nuclear Information System (INIS)

    Gheorghiu, O.C.; Mandache, C.

    1987-01-01

    The principles of operative of atomic frequency-time-length standards and their principle characteristics are described. The role of quartz crystal oscillators which are sloved to active or passive standards is presented. (authors)

  6. Reactor Section standard analytical methods. Part 1

    Energy Technology Data Exchange (ETDEWEB)

    Sowden, D.

    1954-07-01

    the Standard Analytical Methods manual was prepared for the purpose of consolidating and standardizing all current analytical methods and procedures used in the Reactor Section for routine chemical analyses. All procedures are established in accordance with accepted practice and the general analytical methods specified by the Engineering Department. These procedures are specifically adapted to the requirements of the water treatment process and related operations. The methods included in this manual are organized alphabetically within the following five sections which correspond to the various phases of the analytical control program in which these analyses are to be used: water analyses, essential material analyses, cotton plug analyses boiler water analyses, and miscellaneous control analyses.

  7. Atomic power engineering under falsified safety standards

    International Nuclear Information System (INIS)

    Ackerman, A.J.

    1974-01-01

    In July 1970 the United States Department of Justice accused the American Society of Mechanical Engineers (ASME) of violating the Sherman Antitrust Act and of acting in restraint of trade by restricting the ASME Certificate of Authorization and the use of the Code Symbol Stamps to boilers and pressure vessels manufactured in the United States and Canada. During the succeeding two years attorneys for the parties in the case formulated a Consent Decree without a public confrontation in the Court. Furthermore, the membership of ASME was kept uninformed until October of 1972, after the Consent and Final Judgment had become effective and new procedures had been developed for allowing foreign manufacturers to apply the ASME Code Symbol Stamps to their products. As a consequence, a breakdown in engineered safety standards has been sanctioned and this is undermining the engineering profession's overriding reponsibility to protect the public health and safety. This breakdown of professional responsibility is especially serious in the new technology of atomic power. American insurance companies, which have traditionally written 100% insurance coverage for property damage and third party liability against explosions of high pressure steam boilers bearing the ASME Code Stamp, have refused to write such insurance coverage on nuclear reactors. In the author's opinion there is evidence that the Consent was formulated under collusive proceedings and he calls on the members and the Council of ASME to appeal for dismissal of the Consent Decree. 24 refs

  8. Standards for safe operation of research reactors

    International Nuclear Information System (INIS)

    1996-01-01

    The safety of research reactors is based on many factors such as suitable choice of location, design and construction according to the international standards, it also depends on well trained and qualified operational staff. These standards determine the responsibilities of all who are concerned with the research reactors safe operation, and who are responsible of all related activities in all the administrative and technical stages in a way that insures the safe operation of the reactor

  9. General Atomics Sciences Education Foundation Outreach Programs

    Science.gov (United States)

    Winter, Patricia S.

    1997-11-01

    Scientific literacy for all students is a national goal. The General Atomics (GA) Foundation Outreach Program is committed to playing a major role in enhancing pre-college education in science, engineering and new technologies. GA has received wide recognition for its Sciences Education Program, a volunteer effort of GA employees and San Diego science teachers. GA teacher/scientist teams have developed inquiry-based education modules and associated workshops based on areas of core competency at GA: Fusion -- Energy of the Stars; Explorations in Materials Science; Portrait of an Atom; DNA Technology. [http://www.sci-ed-ga.org]. Workshops [teachers receive printed materials and laboratory kits for ``hands-on" modules] have been presented for 700+ teachers from 200+ area schools. Additional workshops include: University of Denver for Denver Public Schools; National Educators Workshop; Standard Experiments in Engineering Materials; Update '96 in Los Alamos; Newspapers in Education Workshop (LA Times); American Chemical Society Regional/National meetings, and California Science Teachers Association Conference. Other outreach includes High School Science Day, school partnerships, teacher and student mentoring and the San Diego Science Alliance [http://www.sdsa.org].

  10. Use of research reactors in Soviet atomic centres

    International Nuclear Information System (INIS)

    1964-01-01

    The manner of controlling and directing research reactors in the USSR was described in October at the IAEA seminar for atomic energy administrators by Dr. U. V. Archangelski, Department Chief, State Committee for Utilization of Atomic Energy, USSR. He also enumerated the research reactors in operation. In addition to the portions of the paper which are quoted below, he gave details of the scientific work being carried out in these reactors.

  11. Draft of standard for graphite core components in high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Shibata, Taiju; Sawa, Kazuhiro; Eto, Motokuni; Kunimoto, Eiji; Shiozawa, Shusaku; Oku, Tatsuo; Maruyama, Tadashi

    2010-01-01

    For the design of the graphite components in the High Temperature Engineering Test Reactor (HTTR), the graphite structural design code for the HTTR etc. were applied. However, general standard systems for the High Temperature Gas-cooled Reactor (HTGR) have not been established yet. The authors had studied on the technical issues which is necessary for the establishment of a general standard system for the graphite components in the HTGR. The results of the study were documented and discussed at a 'Special committee on research on preparation for codes for graphite components in HTGR' at Atomic Energy Society of Japan (AESJ). As a result, 'Draft of Standard for Graphite Core Components in High Temperature Gas-cooled Reactor.' was established. In the draft standard, the graphite components are classified three categories (A, B and C) in the standpoints of safety functions and possibility of replacement. For the components in the each class, design standard, material and product standards, and in-service inspection and maintenance standard are determined. As an appendix of the design standard, the graphical expressions of material property data of 1G-110 graphite as a function of fast neutron fluence are expressed. The graphical expressions were determined through the interpolation and extrapolation of the irradiated data. (author)

  12. Impact of proposed research reactor standards on reactor operation

    Energy Technology Data Exchange (ETDEWEB)

    Ringle, J C; Johnson, A G; Anderson, T V [Oregon State University (United States)

    1974-07-01

    A Standards Committee on Operation of Research Reactors, (ANS-15), sponsored by the American Nuclear Society, was organized in June 1971. Its purpose is to develop, prepare, and maintain standards for the design, construction, operation, maintenance, and decommissioning of nuclear reactors intended for research and training. Of the 15 original members, six were directly associated with operating TRIGA facilities. This committee developed a standard for the Development of Technical Specifications for Research Reactors (ANS-15.1), the revised draft of which was submitted to ANSI for review in May of 1973. The Committee then identified 10 other critical areas for standards development. Nine of these, along with ANS-15.1, are of direct interest to TRIGA owners and operators. The Committee was divided into subcommittees to work on these areas. These nine areas involve proposed standards for research reactors concerning: 1. Records and Reports (ANS-15.3) 2. Selection and Training of Personnel (ANS-15.4) 3. Effluent Monitoring (ANS-15.5) 4. Review of Experiments (ANS-15.6) 5. Siting (ANS-15.7) 6. Quality Assurance Program Guidance and Requirements (ANS-15.8) 7. Restrictions on Radioactive Effluents (ANS-15.9) 8. Decommissioning (ANS-15.10) 9. Radiological Control and Safety (ANS-15.11). The present status of each of these standards will be presented, along with their potential impact on TRIGA reactor operation. (author)

  13. Impact of proposed research reactor standards on reactor operation

    International Nuclear Information System (INIS)

    Ringle, J.C.; Johnson, A.G.; Anderson, T.V.

    1974-01-01

    A Standards Committee on Operation of Research Reactors, (ANS-15), sponsored by the American Nuclear Society, was organized in June 1971. Its purpose is to develop, prepare, and maintain standards for the design, construction, operation, maintenance, and decommissioning of nuclear reactors intended for research and training. Of the 15 original members, six were directly associated with operating TRIGA facilities. This committee developed a standard for the Development of Technical Specifications for Research Reactors (ANS-15.1), the revised draft of which was submitted to ANSI for review in May of 1973. The Committee then identified 10 other critical areas for standards development. Nine of these, along with ANS-15.1, are of direct interest to TRIGA owners and operators. The Committee was divided into subcommittees to work on these areas. These nine areas involve proposed standards for research reactors concerning: 1. Records and Reports (ANS-15.3) 2. Selection and Training of Personnel (ANS-15.4) 3. Effluent Monitoring (ANS-15.5) 4. Review of Experiments (ANS-15.6) 5. Siting (ANS-15.7) 6. Quality Assurance Program Guidance and Requirements (ANS-15.8) 7. Restrictions on Radioactive Effluents (ANS-15.9) 8. Decommissioning (ANS-15.10) 9. Radiological Control and Safety (ANS-15.11). The present status of each of these standards will be presented, along with their potential impact on TRIGA reactor operation. (author)

  14. JAERI thermal reactor standard code system for reactor design and analysis SRAC

    International Nuclear Information System (INIS)

    Tsuchihashi, Keichiro

    1985-01-01

    SRAC, JAERI thermal reactor standard code system for reactor design and analysis, developed in Japan Atomic Energy Research Institute, is for all types of thermal neutron nuclear design and analysis. The code system has undergone extensive verifications to confirm its functions, and has been used in core modification of the research reactor, detailed design of the multi-purpose high temperature gas reactor and analysis of the experiment with a critical assembly. In nuclear calculation with the code system, multi-group lattice calculation is first made with the libraries. Then, with the resultant homogeneous equivalent group constants, reactor core calculation is made. Described are the following: purpose and development of the code system, functions of the SRAC system, bench mark tests and usage state and future development. (Mori, K.)

  15. Standards for reference reactor physics measurements

    International Nuclear Information System (INIS)

    Harris, D.R.; Cokinos, D.M.; Uotinen, V.

    1990-01-01

    Reactor physics analysis methods require experimental testing and confirmation over the range of practical reactor configurations and states. This range is somewhat limited by practical fuel types such as actinide oxides or carbides enclosed in metal cladding. On the other hand, this range continues to broaden because of the trend of using higher enrichment, if only slightly enriched, electric utility fuel. The need for experimental testing of the reactor physics analysis methods arises in part because of the continual broadening of the range of core designs, and in part because of the nature of the analysis methods. Reactor physics analyses are directed primarily at the determination of core reactivities and reaction rates, the former largely for reasons of reactor control, and the latter largely to ensure that material limitations are not violated. Errors in these analyses can be regarded as being from numerics, from the data base, and from human factors. For numerical, data base, and human factor reasons, then, it is prudent and customary to qualify reactor physical analysis methods against experiments. These experiments can be treated as being at low power or at high power, and each of these types is subject to an American National Standards Institute standard. The purpose of these standards is to aid in improving and maintaining adequate quality in reactor physics methods, and it is from this point of view that the standards are examined here

  16. Education for university students, high school teachers and the general public using the Kinki University Reactor

    International Nuclear Information System (INIS)

    Tsuruta, T.

    2007-01-01

    Atomic Energy Research Institute of Kinki University is equipped with a nuclear reactor which is called UTR-KINKI. UTR is the abbreviation for University Teaching and Research Reactor. The reactor is the first one installed in Japanese universities. Though the reactor is owned and operated by Kinki University, its use is widely open to scientists and students from other universities and research institutions. The reactor is made the best of teaching instrument for the training of high school teachers. In addition, the reactor is utilized for general public education concerning atomic energy. (author)

  17. Accelerators and nuclear reactors as tools in hot atom chemistry

    International Nuclear Information System (INIS)

    Lindner, L.

    1975-01-01

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

  18. International standardization of safety requirements for fast reactors

    International Nuclear Information System (INIS)

    2011-06-01

    Japan Atomic Energy Agency (JAEA) is conducting the FaCT (Fast Reactor Cycle Technology Development) project in cooperation with Japan Atomic Power Company (JAPC) and Mitsubishi FBR systems inc. (MFBR), where an advanced loop-type fast reactor named JSFR (Japan Sodium-cooled Fast Reactor) is being developed. It is important to develop software technologies (a safety guideline, safety design criteria, safety design standards etc.) of FBRs as well as hardware ones (a reactor plant itself) in order to address prospective worldwide utilization of FBR technology. Therefore, it is expected to establish a rational safety guideline applicable to the JSFR and harmonized with national nuclear-safety regulations as well, including Japan, the United States and the European Union. This report presents domestic and international status of safety guideline development for sodium-cooled fast reactors (SFRs), results of comparative study for safety requirements provided in existing documents and a proposal for safety requirements of future SFRs with a roadmap for their refinement and worldwide utilization. (author)

  19. Research reactor standards and their impact on the TRIGA reactor community

    International Nuclear Information System (INIS)

    Richards, W.J.

    1980-01-01

    The American Nuclear Society has established a standards committee devoted to writing standards for research reactors. This committee was formed in 1971 and has since that time written over 15 standards that cover all aspects of research reactor operation. The committee has representation from virtually every group concerned with research reactors and their operation. This organization includes University reactors, National laboratory reactors, Nuclear Regulatory commission, Department of Energy and private nuclear companies and insurers. Since its beginning the committee has developed standards in the following areas: Standard for the development of technical specifications for research reactors; Quality control for plate-type uranium-aluminium fuel elements; Records and reports for research reactors; Selection and training of personnel for research reactors; Review of experiments for research reactors; Research reactor site evaluation; Quality assurance program requirements for research reactors; Decommissioning of research reactors; Radiological control at research reactor facilities; Design objectives for and monitoring of systems controlling research reactor effluents; Physical security for research reactor facilities; Criteria for the reactor safety systems of research reactors; Emergency planning for research reactors; Fire protection program requirements for research reactors; Standard for administrative controls for research reactors. Besides writing the above standards, the committee is very active in using communications with the nuclear regulatory commission on proposed rules or positions which will affect the research reactor community

  20. Standard mirror fusion reactor design study

    International Nuclear Information System (INIS)

    Moir, R.W.

    1978-01-01

    This report covers the work of the Magnetic Fusion Energy Division's reactor study group during FY 1976 on the standard mirror reactor. The ''standard'' mirror reactor is characterized as a steady state, neutral beam sustained, D-T fusioning plasma confined by a Yin-Yang magnetic mirror field. The physics parameters are obtained from the same physics model that explains the 2XIIB experiment. The model assumes that the drift cyclotron loss cone mode occurs on the boundary of the plasma, and that it is stabilized by warm plasma with negligible energy investment. The result of the study was a workable mirror fusion power plant, steady-state blanket removal made relatively simple by open-ended geometry, and no impurity problem due to the positive plasma potential. The Q (fusion power/injected beam power) turns out to be only 1.1 because of loss out the ends from Coulomb collisions, i.e., classical losses. This low Q resulted in 77% of the gross electrical power being used to power the injectors, thereby causing the net power cost to be high. The low Q stimulated an intensive search for Q-enhancement concepts, resulting in the LLL reactor design effort turning to the field reversal mirror and the tandem mirror, each having Q of order 5

  1. General Atomic's superconducting toroidal field coil concept

    International Nuclear Information System (INIS)

    Alcorn, J.; Purcell, J.

    1978-01-01

    General Atomic's concept for a superconducting toroidal field coil is presented. The concept is generic for large tokamak devices, while a specific design is indicated for a 3.8 meter (major radius) ignition/burn machine. The concept utilizes bath cooled NbTi conductor to generate a peak field of 10 tesla at 4.2 K. The design is simple and straightforward, requires a minimum of developmental effort, and draws extensively upon the perspective of past experience in the design and construction of large superconducting magnets for high energy physics. Thus, the primary emphasis is upon economy, reliability, and expeditious construction scheduling. (author)

  2. Atomic Energy Law and technical standardization

    International Nuclear Information System (INIS)

    Vieweg, K.

    1982-01-01

    The dissertation given a representation of the origination of the Nuclear Technology Committee (KTA), its responsibilities and organization, and of the procedure of establishing and amending the KTA rules in terms of factural structure, and deals with the legal quality of the KTA and its legal organizational form. In addition, it describes the legal status of the KTA members, the legal quality of the KTA office and of the bodies participating in the establishment of rules. In conclusion, it deals with the legal quality of KTA rules and their legal significances within administrative and court procedures and proceedings bearing on atomic law. The author qualifies the KTA rules as simple administrative acts in the form of abstract-general anticipated expert's opinions. (HSCH) [de

  3. Research and development issues for fast reactor structural design standard (FDS)

    International Nuclear Information System (INIS)

    Kasahara, Naoto; Ando, Masanori; Morishita, Masaki

    2003-01-01

    For realization of safe and economical fast reactor (FR) plants, Japan Nuclear Cycle Development Institute (JNC) and Japan Atomic Power Company (JAPC) are cooperating on 'Feasibility Study on Commercialized FR Cycle Systems'. To certify the design concepts and validate their structural integrity, the research and development of 'Fast Reactor Structural Design Standard (FDS)' is recognized as an essential theme. FDS considers general characteristics of FRs and design needs for their rationalization. Three main subjects were settled in research and development issues of FDS. One is rationalization of failure criteria' taking characteristic design conditions into account. Next is development of 'a guideline on inelastic analysis for design' in order to predict elastic plastic and creep behaviours of high temperature components. Furthermore, efforts are being made toward preparing a guideline on thermal loads modeling' for FR component design where thermal loads are dominant. (author)

  4. A generalized perturbation program for CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Do Heon; Kim, Jong Kyung [Hanyang University, Seoul (Korea, Republic of); Choi, Hang Bok; Roh, Gyu Hong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Yang, Won Sik [Chosun University, Kwangju (Korea, Republic of)

    1999-12-31

    A generalized perturbation program has been developed for the purpose of estimating zonal power variation of a CANDU reactor upon refueling operation. The forward and adjoint calculation modules of RFSP code were used to construct the generalized perturbation program. The numerical algorithm for the generalized adjoint flux calculation was verified by comparing the zone power estimates upon refueling with those of forward calculation. It was, however, noticed that the truncation error from the iteration process of the generalized adjoint flux is not negligible. 2 refs., 1 figs., 1 tab. (Author)

  5. A generalized perturbation program for CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Do Heon; Kim, Jong Kyung [Hanyang University, Seoul (Korea, Republic of); Choi, Hang Bok; Roh, Gyu Hong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Yang, Won Sik [Chosun University, Kwangju (Korea, Republic of)

    1998-12-31

    A generalized perturbation program has been developed for the purpose of estimating zonal power variation of a CANDU reactor upon refueling operation. The forward and adjoint calculation modules of RFSP code were used to construct the generalized perturbation program. The numerical algorithm for the generalized adjoint flux calculation was verified by comparing the zone power estimates upon refueling with those of forward calculation. It was, however, noticed that the truncation error from the iteration process of the generalized adjoint flux is not negligible. 2 refs., 1 figs., 1 tab. (Author)

  6. Internal structure of reactor building for Madras Atomic Power Project

    International Nuclear Information System (INIS)

    Pandit, D.P.

    1975-01-01

    The structural configuration and analysis of structural elements of the internal structure of reactor building for the Madras Atomic Power Project has been presented. Two methods of analysis of the internal structure, viz. Equivalent Plane Frame and Finite Element Method, are explained and compared with the use of bending moments obtained. (author)

  7. Generalized Pauli constraints in small atoms

    Science.gov (United States)

    Schilling, Christian; Altunbulak, Murat; Knecht, Stefan; Lopes, Alexandre; Whitfield, James D.; Christandl, Matthias; Gross, David; Reiher, Markus

    2018-05-01

    The natural occupation numbers of fermionic systems are subject to nontrivial constraints, which include and extend the original Pauli principle. A recent mathematical breakthrough has clarified their mathematical structure and has opened up the possibility of a systematic analysis. Early investigations have found evidence that these constraints are exactly saturated in several physically relevant systems, e.g., in a certain electronic state of the beryllium atom. It has been suggested that, in such cases, the constraints, rather than the details of the Hamiltonian, dictate the system's qualitative behavior. Here, we revisit this question with state-of-the-art numerical methods for small atoms. We find that the constraints are, in fact, not exactly saturated, but that they lie much closer to the surface defined by the constraints than the geometry of the problem would suggest. While the results seem incompatible with the statement that the generalized Pauli constraints drive the behavior of these systems, they suggest that the qualitatively correct wave-function expansions can in some systems already be obtained on the basis of a limited number of Slater determinants, which is in line with numerical evidence from quantum chemistry.

  8. STANDARD ATOMIC WEIGHT VALUES FOR THE MONONUCLIDIC ELEMENTS - 2001

    International Nuclear Information System (INIS)

    HOLDEN, N.E.

    2001-01-01

    Atomic Mass Evaluations have had a major impact on the values of the atomic weights for the twenty mononuclidic elements plus two elements, Thorium and Protactinium, which have no stable nuclides but a characteristic terrestrial isotopic composition. This paper reviews the history of the atomic weight values of these elements in the years, since the reference mass standard changed from 16 O to 12 C. There is a problem for Thorium, which is considered to have an abundance value of 100%, but is not treated as such in the Standard Atomic Weights' Table. Recommendations for handling the Standard Atomic Weight values for 2001 are presented

  9. General philosophy of safety standards

    International Nuclear Information System (INIS)

    Dunster, H.J.

    1987-01-01

    Safety standards should be related to the form and magnitude of the risk they aim to limit. Because of the lack of direct information at the exposure levels experienced, radiation protection standards have to be based on risk assumptions that, while plausible, are not proven. The pressure for standards has come as much from public perceptions and fears as from the reality of the risk. (author)

  10. Innovation and reliability of atomic standards for PTTI applications

    Science.gov (United States)

    Kern, R.

    1981-01-01

    Innovation and reliability in hyperfine frequency standards and clock systems are discussed. Hyperfine standards are defined as those precision frequency sources and clocks which use a hyperfine atomic transition for frequency control and which have realized significant commercial production and acceptance (cesium, hydrogen, and rubidium atoms). References to other systems such as thallium and ammonia are excluded since these atomic standards have not been commercially exploited in this country.

  11. IAEA safety standards and approach to safety of advanced reactors

    International Nuclear Information System (INIS)

    Gasparini, M.

    2004-01-01

    The paper presents an overview of the IAEA safety standards including their overall structure and purpose. A detailed presentation is devoted to the general approach to safety that is embodied in the current safety requirements for the design of nuclear power plants. A safety approach is proposed for the future. This approach can be used as reference for a safe design, for safety assessment and for the preparation of the safety requirements. The method proposes an integration of deterministic and risk informed concepts in the general frame of a generalized concept of safety goals and defence in depth. This methodology may provide a useful tool for the preparation of safety requirements for the design and operation of any kind of reactor including small and medium sized reactors with innovative safety features.(author)

  12. Atomization of U3Si2 for research reactor fuel

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  13. Investigations of Laser Pumped Gas Cell Atomic Frequency Standard

    National Research Council Canada - National Science Library

    Volk, C. H; Camparo, J. C; Frueholz, R. P

    1981-01-01

    Recently it has been suggested that the performance characteristics of a rubidium gas cell atomic frequency standard might be improved by replacing the standard rubidium discharge lamp with a single mode laser diode...

  14. The processing and management of wastes from atomic reactors

    International Nuclear Information System (INIS)

    Cerre, P.; Mestre, E.; Bourdrez, J.

    1964-01-01

    The policy concerning radioactive wastes studied by all Atomic Centres has led to various procedures which, while apparently numerous, come under a few standard headings. Whether the wastes are in the liquid or solid state their management depends on their physical and chemical nature. The procedure adopted is governed by three general principles: - determination of the most economical means possible of storage and processing by volume reduction; - conversion to a solid compact form; - complete acceptance of the accepted standards at all places and all times. In this communication all the standard solutions adopted and used by the various Centres of the Commissariat a l'Energie Atomique will be examined bearing in mind the preceding remarks. Particular mention will be made of the following: - For liquids, physical, chemical and physico-chemical processing - For solids, decontamination, volume reduction and long-term conditioning techniques. The different procedures for collecting and storing solid wastes before and after processing are also discussed. The paper ends with a brief review of the studies, both technical and economic, being pursued on this subject. (authors) [fr

  15. Advances in U.S. reactor physics standards

    International Nuclear Information System (INIS)

    Cokinos, Dimitrios

    2008-01-01

    The standards for Reactor Design, widely used in the nuclear industry, provide guidance and criteria for performing and validating a wide range of nuclear reactor calculations and measurements. Advances, over the past decades in reactor technology, nuclear data and infrastructure in the data handling field, led to major improvements in the development and application of reactor physics standards. A wide variety of reactor physics methods and techniques are being used by reactor physicists for the design and analysis of modern reactors. ANSI (American National Standards Institute) reactor physics standards, covering such areas as nuclear data, reactor design, startup testing, decay heat and fast neutron fluence in the pressure vessel, are summarized and discussed. These standards are regularly undergoing review to respond to an evolving nuclear technology and are being successfully used in the U.S and abroad contributing to improvements in reactor design, safe operation and quality assurance. An overview of the overall program of reactor physics standards is presented. New standards currently under development are also discussed. (authors)

  16. Forecast of Standard Atomic Weights for the Mononuclidic Elements - 2011

    International Nuclear Information System (INIS)

    Holden, N.E.

    2011-01-01

    In this short report, I will provide an early warning about potential changes to the standard atomic weight values for the twenty mononuclidic and the so-called pseudo-mononuclidic ( 232 Th and 231 Pa) chemical elements due to the estimated changes in the mass values to be published in the next Atomic Mass Tables within the next two years. There have been many new measurements of atomic masses, since the last published Atomic Mass Table. The Atomic Mass Data Center has released an unpublished version of the present status of the atomic mass values as a private communication. We can not update the Standard Atomic Weight Table at this time based on these unpublished values but we can anticipate how many changes are probably going to be expected in the next few years on the basis of the forthcoming publication of the Atomic Mass Table. I will briefly discuss the procedures that the Atomic Weights Commission used in deriving the recommended Standard Atomic Weight values and their uncertainties from the atomic mass values. I will also discuss some concern raised about a proposed change in the definition of the mole. The definition of the mole is now connected directly to the mass of a 12 C isotope (which is defined as 12 exactly) and to the kilogram. A change in the definition of the mole will probably impact the mass of 12 C.

  17. Brief overview of American Nuclear Society's research reactor standards

    International Nuclear Information System (INIS)

    Richards, Wade J.

    1984-01-01

    The American Nuclear Society (ANS) established the research reactor standards group in 1968. The standards group, known as ANS-15, was established for the purpose of developing, preparing, and maintaining standards for the design, construction, operation, maintenance, and decommissioning of nuclear reactors intended for research and training

  18. Proposed general amendments to the atomic energy control regulations

    International Nuclear Information System (INIS)

    1986-01-01

    Canada's Atomic Energy Control Act defines the powers and responsibilities of the Atomic Energy Control Board (AECB). Among these is to make regulations to control the development, application and use of atomic energy. In these proposed general amendments to the Atomic Energy Control Regulations substantial changes are proposed in the designation of the authority of AECB staff, exemptions from licensing, international safeguards, duties of licensees and atomic radiation workers, security of information, and provision for hearings. The scope of the control of atomic energy has been redefined as relating to matters of health, safety, security, international safeguards, and the protection of the environment

  19. An overview of reactor physics standards: Past, present and future

    International Nuclear Information System (INIS)

    Cokinos, D.M.

    1992-07-01

    This report discusses for determining key static reactor physics parameters which have been developed by groups of experts (working groups) under the aegis of ANS-19, the ANS Reactor Physics Standards Committee. Following a series of sequential reviews, augmented by feedback from potential users, a proposed standard is brought into final form by the working group before it is adopted as a formal standard by the American National Standards Institute (ANSI); Reactor Physics standards are intended to provide guidance in the performance and qualification of complex sequences of reactor calculations and/or measurements and are regularly reviewed for possible updates and/or revisions. The reactor physics standards developed to date are listed and standards now being developed by the respective working groups are also provided

  20. Nuclear powerplant standardization: light water reactors. Volume 2. Appendixes

    International Nuclear Information System (INIS)

    1981-06-01

    This volume contains working papers written for OTA to assist in preparation of the report, NUCLEAR POWERPLANT STANDARDIZATION: LIGHT WATER REACTORS. Included in the appendixes are the following: the current state of standardization, an application of the principles of the Naval Reactors Program to commercial reactors; the NRC and standardization, impacts of nuclear powerplant standardization on public health and safety, descriptions of current control room designs and Duke Power's letter, Admiral Rickover's testimony, a history of standardization in the NRC, and details on the impact of standardization on public health and safety

  1. Clarifying atomic weights: A 2016 four-figure table of standard and conventional atomic weights

    Science.gov (United States)

    Coplen, Tyler B.; Meyers, Fabienne; Holden, Norman E.

    2017-01-01

    To indicate that atomic weights of many elements are not constants of nature, in 2009 and 2011 the Commission on Isotopic Abundances and Atomic Weights (CIAAW) of the International Union of Pure and Applied Chemistry (IUPAC) replaced single-value standard atomic weight values with atomic weight intervals for 12 elements (hydrogen, lithium, boron, carbon, nitrogen, oxygen, magnesium, silicon, sulfur, chlorine, bromine, and thallium); for example, the standard atomic weight of nitrogen became the interval [14.00643, 14.00728]. CIAAW recognized that some users of atomic weight data only need representative values for these 12 elements, such as for trade and commerce. For this purpose, CIAAW provided conventional atomic weight values, such as 14.007 for nitrogen, and these values can serve in education when a single representative value is needed, such as for molecular weight calculations. Because atomic weight values abridged to four figures are preferred by many educational users and are no longer provided by CIAAW as of 2015, we provide a table containing both standard atomic weight values and conventional atomic weight values abridged to four figures for the chemical elements. A retrospective review of changes in four-digit atomic weights since 1961 indicates that changes in these values are due to more accurate measurements over time or to the recognition of the impact of natural isotopic fractionation in normal terrestrial materials upon atomic weight values of many elements. Use of the unit “u” (unified atomic mass unit on the carbon mass scale) with atomic weight is incorrect because the quantity atomic weight is dimensionless, and the unit “amu” (atomic mass unit on the oxygen scale) is an obsolete term: Both should be avoided.

  2. Joint General Atomic-TAERF fusion program

    Energy Technology Data Exchange (ETDEWEB)

    Kerst, D W [John Jay Hopkins Laboratory for Pure and Applied Science, General Atomic Division of General Dynamics Corporation, San Diego, CA (United States)

    1958-07-01

    The experimental work has consisted of several parts: the study of charge exchange in hydrogen ionic and atomic collisions, the study of some linear pinch discharge systems with high stabilizing axial magnetic fields, developments on a small scale for a large toroidal geometry, and experiments with various diagnostic methods, including electrical, optical, and shock-tube methods. The experiments on atomic collisions have consisted of measurements of cross sections for the ionization, the excitation of Lyman-alpha radiation, and elastic scattering for the case of electron bombardment. In addition, charge-exchange cross sections between deuterons and deuterium atoms have been measured. The calculations of Dalgarno and Yadav, using a perturbed stationary-state approximation are close to the experimental results which show a very large cross section for charge exchange.

  3. Generalized Pauli constraints in small atoms

    DEFF Research Database (Denmark)

    Schilling, Christian; Altunbulak, Murat; Knecht, Stefan

    2018-01-01

    investigations have found evidence that these constraints are exactly saturated in several physically relevant systems, e.g., in a certain electronic state of the beryllium atom. It has been suggested that, in such cases, the constraints, rather than the details of the Hamiltonian, dictate the system......'s qualitative behavior. Here, we revisit this question with state-of-the-art numerical methods for small atoms. We find that the constraints are, in fact, not exactly saturated, but that they lie much closer to the surface defined by the constraints than the geometry of the problem would suggest. While...

  4. Evaluation of the General Atomic codes TAP and RECA for HTGR accident analyses

    International Nuclear Information System (INIS)

    Ball, S.J.; Cleveland, J.C.; Sanders, J.P.

    1978-01-01

    The General Atomic codes TAP (Transient Analysis Program) and RECA (Reactor Emergency Cooling Analysis) are evaluated with respect to their capability for predicting the dynamic behavior of high-temperature gas-cooled reactors (HTGRs) for postulated accident conditions. Several apparent modeling problems are noted, and the susceptibility of the codes to misuse and input errors is discussed. A critique of code verification plans is also included. The several cases where direct comparisons could be made between TAP/RECA calculations and those based on other independently developed codes indicated generally good agreement, thus contributing to the credibility of the codes

  5. Standard interface files and procedures for reactor physics codes. Version IV

    International Nuclear Information System (INIS)

    O'Dell, R.D.

    1977-09-01

    Standards, procedures, and recommendations of the Committee on Computer Code Coordination for promoting the exchange of reactor physics codes are updated to Version IV status. Standards and procedures covering general programming, program structure, standard interface files, and file management and handling subroutines are included

  6. Atomic layer deposition on nanoparticles in a rotary reactor

    Science.gov (United States)

    McCormick, Jarod Alan

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

  7. 5 CFR 2635.202 - General standards.

    Science.gov (United States)

    2010-01-01

    ... 5 Administrative Personnel 3 2010-01-01 2010-01-01 false General standards. 2635.202 Section 2635.202 Administrative Personnel OFFICE OF GOVERNMENT ETHICS GOVERNMENT ETHICS STANDARDS OF ETHICAL... contrary to applicable regulations, policies or guidance relating to the procurement of supplies and...

  8. General remarks on fast neutron reactor physics

    International Nuclear Information System (INIS)

    Barre, J.Y.

    1980-01-01

    The main aspects of fast reactor physics, presented in these lecture notes, are restricted to LMFBR's. The emphasis is placed on the core neutronic balance and the burn-up problems. After a brief description of the power reactor main components and of the fast reactor chronology, the fundamental parameters of the one-group neutronic balance are briefly reviewed. Then the neutronic burn-up problems related to the Pu production and to the doubling time are considered

  9. General aspects of CAREM-25 reactor

    International Nuclear Information System (INIS)

    Delmastro, Dario F.; Gomez, Silvia; Ishida, Viviana; Mazzi, Ruben; Santecchia, Alberto; Gomez de Soler, Susana M.

    2000-01-01

    CAREM project consists on the development and design of an advanced nuclear power plant. In order to verify its innovative features the construction of a prototype is planned. In this paper the main technical characteristics of CAREM-25 prototype reactor are presented. This is a very low power innovative reactor (100 M Wth) conceived with new generation design solutions. Based on an indirect cycle integrated light water reactor using enriched uranium, CAREM has some distinctive features that greatly simplify the reactor and also contribute to a high level of safety: -) Integrated primary system; -) Primary system cooling by natural convection; -) Self pressurization; -) and Passive safety systems. (author)

  10. General Aspects of CAREM-25 Reactor

    International Nuclear Information System (INIS)

    Delmastro, Dario; Gomez, S.; Mazzi, R.; Gomez de Soler, S.; Santecchia, A.; Ishida, V.

    2000-01-01

    CAREM project consists on the development and design of an advanced Nuclear Power Plant. In order to verify its innovative features the construction of a prototype is planned. In this paper the main technical characteristics of CAREM-25 prototype reactor are presented. This is a very low power innovative reactor (100MWth) conceived with new generation design solutions. Based on an indirect cycle integrated light water reactor using enriched uranium, CAREM has some distinctive features that greatly simplify the reactor and also contribute to a high level of safety: integrated primary system, primary system cooling by natural convection, selfpressurization, and passive safety systems

  11. Japan Atomic Energy Research Institute, Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    1981-09-01

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

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

    International Nuclear Information System (INIS)

    Vidovszky, I.

    2009-10-01

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

  13. Developing detection efficiency standards for atom probe tomography

    Science.gov (United States)

    Prosa, Ty J.; Geiser, Brian P.; Lawrence, Dan; Olson, David; Larson, David J.

    2014-08-01

    Atom Probe Tomography (APT) is a near-atomic-scale analytical technique which, due to recent advances in instrumentation and sample preparation techniques, is being used on a variety of 3D applications. Total system detection efficiency is a key parameter for obtaining accurate spatial reconstruction of atomic coordinates from detected ions, but experimental determination of efficiency can be difficult. This work explores new ways to measure total system detection efficiency as well as the specimen characteristics necessary for such measurements. Composite specimens composed of a nickel/chromium multilayer core, National Institute of Standards and Technology Standard Reference Material 2135c, encapsulated with silver, silicon, or nickel were used to demonstrate the suitability of this approach for providing a direct measurement of APT efficiency. Efficiency measurements based on this multilayer encapsulated in nickel are reported.

  14. A standard form for generalized CP transformations

    International Nuclear Information System (INIS)

    Ecker, G.; Grimus, W.; Neufeld, H.

    1987-01-01

    The investigation of general CP transformations leads to transformations of the form U → W T UW with unitary matrices U, W. It is shown that a basis for weak eigenstates can always be chosen such that W T UW has a certain real standard form. (Author)

  15. Japan Atomic Energy Research Institute, Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    1979-09-01

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

  16. Light-water reactor pressure vessel surveillance standards

    International Nuclear Information System (INIS)

    Anon.

    1981-01-01

    The master matrix standard describes a series of standard practices, guides, and methods for the prediction of neutron-induced changes in light-water reactor (LWR) pressure vessel steels throughout a pressure vessel's service life. Some of these are existing American Society for Testing and Materials (ASTM) standards, some are ASTM standards that have been modified, and some are newly proposed ASTM standards. The current (1) scope, (2) areas of application, (3) interrelationships, and (4) status and time table of development, improvement, validation, and calibration for a series of 16 ASTM standards are defined. The standard also includes a discussion of LWR pressure vessel surveillance - justification, requirements, and status of work

  17. General description of advanced heavy water reactor

    International Nuclear Information System (INIS)

    Kakodkar, A.; Sinha, R.K.; Dhawan, M.L.

    1999-01-01

    Advanced Heavy Water Reactor is a boiling light water cooled, heavy water moderated and vertical pressure tube type reactor with its design optimised for utilisation of thorium for power generation. The core consists of (Th-U 233 )O 2 and (Th-Pu)O 2 fuel with a discharge burn up of 20,000 MWd/Te. This reactor incorporates several features to simplify the design, which eliminate certain systems and components. AHWR design is also optimised for easy replaceability of coolant channels, facilitation of in-service inspection and maintenance and ease of erection. The AHWR design also incorporates several passive systems for performing safety-related functions in the event of an accident. In case of LOCA, emergency coolant is injected through 4 accumulators of 260 m 3 capacity directly into the core. Gravity driven water pool of capacity 6000 m 3 serves to cool the core for 3 days without operator's intervention. Core submergence, passive containment isolation and passive containment cooling are the added features in AHWR. The paper describes the various process systems, core and fuel design, primary components and safety concepts of AHWR. Plant layout and technical data are also presented. The conceptual design of the reactor has been completed, and the detailed design and development is scheduled for completion in the year 2002. (author)

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

    International Nuclear Information System (INIS)

    DiMeglio, A.F.

    1987-01-01

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

  19. Standard Master Matrix for Light-Water Reactor Pressure Vessel Surveillance Standards, E706(0)

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2002-01-01

    1.1 This master matrix standard describes a series of standard practices, guides, and methods for the prediction of neutron-induced changes in light-water reactor (LWR) pressure vessel (PV) and support structure steels throughout a pressure vessel's service life (Fig. 1). Some of these are existing ASTM standards, some are ASTM standards that have been modified, and some are proposed ASTM standards. General requirements of content and consistency are discussed in Section 6 . More detailed writers' and users' information, justification, and specific requirements for the nine practices, ten guides, and three methods are provided in Sections 3-5. Referenced documents are discussed in Section 2. The summary-type information that is provided in Sections 3 and 4 is essential for establishing proper understanding and communications between the writers and users of this set of matrix standards. It was extracted from the referenced documents, Section 2 and references (1-106) for use by individual writers and users. 1...

  20. American National Standard: nuclear data sets for reactor design calculations

    International Nuclear Information System (INIS)

    1983-01-01

    This standard identifies and describes the specifications for developing, preparing, and documenting nuclear data sets to be used in reactor design calculations. The specifications include criteria for acceptance of evaluated nuclear data sets, criteria for processing evaluated data and preparation of processed continuous data and averaged data sets, and identification of specific evaluated, processed continuous, and averaged data sets which meet these criteria for specific reactor types

  1. American National Standard nuclear data sets for reactor design calculations

    International Nuclear Information System (INIS)

    Anon.

    1975-01-01

    A standard is presented which identifies and describes the specifications for developing, preparing, and documenting nuclear data sets to be used in reactor design calculations. The specifications include (a) criteria for acceptance of evaluated nuclear data sets, (b) criteria for processing evaluated data and preparation of processed continuous data and averaged data sets, and (c) identification of specific evaluated, processed continuous, and averaged data sets which meet these criteria for specific reactor types

  2. Standard Technical Specifications for Westinghouse pressurized water reactors

    International Nuclear Information System (INIS)

    Virgilio, M.J.

    1980-09-01

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

  3. Standard Technical Specifications for Combustion Engineering Pressurized Water Reactors

    International Nuclear Information System (INIS)

    Vito, D.J.

    1980-12-01

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

  4. Overview of standards subcommittee 8, fissionable materials outside reactors

    International Nuclear Information System (INIS)

    McLaughlin, T.P.

    1996-01-01

    The American Nuclear Society's Standards Subcommittee 8, titled open-quotes Fissionable Materials Outside Reactors,close quotes has worked for the past 35 yr to prepare and promote standards on nuclear criticality safety for the handling, processing, storing, and transportation of fissionable materials outside reactors. The reader is referred to the Transactions of the American Nuclear Society, Vols. 39 (1981) and 64 (1991), for previous papers associated with ANS-8 poster sessions. In addition to discussions on the then-current standards, the reader will find articles on working group efforts that never materialized into standards, such as proposed 8.13, open-quotes Use of the Solid-Angle Method in Nuclear Criticality Safety,close quotes and on applications and critiques of current standards. The paper by McLendon in Vol. 39 is particularly interesting as an overview of the early history of ANS-8 and its standards

  5. International standardization of nuclear reactor designs - the way forward

    International Nuclear Information System (INIS)

    Raetzke, Christian

    2010-01-01

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

  6. Improvement of Frequency Locking Algorithm for Atomic Frequency Standards

    Science.gov (United States)

    Park, Young-Ho; Kang, Hoonsoo; Heyong Lee, Soo; Eon Park, Sang; Lee, Jong Koo; Lee, Ho Seong; Kwon, Taeg Yong

    2010-09-01

    The authors describe a novel method of frequency locking algorithm for atomic frequency standards. The new algorithm for locking the microwave frequency to the Ramsey resonance is compared with the old one that had been employed in the cesium atomic beam frequency standards such as NIST-7 and KRISS-1. Numerical simulations for testing the performance of the algorithm show that the new method has a noise filtering performance superior to the old one by a factor of 1.2 for the flicker signal noise and 1.4 for random-walk signal noise. The new algorithm can readily be used to enhance the frequency stability for a digital servo employing the slow square wave frequency modulation.

  7. Standard irradiation facilities for use in TRIGA reactors

    International Nuclear Information System (INIS)

    Kolbasov, B.N.; Luse, R.A.

    1972-01-01

    The standard neutron irradiation facility (SNIP) was developed under IAEA and FAO co-ordinated research program for the standardization of neutron irradiation facilities for radiobiological research, resulting in the possibility to use fast neutrons from pool-type reactors for radiobiological studies. The studies include irradiation of seeds for crop improvement, of Drosophila for genetic studies, and of microorganisms for developing industrially useful mutants, as well as fundamental studies in radiation biology. The facilities, located in the six pool-type reactors (in Austria, Bulgaria, India, Philippines, Thailand and Taiwan), have been calibrated and utilized to compare the response to fast neutrons of barley seeds (variety Himalaya CI 000620) which were selected as a standard biological monitor by which to estimate neutron fluxes in different reactors. These comparative irradiation studies showed excellent agreement and reproducibility

  8. Portable atomic frequency standard based on coherent population trapping

    Science.gov (United States)

    Shi, Fan; Yang, Renfu; Nian, Feng; Zhang, Zhenwei; Cui, Yongshun; Zhao, Huan; Wang, Nuanrang; Feng, Keming

    2015-05-01

    In this work, a portable atomic frequency standard based on coherent population trapping is designed and demonstrated. To achieve a portable prototype, in the system, a single transverse mode 795nm VCSEL modulated by a 3.4GHz RF source is used as a pump laser which generates coherent light fields. The pump beams pass through a vapor cell containing atom gas and buffer gas. This vapor cell is surrounded by a magnetic shield and placed inside a solenoid which applies a longitudinal magnetic field to lift the Zeeman energy levels' degeneracy and to separate the resonance signal, which has no first-order magnetic field dependence, from the field-dependent resonances. The electrical control system comprises two control loops. The first one locks the laser wavelength to the minimum of the absorption spectrum; the second one locks the modulation frequency and output standard frequency. Furthermore, we designed the micro physical package and realized the locking of a coherent population trapping atomic frequency standard portable prototype successfully. The short-term frequency stability of the whole system is measured to be 6×10-11 for averaging times of 1s, and reaches 5×10-12 at an averaging time of 1000s.

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  10. Outlines of revised regulation standards for experimental research reactors

    International Nuclear Information System (INIS)

    Hohara, Shinya

    2015-01-01

    In response to the accident of TEPCO Fukushima Daiichi Nuclear Power Station, the government took actions through the revision of regulatory standards as well as the complete separation of regulation administrative department from promotion administrative department. The Nuclear and Industrial Safety Agency of the Ministry of Economy, Trade and Industry, which has been in charge of the regulations of commercial reactors, and the Office of Nuclear Regulations of the Ministry of Education, Culture, Sports, Science and Technology, which has been in charge of the regulations of reactors for experiment and research, were separated from both ministries, and integrated into the Nuclear Regulation Authority, which was newly established as the affiliated agency of the Ministry of the Environment. As for the revision of regulations and standards, the Nuclear Safety Commission was dismantled, and regulation enacting authority was given to the new Nuclear Regulation Authority, and the regulations that stipulated new regulatory standards were enacted. This paper outlines the contents of regulations related mainly to the reactors for experiment and research, and explains the following: (1) retroactive application of the new regulatory standards to existing reactor facilities, (2) examinations at the Nuclear Regulatory Agency, (3) procedures to confirm the compliance to the new standards, (4) seismic design classification, and (5) importance classification of safety function. (A.O.)

  11. Fusion reactor design studies: standard accounts for cost estimates

    International Nuclear Information System (INIS)

    Schulte, S.C.; Willke, T.L.; Young, J.R.

    1978-05-01

    The fusion reactor design studies--standard accounts for cost estimates provides a common format from which to assess the economic character of magnetically confined fusion reactor design concepts. The format will aid designers in the preparation of design concept costs estimates and also provide policymakers with a tool to assist in appraising which design concept may be economically promising. The format sets forth a categorization and accounting procedure to be used when estimating fusion reactor busbar energy cost that can be easily and consistently applied. Reasons for developing the procedure, explanations of the procedure, justifications for assumptions made in the procedure, and the applicability of the procedure are described in this document. Adherence to the format when evaluating prospective fusion reactor design concepts will result in the identification of the more promising design concepts thus enabling the fusion power alternatives with better economic potential to be quickly and efficiently developed

  12. The Atomics International (AI) prototype large breeder reactor (PLBR)

    International Nuclear Information System (INIS)

    McDonald, J.S.; Campise, A.V.; Brunings, J.

    1978-01-01

    The AI-PLBR breeder plant design prepared for ERDA and EPRI is of 1000 MWe size, utilizing a loop-type sodium system configuration and producing 2200 psig/850 0 F steam. A 'bullseye' core geometry type sodium system configuration and is employed with Pu0 2 -UO 2 fuel and UO 2 fertile material. The reactor outlet coolant temperature is 930 0 F. A modified 'A'-frame refueling system is employed, which is capable of handling 1/3 of a core loading in 12 days. An inducer-type mechanical pump is used in the primary circuit because of its excellent NPSH characteristics. Hockey sticks steam generators are used to produce near-fossil steam conditions at the turbine throttle. The balance of plant (BOP) design was developed by the architectural-engineering firm of Burns and Roe. It includes Allis-Chalmers - KWU 1800-rpm tandem-compound turbine, selected with high-,intermediate-, and two double-flow low-pressure cylinders equipped with two moisture separator cyclones. A design feature to enhance the licensability of the reference design is the three-level Decay Heat Removal System (DHRS), which consists of normal, backup, and diverse decay heat removal paths. The Atomics International PLBR design illustrated the technical soundness of the LMFBR system for meeting the world's long-term electrical energy supply needs. The design includes design features to assure licensability and innovative engineering features to enhance reliability, constructability, economics, and U.S. utility grid compatibility. The design concept provides a sound basis for the future detailed design of a prototype plant and subsequent development of larger LMFBR plants. (author)

  13. Autoionizing states of atoms calculated using generalized sturmians

    DEFF Research Database (Denmark)

    Avery, James Emil; Avery, John Scales

    2005-01-01

    The generalized Sturmian method is applied to autoionizing states of atoms and ions. If the Goscinskian basis sets allow for a sufficient amount of angular correletion, the calculated energies of doubly-excited (autoionizing) states are found to agree well with the few available experimental...... energies. A large-Z approximation is discussed, and simple formulas are derived which are valid not only for autoionizing states, but for all states of an isoelectronic atomic series. Diagonalization of a small block of the interelectron repulsion matrix yields roots that can be used for a wide range of Z...

  14. ASTM Standards for Reactor Dosimetry and Pressure Vessel Surveillance

    International Nuclear Information System (INIS)

    GRIFFIN, PATRICK J.

    1999-01-01

    The ASTM standards provide guidance and instruction on how to field and interpret reactor dosimetry. They provide a roadmap towards understanding the current ''state-of-the-art'' in reactor dosimetry, as reflected by the technical community. The consensus basis to the ASTM standards assures the user of an unbiased presentation of technical procedures and interpretations of the measurements. Some insight into the types of standards and the way in which they are organized can assist one in using them in an expeditious manner. Two example are presented to help orient new users to the breadth and interrelationship between the ASTM nuclear metrology standards. One example involves the testing of a new ''widget'' to verify the radiation hardness. The second example involves quantifying the radiation damage at a pressure vessel critical weld location through surveillance dosimetry and calculation

  15. Shutdown channels and fitted interlocks in atomic reactors

    International Nuclear Information System (INIS)

    Furet, J.; Landauer, C.

    1968-01-01

    This catalogue consists of tables (one per reactor) giving the following information: number and type of detectors, range of the shutdown channels, nature of the associated electronics, thresholds setting off the alarms, fitted interlocks. These cards have been drawn up with a view to an examination of the reactors safety by the 'Reactor Safety Sub-Commission', they take into account the latest decisions. The reactors involved in this review are: Azur, Cabri, Castor-Pollux, Cesar-Marius-2, Edf-2, EL3, EL4, Eole, G1, G2-G3, Harmonie, Isis, Masurca, Melusine, Minerve, Osiris, Pegase, Peggy, PAT, Rapsodie, SENA, Siloe, Siloette, Triton-Nereide, and Ulysse. (authors) [fr

  16. Application of objective provision tree to development of standard review plan for sodium-cooled fast reactor nuclear design

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Moo-Hoon; Suh, Namduk; Choi, Yongwon; Shin, Andong [Korea Institute of Nuclear Safety, Daejon (Korea, Republic of)

    2016-06-15

    A systematic methodology was developed for the standard review plan for sodium-cooled fast reactor nuclear design. The process is first to develop an objective provision tree of sodium-cooled fast reactor for the reactivity control safety function. The provision tree is generally developed by designer to confirm whether the design satisfies the defense-in-depth concept. Then applicability of the current standard review plan of nuclear design for light water reactor to sodium-cooled fast reactor was evaluated and complemented by the developed objective provision tree.

  17. Atoms

    International Nuclear Information System (INIS)

    Fuchs, Alain; Villani, Cedric; Guthleben, Denis; Leduc, Michele; Brenner, Anastasios; Pouthas, Joel; Perrin, Jean

    2014-01-01

    Completed by recent contributions on various topics (atoms and the Brownian motion, the career of Jean Perrin, the evolution of atomic physics since Jean Perrin, relationship between scientific atomism and philosophical atomism), this book is a reprint of a book published at the beginning of the twentieth century in which the author addressed the relationship between atomic theory and chemistry (molecules, atoms, the Avogadro hypothesis, molecule structures, solutes, upper limits of molecular quantities), molecular agitation (molecule velocity, molecule rotation or vibration, molecular free range), the Brownian motion and emulsions (history and general features, statistical equilibrium of emulsions), the laws of the Brownian motion (Einstein's theory, experimental control), fluctuations (the theory of Smoluchowski), light and quanta (black body, extension of quantum theory), the electricity atom, the atom genesis and destruction (transmutations, atom counting)

  18. Non-standard interaction effects at reactor neutrino experiments

    International Nuclear Information System (INIS)

    Ohlsson, Tommy; Zhang, He

    2009-01-01

    We study non-standard interactions (NSIs) at reactor neutrino experiments, and in particular, the mimicking effects on θ 13 . We present generic formulas for oscillation probabilities including NSIs from sources and detectors. Instructive mappings between the fundamental leptonic mixing parameters and the effective leptonic mixing parameters are established. In addition, NSI corrections to the mixing angles θ 13 and θ 12 are discussed in detailed. Finally, we show that, even for a vanishing θ 13 , an oscillation phenomenon may still be observed in future short baseline reactor neutrino experiments, such as Double Chooz and Daya Bay, due to the existences of NSIs

  19. A proposed standard on medical isotope production in fission reactors

    International Nuclear Information System (INIS)

    Schenter, R. E.; Brown, G. J.; Holden, C. S.

    2006-01-01

    Authors Robert E. Sehenter, Garry Brown and Charles S. Holden argue that a Standard for 'Medical Isotope Production' is needed. Medical isotopes are becoming major components of application for the diagnosis and treatment of all the major diseases including all forms of cancer, heart disease, arthritis, Alzheimer's, among others. Current nuclear data to perform calculations is incomplete, dated or imprecise or otherwise flawed for many isotopes that could have significant applications in medicine. Improved data files will assist computational analyses to design means and methods for improved isotope production techniques in the fission reactor systems. Initial focus of the Standard is expected to be on neutron cross section and branching data for both fast and thermal reactor systems. Evaluated and reviewed tables giving thermal capture cross sections and resonance integrals for the major target and product medical isotopes would be the expected 'first start' for the 'Standard Working Group'. (authors)

  20. Safety Culture Evaluation at Research Reactors of Pakistan Atomic Energy Commission

    International Nuclear Information System (INIS)

    Qamar, M.A.; Saeed, A.; Shah, J.H.

    2016-01-01

    The concept of safety culture was presented by IAEA in document INSAG-4 (1991), delineated as “assembly of characteristics and attitudes in organizations and individuals which establish that, as an overriding priority, nuclear plant safety issues receive the attention warranted by their significance”. The purpose of this paper is to describe the evaluation of safety culture at research reactors of the Pakistan Atomic Energy Commission (PAEC). Evaluating the safety culture of a particular organization poses some challenges which can be resolved by using safety culture evaluation models like those of Sachein (1992) and Harber-Barrier(1998). In PAEC, safety culture is the integral part of management system which not only promotes safety culture throughout the organization but also enhances its significance. To strengthen the safety culture, PAEC is also participating in a number of international and regional meetings of IAEA regarding safety culture. PAEC and the national regulator Pakistan Nuclear Regulatory Authority (PNRA) are also arranging workshops, peer reviews, sharing operational experiences and interacting with IAEA missions to enhance its capabilities in the field of safety culture. The Directorate General of Safety (DOS) is a corporate office of PAEC for safety and regulatory matters. DOS is in the process of implementing a program to evaluate safety culture at nuclear installations of PAEC to ensure that safety culture is included as a vital segment of the Integral Management System of the establishment. In this regard, training sessions and lectures on safety culture evaluation are normally conducted in PAEC for awareness and enhancement of the safety culture program. Safety culture is also addressed in PNRA Regulations like PAK-909 and PAK-913. In this paper we will focus on the safety culture evaluation in our research reactors, i.e., PARR-1 and PARR-2. The evaluation results will be based on observations, interviews of employees, group discussions

  1. Cirus reactor: a milestone in Indian Atomic Energy Programme

    International Nuclear Information System (INIS)

    Ranjan, Rakesh; Karhadkar, C.G.; Bhattacharya, S.

    2017-01-01

    Cirus, a 40 MW_t_h, high flux, thermal neutron research reactor achieved first criticality on 10"t"h July 1960. It had vertical core, natural metallic Uranium rods in Aluminium clad as fuel, demineralised light water as coolant, heavy water as moderator, Helium as cover gas and graphite as reflector. A low-pressure containment was provided for the reactor and some of the important associated reactor systems. Reactor start-up and power regulation was effected by controlled adjustment of moderator level in the reactor vessel. Boron carbide rods were used as primary shut down devices. Dumping of heavy water from core worked as secondary shut down device. Seawater was used as secondary coolant for removal of the fission heat of the reactor. Initial operation of Cirus was marred by several difficulties, primarily arising out of water chemistry in primary cooling water system. It took almost 3 years to systematically resolve these problems and achieve stable operation of reactor. Cirus could be operated at its rated power by the year 1963

  2. Standard Technical Specifications, General Electric plants, BWR/4

    International Nuclear Information System (INIS)

    1992-09-01

    This NUREG contains improved Standard Technical Specifications (STS) for General Electric Plants, BWR/4, and documents the positions of the Nuclear Regulatory Commission based on the BWR Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. The improved STS will be used as the basis for individual nuclear power specifications. This report contains three volumes. This document, Volume 2, contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS

  3. Prospect of atomic airplane and mini-reactor

    International Nuclear Information System (INIS)

    Shi Lei; Li Jinying; Ji Cunxing; Ma Huirou; Hou Yanli; Duan Zhihui

    2012-01-01

    History of R and D of atomic airplane was reviewed. inclusive of problem and actuality, principle and attaching. Prospect of R and D of atomic airplane in China is described extensively. Some suggestions to the government and establishments were made as well. (authors)

  4. Standard Guide for Benchmark Testing of Light Water Reactor Calculations

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This guide covers general approaches for benchmarking neutron transport calculations in light water reactor systems. A companion guide (Guide E2005) covers use of benchmark fields for testing neutron transport calculations and cross sections in well controlled environments. This guide covers experimental benchmarking of neutron fluence calculations (or calculations of other exposure parameters such as dpa) in more complex geometries relevant to reactor surveillance. Particular sections of the guide discuss: the use of well-characterized benchmark neutron fields to provide an indication of the accuracy of the calculational methods and nuclear data when applied to typical cases; and the use of plant specific measurements to indicate bias in individual plant calculations. Use of these two benchmark techniques will serve to limit plant-specific calculational uncertainty, and, when combined with analytical uncertainty estimates for the calculations, will provide uncertainty estimates for reactor fluences with ...

  5. On the actual controlling of standards concerning the 'fast breeder reactor'

    International Nuclear Information System (INIS)

    Rengeling, H.W.

    1978-01-01

    If the decision of the OVG Muenster to present the case to the Federal Constitutional Court asking whether Article 7 of the atomic energy law corresponds to the constitution or not, as far as the article allows the licensing of a fast breeder reactor, two problems arise: The legal question in how far the actual controlling of standards is to be preceeded by a statement of facts given by the court of first instance, and the problem behind concerning the responsibility of decision. The Federal Constitutional Court should accept the responsibility of decision to be borne by it. (orig.) [de

  6. Nuclear reactors. Use of the protection system for non-safety purposes (International Electrotechnical Commission Standard Publication 639:1979)

    International Nuclear Information System (INIS)

    Stefanik, J.

    1996-01-01

    This standard applies to the protection system of a nuclear reactor and, more especially, to all interconnections between a reactor protection system (as defined and explained in International Electrotechnical Commission Publication 231 A, first supplement to Publication 231, General Principles of Nuclear Reactor Instrumentation) and all other systems and equipment not part of the protection system, except: a) the physical connection between sensors of the protection system and the physical variables that they monitor, such as for example, thermo wells, moderating medium for neutron sensors, etc.; b) the electrical connection between the protection system and the reactor control rods or other safety mechanism; c) the electrical and pneumatic connections to the power distribution system (mains) and pneumatic supplies that supply power to the protection system. Although many clauses relate to all reactor protection systems, this standard applies mainly to protection systems in nuclear power reactors

  7. Mercury Atomic Frequency Standards for Space Based Navigation and Timekeeping

    Science.gov (United States)

    Tjoelker, R. L.; Burt, E. A.; Chung, S.; Hamell, R. L.; Prestage, J. D.; Tucker, B.; Cash, P.; Lutwak, R.

    2012-01-01

    A low power Mercury Atomic Frequency Standard (MAFS) has been developed and demonstrated on the path towards future space clock applications. A self contained mercury ion breadboard clock: emulating flight clock interfaces, steering a USO local oscillator, and consuming approx 40 Watts has been operating at JPL for more than a year. This complete, modular ion clock instrument demonstrates that key GNSS size, weight, and power (SWaP) requirements can be achieved while still maintaining short and long term performance demonstrated in previous ground ion clocks. The MAFS breadboard serves as a flexible platform for optimizing further space clock development and guides engineering model design trades towards fabrication of an ion clock for space flight.

  8. Black-body radiation effects and light shifts in atomic frequency standards

    Energy Technology Data Exchange (ETDEWEB)

    Pal' chikov, V G; Domnin, Yu S; Novoselov, A V [Institute of Metrology for Time and Space at National Research Institute for Physical-Technical and Radiotechnical Measurements - IMVP GP VNIIFTRI, Mendeleevo, Moscow Region, 141570 (Russian Federation)

    2003-04-01

    A general method is presented for calculating the higher-order terms of series in powers of the black-body radiation field for the Stark-state wavefunctions, dipole transition matrix elements and corresponding frequency shifts of hyperfine splitting in the ground states for Cs and Rb atoms. A numerical method for calculating the light shifts in Sr atoms is described. It is based on the Green function method for summation over all intermediate states and exact Dirac-Fock wavefunctions for the resonant transitions to the first excited s-, p- and d-states. By comparing the calculated Stark shift with results of measurements employing atomic frequency standards, the black-body radiation effects on the ground state are analysed.

  9. Neutron standard cross sections in reactor physics - Need and status

    International Nuclear Information System (INIS)

    Carlson, A.D.

    1990-01-01

    The design and improvement of nuclear reactors require detailed neutronics calculations. These calculations depend on comprehensive libraries of evaluated nuclear cross sections. Most of the cross sections that form the data base for these evaluations have been measured relative to neutron cross-section standards. The use of these standards can often simplify the measurement process by eliminating the need for a direct measurement of the neutron fluence. The standards are not known perfectly, however; thus the accuracy of a cross-section measurement is limited by the uncertainty in the standard cross section relative to which it is measured. Improvements in a standard cause all cross sections measured relative to that standard to be improved. This is the reason for the emphasis on improving the neutron cross-section standards. The continual process of measurement and evaluation has led to improvements in the accuracy and range of applicability of the standards. Though these improvements have been substantial, this process must continue in order to obtain the high-quality standards needed by the user community

  10. An inspection standard of fuel for the high temperature engineering test reactor

    International Nuclear Information System (INIS)

    Kobayashi, Fumiaki; Shiozawa, Shusaku; Sawa, Kazuhiro; Sato, Sadao; Hayashi, Kimio; Fukuda, Kosaku; Kaneko, Mitsunobu; Sato, Tsutomu.

    1992-06-01

    The High Temperature Engineering Test Reactor (HTTR) uses the fuel comprising coated fuel particles. A general inspection standard for the coated particle fuel, however, has not been established in Japan. Therefore, it has been necessary to prescribe the inspection standard of the fuel for HTTR. Under these circumstances, a fuel inspection standard of HTTR has been established under cooperation of fuel specialists both inside and outside of JAERI on referring to the inspection methods adopted in USA, Germany and Japan for HTGR fuels. Since a large number of coated fuel particle samples is needed to inspect the HTTR fuel, the sampling inspection standard has also been established considering the inspection efficiency. This report presents the inspection and the sampling standards together with an explanation of these standards. These standards will be applied to the HTTR fuel acceptance tests. (author)

  11. Monitoring system for accuracy and reliability characteristics of standard temperature measurements in WWER-440 reactors

    International Nuclear Information System (INIS)

    Stanc, S.; Repa, M.

    2001-01-01

    Description of a monitoring system for accuracy and reliability characteristics of standard temperature measurements in WWER-440 reactors and benefits obtained from its use are shown in the presentation. As standard reactor temperature measurement, coolant temperature measurement at fuel assembly outlets and in loops, entered into the In-Reactor Control System , are considered. Such systems have been implemented at two V-230 reactors and are under implementation at other four V-213 reactors. (Authors)

  12. Non-Power Reactor Operator Licensing Examiner Standards. Revision 1

    International Nuclear Information System (INIS)

    1995-06-01

    The Non-Power Reactor Operator Licensing Examiner Standards provide policy and guidance to NRC examiners and establish the procedures and practices for examining and licensing of applicants for NRC operator licenses pursuant to Part 55 of Title 10 of the Code of Federal Regulations (10 CFR 55). They are intended to assist NRC examiners and facility licensees to understand the examination process better and to provide for equitable and consistent administration of examinations to all applicants by NRC examiners. These standards are not a substitute for the operator licensing regulations and are subject to revision or other internal operator examination licensing policy changes. As appropriate, these standards will be revised periodically to accommodate comments and reflect new information or experience

  13. Non-Power Reactor Operator Licensing Examiner Standards

    International Nuclear Information System (INIS)

    1994-06-01

    The Non-Power Reactor Operator Licensing Examiner Standards provide policy and guidance to NRC examiners and establish the procedures and practices for examining and licensing of applicants for NRC operator licenses pursuant to Part 55 of Title 10 of the Code of Federal Regulations (10 CFR Part 55). They are intended to assist NRC examiners and facility licensees to understand the examination process better and to provide for equitable and consistent administration of examinations to all applicants by NRC examiners. These standards are not a substitute for the operator licensing regulations and are subject to revision or other internal operator examination licensing policy changes. As appropriate, this standard will be revised periodically to accommodate comments and reflect new information or experience

  14. Non-Power Reactor Operator Licensing Examiner Standards. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-06-01

    The Non-Power Reactor Operator Licensing Examiner Standards provide policy and guidance to NRC examiners and establish the procedures and practices for examining and licensing of applicants for NRC operator licenses pursuant to Part 55 of Title 10 of the Code of Federal Regulations (10 CFR 55). They are intended to assist NRC examiners and facility licensees to understand the examination process better and to provide for equitable and consistent administration of examinations to all applicants by NRC examiners. These standards are not a substitute for the operator licensing regulations and are subject to revision or other internal operator examination licensing policy changes. As appropriate, these standards will be revised periodically to accommodate comments and reflect new information or experience.

  15. Generalized Hartree-Fock method for electron-atom scattering

    International Nuclear Information System (INIS)

    Rosenberg, L.

    1997-01-01

    In the widely used Hartree-Fock procedure for atomic structure calculations, trial functions in the form of linear combinations of Slater determinants are constructed and the Rayleigh-Ritz minimum principle is applied to determine the best in that class. A generalization of this approach, applicable to low-energy electron-atom scattering, is developed here. The method is based on a unique decomposition of the scattering wave function into open- and closed-channel components, so chosen that an approximation to the closed-channel component may be obtained by adopting it as a trial function in a minimum principle, whose rigor can be maintained even when the target wave functions are imprecisely known. Given a closed-channel trial function, the full scattering function may be determined from the solution of an effective one-body Schroedinger equation. Alternatively, in a generalized Hartree-Fock approach, the minimum principle leads to coupled integrodifferential equations to be satisfied by the basis functions appearing in a Slater-determinant representation of the closed-channel wave function; it also provides a procedure for optimizing the choice of nonlinear parameters in a variational determination of these basis functions. Inclusion of additional Slater determinants in the closed-channel trial function allows for systematic improvement of that function, as well as the calculated scattering parameters, with the possibility of spurious singularities avoided. Electron-electron correlations can be important in accounting for long-range forces and resonances. These correlation effects can be included explicitly by suitable choice of one component of the closed-channel wave function; the remaining component may then be determined by the generalized Hartree-Fock procedure. As a simple test, the method is applied to s-wave scattering of positrons by hydrogen. copyright 1997 The American Physical Society

  16. Interaction of relativistic elementary atoms with matter. I. General formulas

    International Nuclear Information System (INIS)

    Mrowczyn'ski, S.

    1987-01-01

    The problem of the interaction of relativistic elementary atoms (Coulomb bound states of elementary particles such as positronium, pionium, etc.) with matter is studied in the reference frame where the atom is initially at rest. An atom of matter is treated as a spinless structureless fast particle. The amplitudes of elementary-atom interaction are derived in the Born approximation under the assumption that a momentum transfer to the atom does not significantly exceed an inverse Bohr radius of the atom. The elementary-atom excitation and ionization processes are considered. The transitions where the spin projection of the atom component is reversed are also studied. In particular the matrix elements for para-ortho and ortho-para transitions are given. The spin structure of the amplitudes is discussed in detail. The sum rules, which allow the calculation of the cross sections summed over atom final states are found. Finally the formulas of the atom interaction cross sections are presented

  17. The conceptual design of the standard and the reduced fuel assemblies for an advanced research reactor

    International Nuclear Information System (INIS)

    Ryu, Jeong Soo; Cho, Yeong Garp; Yoon, Doo Byung; Dan, Ho Jin; Chae, Hee Tack; Park, Cheol

    2005-01-01

    HANARO (Hi-flux Advanced Neutron Application Reactor), is an open-tank-in-pool type research reactor with a thermal power of 30MW. The HANARO has been operating at Korea Atomic Energy Research Institute since 1995. Based on the technical experiences in design and operation for the HANARO, the design of an Advanced Research Reactor (ARR) was launched by KAERI in 2002. The final goal of the project is to develop a new and advanced research reactor model which is superior in safety and economical aspects. This paper summarizes the design improvements of the conceptually designed standard fuel assembly based on the analysis results for the nuclear physics. It includes also the design of the reduced fuel assembly in conjunction with the flow tube as the fuel channel and the guide of the absorber rod. In the near future, the feasibility of the conceptually designed fuel assemblies of the ARR will be verified by investigating the dynamic and the thermal behaviors of the fuel assembly submerged in coolant

  18. Recent developments for fast reactor structural design standard (FDS)

    International Nuclear Information System (INIS)

    Kasahara, N.; Nakamuria, K.; Morishita, M.; Shibamoto, H.; Nagashima, H.; Inoue, K.

    2005-01-01

    For realization of reliable and economical fast reactor (FR) plants, Japan Nuclear Cycle Development Institute(JNC) and Japan Atomic Power Company(JAPC) are cooperating on 'Feasibility Study on Commercialized FR Cycle Systems'. To certify the design concepts through evaluation of their structural integrity, the research and development of 'Elevated Temperature Structural Design Guide for Commercialized Fast Reactor (FDS)' is recognized as an essential theme. FDS focuses on particular failure modes of FRs such as ratchet deformation and creep fatigue damages due to cyclic thermal loads. To evaluate these modes, three main developments are in progress. One is 'Refinement of Failure Criteria' for particular modes of FRs. Next is development of 'Guidelines for Inelastic Design Analysis' in order to predict elastic plastic and creep behaviors. Furthermore, efforts are being made toward preparing 'Guidelines for Thermal Load Modeling' for FR component design where thermal loads are dominant. These studies were performed under the sponsorship of the Ministry of Economy, Trade and Industry of Japanese government. (authors)

  19. Activities of the data centers on atomic spectroscopy at the National Institute of Standards and Technology

    International Nuclear Information System (INIS)

    Wiese, W.L.

    1990-01-01

    The activities concerning Atomic Energy Levels and Wavelengths, Atomic Transition Probabilities and Spectral Line Shapes and Shifts at the National Institute of Standards and Technology (Gaithersburg, USA) are listed together with the corresponding lists of publications

  20. General safety orientations of the Jules Horowitz Reactor Project (JHRP)

    International Nuclear Information System (INIS)

    Tremodeux, P.; Fiorini, G.L.

    2000-01-01

    After a brief reminder of the JHR purpose, the document outlines the General Safety related Orientations/Recommendations used for the design and the safety assessment of the facility. As far as the JHR design is new, the safety philosophy adopted for this reactor will be as consistent as possible with that recommended for future (power...) reactors. The general recommendations developed in the paper are: the general nuclear safety approach for the design, operation and analysis with, in particular, the adoption of the Defence In Depth principle; the general safety objectives in terms of radiological consequences; the use of Probabilistic Safety Studies; quality assurance. The 'Defence in Depth' concept using amongst others the 'Barrier' principle remains the basis of the JHR safety. 'Defence In Depth' is applied both to design and operation. Its adequacy is checked during the safety assessment and the paper gives the technical recommendations that should allow the designer to implement this concept into the final design. Built mainly for experimental irradiation the JHR facilities will be handled according to conventional or new operation rules which could put materials under stress and entail handling errors. Specific recommendations are defined to take into account the corresponding peculiarities; they are discussed in the paper. The safety design of the JHR takes into account the experience accumulated through the CEA experimental irradiation programmes, which represents several dozen reactor years; the consultation of CEA reactor facilities operators is ongoing. The corresponding feedback is shortly described. Recommendations related to maintenance and associated operation are indicated as well as those regarding the human factor. Details are given on the JHR safety practical implementation through the CEA/DRN Safety approach. Details of the corresponding Safety Objectives are also discussed. Finally the designer position on the role of probabilistic safety

  1. Ministry of ordinance determining the technical standard concerning atomic energy facilities for power generation

    International Nuclear Information System (INIS)

    1985-01-01

    The ministerial ordinance provides for the technical standards for the power generation of nuclear facilities; i.e., electric power facilities generating electricity with nuclear energy for motive power, according to the Electricity Enterprises Act. The contents are as follows: protection against fires, aseismatic design, radiation protective barriers, structural protection for sitings, reactor installation, safety measures, materials and structures, safety valves, pressure resistance tests, reactor core, radiation shields, reactor cooling, emergency core cooling system, facility equipment, alarm system, reactor control system, reactor control room, fuel storage facility, fuel handling facility, ventilation equipment, radioactive contamination prevention, radioactive waste management facility, reactor containment facility, and so on. (Kubozono, M.)

  2. On the path to ordering standardized advanced light water reactors

    International Nuclear Information System (INIS)

    Sliter, G.E.

    1997-01-01

    The international Advanced Light Water Reactor (ALWR) program is specifying, designing, and certifying the next generation of nuclear power plants. Begun in the mid-1980's, the program is on track to permit ordering and construction of families of standardized plants at the start of the twenty-first century. ALWRs will be constructed only if they are economically competitive with alternative forms of electricity generation and are recognized as acceptable and favorable by the public, prospective owners, and investors. This paper first gives an overview of the major building blocks ensuring safe, reliable, and economic designs and the status of those designs. Next it lays out the path the industry has charted toward adopting the ALWR option and indicates the status of three key steps -- design certification, utility requirements, and first-of-a-kind engineering. Lastly, the paper focuses on one of the most important building blocks for ensuring economic viability -- life-cycle standardization. Among the topics are the definition and scope of standardization; its advantages and disadvantages; design team standardization plans that describe the desired or optimum degree of standardization and the processes used to achieve it; and the need for an agreement among all plant owners and operators for implementing and sustaining standardization in families of ALWRs. 10 refs., 5 figs

  3. Ways of prevention of accidents at atomic reactor

    International Nuclear Information System (INIS)

    Takibaev, Zh. S.

    2000-01-01

    The methods proposed to prevent such a move are discussed as well as the scheme of their realization. To improve reactor operation characteristics the safeguard system of quick response is used. Nowadays direct-acting safeguard system (DAS) is to be worked out. It reacts on the main cause of the accident the rapid growth of neutron flux. The time delay of combined gas-liquid DAS unit and fluctuation of nuclear power are calculated. The DAS grid disposed in active zone is developed. Fissile materials are employed because their heating almost immediately follows the growth of neutron flux. There are several systems proposed: uranium bimetal dispersed absorber, uranium hexafluoride liquid absorber (gadolinium solution).Neutronic calculation is done for WWR-1000. The model suggested acts over 0.12 sec. after reactivity swing of 0.003, becomes a 'safety rod' over time delay of 1.49 sec. and cleans itself over 3.0 sec. after.The study presents its improved version. Absorber is injected dose by dose and thus negative reactivity is introduced discretely. Accordingly the same system can act by extracting some parts of fuel from the core. Bimetal safeguard systems are studied. The methods suggested above seem proved in the sense of strengthening nuclear energy development in the future. The problem of DAS and other safeguard systems to prevent reactivity accidents for various reactor types including computer simulation is set to be studied further

  4. SRAC: JAERI thermal reactor standard code system for reactor design and analysis

    International Nuclear Information System (INIS)

    Tsuchihashi, Keichiro; Takano, Hideki; Horikami, Kunihiko; Ishiguro, Yukio; Kaneko, Kunio; Hara, Toshiharu.

    1983-01-01

    The SRAC (Standard Reactor Analysis Code) is a code system for nuclear reactor analysis and design. It is composed of neutron cross section libraries and auxiliary processing codes, neutron spectrum routines, a variety of transport, 1-, 2- and 3-D diffusion routines, dynamic parameters and cell burn-up routines. By making the best use of the individual code function in the SRAC system, the user can select either the exact method for an accurate estimate of reactor characteristics or the economical method aiming at a shorter computer time, depending on the purpose of study. The user can select cell or core calculation; fixed source or eigenvalue problem; transport (collision probability or Sn) theory or diffusion theory. Moreover, smearing and collapsing of macroscopic cross sections are separately done by the user's selection. And a special attention is paid for double heterogeneity. Various techniques are employed to access the data storage and to optimize the internal data transfer. Benchmark calculations using the SRAC system have been made extensively for the Keff values of various types of critical assemblies (light water, heavy water and graphite moderated systems, and fast reactor systems). The calculated results show good prediction for the experimental Keff values. (author)

  5. General aspects to be considered in a research reactor decommissioning plan

    International Nuclear Information System (INIS)

    Grossi, Pablo Andrade; Tello, Cledola Cassia Oliveira de

    2009-01-01

    There are more than 200 research reactors that have either not operated for a considerable period of time and may never return to operation or, are close to permanent shutdown. Many countries do not have a decommissioning policy, and like Brazil not all installations have their decommissioning plan as part of the licensing documentation. Brazil is signatory of Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management, but until now there is no decommissioning policy, and specifically for research reactor there is no decommissioning guidelines in the standards. The Nuclear Technology Development Centre (CDTN/CNEN-MG) has a TRIGA Mark I Research Reactor IPR-R1 in operation for 47 years with 3.6% average fuel burn-up. The original power was 100 kW and it is being licensed for 250 kW, and it needs the decommissioning plan as part of the licensing requirements. In the paper it is presented general aspects and contents of a Research Reactors Decommissioning Plan. As the Brazilian regulatory body so far does not have a decommissioning policy established neither a regulatory framework in this issue, individual efforts are being integrated to establish a National Decommissioning Group (matrix structure) to perform the decommissioning planning and activities. The approach used for IPR-R1 is presented as suggestions to develop the national regulatory standards on this issue and applied to Brazilian Research Reactors and other nuclear facilities. (author)

  6. 49 CFR 178.915 - General Large Packaging standards.

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 2 2010-10-01 2010-10-01 false General Large Packaging standards. 178.915 Section... PACKAGINGS Large Packagings Standards § 178.915 General Large Packaging standards. (a) Each Large Packaging.... Large Packagings intended for solid hazardous materials must be sift-proof and water-resistant. (b) All...

  7. 40 CFR 63.102 - General standards.

    Science.gov (United States)

    2010-07-01

    ... extent practical. The general duty to minimize emissions during a period of startup, shutdown, or... procedures (including the startup, shutdown, and malfunction plan required in § 63.6(e)(3)), review of... § 63.100(k) of this subpart. [59 FR 19454, Apr. 22, 1994, as amended at 60 FR 63626, Dec. 12, 1995; 61...

  8. The safety of Ontario's nuclear power reactors. A scientific and technical review. A submission to the Ontario Nuclear Safety Review by Atomic Energy Canada Limited

    International Nuclear Information System (INIS)

    1987-01-01

    This submission comments on the evolution of the Canadian nuclear program, the management of safety, and the reactor design, analysis, operation and research programs that contribute to the safety of the CANDU reactor and provide assurance of safety to the regulatory agency and to the public. The CANDU reactor system has been designed and developed with close cooperation between Atomic Energy of Canada Ltd. (AECL), utilities, manufacturers, and the Atomic Energy Control Board (AECB). The AECB has the responsibility, on behalf of the public, for establishing acceptable standards with respect to public risk and for establishing through independent review that these standards are satisfied. The plant designer has responsibility for defining how those standards will be met. The plant operator has responsibility for operating within the framework of those standards. The Canadian approach to safety design is based on the philosophy of defence in depth. Defence in depth is achieved through a high level of equipment quality, system redundancy and fail-safe design; regulating and process systems designed to maintain all process systems within acceptable operating parameters; and, independent safety systems to shut down the reactor, provide long-term cooling, and contain potential release of radioactivity in the event of an accident. The resulting design meets regulatory requirements not only in Canada but also in other countries. Probabilistic safety and risk evaluations show that the CANDU design offers a level of safety and least as good as other commercially available reactor designs

  9. 42 CFR 493.1469 - Standard: Cytology general supervisor qualifications.

    Science.gov (United States)

    2010-10-01

    ... 42 Public Health 5 2010-10-01 2010-10-01 false Standard: Cytology general supervisor... Nonwaived Testing Laboratories Performing High Complexity Testing § 493.1469 Standard: Cytology general supervisor qualifications. The cytology general supervisor must be qualified to supervise cytology services...

  10. The standard calibration instrument automation system for the atomic absorption spectrophotometer. Part 3: Program documentation

    Science.gov (United States)

    Ryan, D. P.; Roth, G. S.

    1982-04-01

    Complete documentation of the 15 programs and 11 data files of the EPA Atomic Absorption Instrument Automation System is presented. The system incorporates the following major features: (1) multipoint calibration using first, second, or third degree regression or linear interpolation, (2) timely quality control assessments for spiked samples, duplicates, laboratory control standards, reagent blanks, and instrument check standards, (3) reagent blank subtraction, and (4) plotting of calibration curves and raw data peaks. The programs of this system are written in Data General Extended BASIC, Revision 4.3, as enhanced for multi-user, real-time data acquisition. They run in a Data General Nova 840 minicomputer under the operating system RDOS, Revision 6.2. There is a functional description, a symbol definitions table, a functional flowchart, a program listing, and a symbol cross reference table for each program. The structure of every data file is also detailed.

  11. Standard Technical Specifications, General Electric plants, BWR/4

    International Nuclear Information System (INIS)

    1992-09-01

    This NUREG contains improved Standard Technical Specifications (STS) for General Electric Plants, BWR/6, and documents the positions of the Nuclear Regulatory Commission based on the B ampersand W Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume 1 contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. This document Volume 3, contains the Bases for Sections 3.4--3.10 of the improved STS

  12. Standard Technical Specifications, General Electric Plants, BWR/6

    International Nuclear Information System (INIS)

    1992-09-01

    This NUREG contains improved Standard Technical Specifications (STS) for General Electric Plants, BWR/6, and documents the positions of the Nuclear Regulatory Commission based on the B ampersand W Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. This document Volume 1, contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. Volume 3 contains the Bases for Sections 3.4--3.10 of the improved STS

  13. Standard Technical Specifications, General Electric plants, BWR/4

    International Nuclear Information System (INIS)

    1992-09-01

    This NUREG contains improved Standard Technical Specifications (STS) for General Electric Plants, BWR/6, and documents the positions of the Nuclear Regulatory Commission based on the BWR Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. This document Volume 1 contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. Volume 3 contains the Bases for Sections 3.4--3.10 of the improved STS

  14. Standard Technical Specifications, General Electric Plants, BWR/6

    International Nuclear Information System (INIS)

    1992-09-01

    This NUREG contains improved Standard Technical Specifications (STS) for General Electric Plants, BWR/4, and documents the positions of the Nuclear Regulatory Commission based on the BWR Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume 1 contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. Volume 3, contains the Bases for Sections 3.4--3.10 of the improved STS

  15. Standard Technical Specifications, General Electric plants, BWR/6

    International Nuclear Information System (INIS)

    1992-09-01

    This NUREG contains improved Standard Technical Specifications (STS) for General Electric Plants, BWR/6, and documents the positions of the Nuclear Regulatory Commission based on the B ampersand W Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume 1 contains the Specifications for all chapters and sections of the improved STS. This document Volume 2, contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. Volume 3 contains the Bases for Sections 3.4--3.10 of the improved STS

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

    Directory of Open Access Journals (Sweden)

    Curtisha D. Travis

    2013-08-01

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

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

    International Nuclear Information System (INIS)

    Hayashi, Takeo

    1979-01-01

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

  18. Results of the General Atomic deposition loop program

    International Nuclear Information System (INIS)

    Hanson, D.L.

    1976-01-01

    The transport behavior of fission products in flowing helium streams has been studied to determine their deposition and re-entrainment characteristics. Such information is required for the design and safety analysis of high-temperature gas-cooled reactors (HTGRs). A small high-pressure, high-temperature loop was constructed for deposition studies at near-HTGR conditions. Five loop experiments were performed to determine the plateout distribution of iodine, strontium, and cesium. In general, the plateout activity showed an exponential decrease with distance from the source with enhanced plateout at flow disturber locations (contractions, bends, etc.) and especially in a chill section where the surface was cooled. Blowdown tests were performed on selected loop specimens to determine the amount of re-entrainment caused by abnormally high wall shear stresses. The liftoff fraction (fractional amount removed) was shown to vary approximately linearly with the shear ratio (defined as the ratio of the steady state wall shear stress under blowdown conditions to that under normal operating conditions). Blowdown results are also reported for pipe sections taken from the GAIL-IV in-pile loop. Attempts were made to correlate these plateout data with the PAD code (Plateout Activity Distribution) which was developed for prediction of plateout distribution in an HTGR primary circuit. Because of inadequate modeling of the effects of the chill section, the agreement was generally poor. Consequently, to test further the PAD code, a review of the available plateout literature was made. Plateout distributions in the Peach Bottom and Dragon HTGRs and the Battelle Memorial Institute out-of-pile loop were successfully modeled

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

    International Nuclear Information System (INIS)

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

    1983-05-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1983-05-01

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

  1. Presentation Stations of the General Atomics Fusion Educational Program

    Science.gov (United States)

    Lee, R. L.; Fusion Group Education Outreach Team

    1996-11-01

    The General Atomics Fusion Group's Educational Program has been actively promoting fusion science and applications throughout San Diego County's secondary school systems for over three years. The educational program allows many students to learn more about nuclear fusion science, its applications, and what it takes to become an active participant in an important field of study. It also helps educators to better understand how to teach fusion science in their classroom. Tours of the DIII--D facility are a centerpiece of the program. Over 1000 students visited the DIII--D research facility during the 1995--1996 school year for a half-day of presentations, discussions, and hands-on learning. Interactive presentations are provided at six different stations by GA scientists and engineers to small groups of students during the tours. Stations include topics on energy, plasma science, the electromagnetic spectrum, radiation and risk assessment, and data acquisition. Included also is a tour of the DIII--D machine hall and model where students can see and discuss many aspects of the tokamak. Portions of each station will be presented and discussed.

  2. Comparison of standard fast reactor calculations (Baker model)

    Energy Technology Data Exchange (ETDEWEB)

    Voropaev, A I; Van' kov, A A; Tsybulya, A M

    1978-12-01

    Compared are standard fast reactor calculations performed at different laboratories using several nuclear data files: BNAB-70 and OSKAR-75 (the USSR), CARNAVAL-4 (France), FD-5 (Great Britain), KFK-INR (West Germany), ENDF/B4 (the USA). Three fuel compositions were chosen: (1) /sup 239/Pu and /sup 238/U; (2) /sup 239/Pu, /sup 238/U and fission products; (3) /sup 239/Pu, /sup 240/Pu, /sup 238/U and fission products. Medium temperature was 300K. The calculations have been conducted in the diffusion approximation. Data on critical masses and breeding ratios are tabulated. Discrepancies in the calculations of all the characteristics are small since all the countries possess practically the same nuclear data files.

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

    International Nuclear Information System (INIS)

    Balakrishnan, M.R.

    1993-01-01

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

  4. 42 CFR 493.1471 - Standard: Cytology general supervisor responsibilities.

    Science.gov (United States)

    2010-10-01

    ... 42 Public Health 5 2010-10-01 2010-10-01 false Standard: Cytology general supervisor... Nonwaived Testing Laboratories Performing High Complexity Testing § 493.1471 Standard: Cytology general supervisor responsibilities. The technical supervisor of cytology may perform the duties of the cytology...

  5. 42 CFR 493.837 - Standard; General immunology.

    Science.gov (United States)

    2010-10-01

    ... 42 Public Health 5 2010-10-01 2010-10-01 false Standard; General immunology. 493.837 Section 493.837 Public Health CENTERS FOR MEDICARE & MEDICAID SERVICES, DEPARTMENT OF HEALTH AND HUMAN SERVICES... These Tests § 493.837 Standard; General immunology. (a) Failure to attain a score of at least 80 percent...

  6. 47 CFR 51.503 - General pricing standard.

    Science.gov (United States)

    2010-10-01

    ... 47 Telecommunication 3 2010-10-01 2010-10-01 false General pricing standard. 51.503 Section 51.503... Pricing of Elements § 51.503 General pricing standard. (a) An incumbent LEC shall offer elements to... commission— (1) Pursuant to the forward-looking economic cost-based pricing methodology set forth in §§ 51...

  7. 78 FR 73898 - Operator Licensing Examination Standards for Power Reactors

    Science.gov (United States)

    2013-12-09

    [email protected] . Both of the Office of New Reactors; or Timothy Kolb, Office of Nuclear Reactor Regulation, U...: timothy.kolb@nrc.gov . SUPPLEMENTARY INFORMATION: I. Accessing Information and Submitting Comments A...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-01-15

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

  9. Compact, Low-Power Atomic Time and Frequency Standards

    Science.gov (United States)

    2008-12-01

    2007). This is consistent with other reports of survival of CSAC devices with thin polymide tethers to 500g ( Lutwak et al., 2007). • Humidity...InterPACK 󈧋 , July 8-12, 2007, Vancouver, British Columbia, CANADA Lutwak , R., et al., “The chip-scale atomic clock – prototype evaluation

  10. The Brazilian time and frequency atomic standards program

    Directory of Open Access Journals (Sweden)

    Mushtaq Ahmed

    2008-06-01

    Full Text Available Cesium atomic beam clocks have been the workhorse for many demanding applications in science and technology for the past four decades. Tests of the fundamental laws of physics and the search for minute changes in fundamental constants, the synchronization of telecommunication networks, and realization of the satellite-based global positioning system would not be possible without atomic clocks. The adoption of optical cooling and trapping techniques, has produced a major advance in atomic clock precision. Cold-atom fountain and compact cold-atom clocks have also been developed. Measurement precision of a few parts in 10(15 has been demonstrated for a cold-atom fountain clock. We present here an overview of the time and frequency metrology program based on cesium atoms under development at USP São Carlos. This activity consists of construction and characterization of atomic-beam, and several variations of cold-atom clocks. We discuss the basic working principles, construction, evaluation, and important applications of atomic clocks in the Brazilian program.Relógios atômicos de feixe de Césio têm sido a base para diversas aplicações em ciência e tecnologia nas últimas quatro décadas. Testes de leis fundamentais de física, buscas por mínimas variações em constantes fundamentais, sincronização de redes de telecomunicações e o funcionamento do sistema de posicionamento global, baseado em satélites de navegação, não seriam possíveis sem os relógios atômicos. A adoção de técnicas de aprisionamento e resfriamento ópticos tem permitido um grande avanço na precisão dos relógios atômicos. Chafarizes de átomos frios e relógios compactos de átomos frios também têm sido desenvolvidos. Precisões de medida de algumas partes em 1015 foram demonstradas para relógios do tipo chafariz de átomos frios. Apresentamos uma visão geral do programa de metrologia de tempo e freqüência baseado em átomos de césio, em

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

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  12. Reactor safety

    International Nuclear Information System (INIS)

    Butz, H.P.; Heuser, F.W.; May, H.

    1985-01-01

    The paper comprises an introduction into nuclear physics bases, the safety concept generally speaking, safety devices of pwr type reactors, accident analysis, external influences, probabilistic safety assessment and risk studies. It further describes operational experience, licensing procedures under the Atomic Energy Law, research in reactor safety and the nuclear fuel cycle. (DG) [de

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

    Science.gov (United States)

    Mozetic, Miran; Cvelbar, Uros

    2009-08-01

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

  14. Proposal of an ISO Standard: Classification of Transients and Accidents for Pressurized Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Jo, Jong Chull [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Chung, Bub Dong; Lee, Doo-Jeong; Kim, Jong In; Yoon, Ju Hyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Jeong, Jae Jun [Pusan National Univ., Busan (Korea, Republic of); Kim, An Sup; Lee, Sang Yoon [Korea Electric Association, Seoul (Korea, Republic of)

    2016-05-15

    Classification of the events for a nuclear power plant is a fundamental basis for defining nuclear safety functions, safety systems performing those functions, and specific acceptance criteria for safety analyses. Presently, the approaches for the event classification adopted by the nuclear suppliers are different, which makes a nuclear technology trade barrier. The IAEA and WENRA are making efforts to establish general requirements or guidelines on the classification of either plant states or defence-in-depth levels for the design of nuclear power plants. However, the requirements and guidelines do not provide the details for practical application to various types of commercial PWRs. Recently, Korea proposed a new ISO standardisation project to develop a harmonized or consolidated international standard for classifying the events in PWRs and for defining (or imposing) the acceptance criteria for reactor design and/or radiation protection corresponding to each event class. This paper briefs the method with strategies for developing the standard, the current various practices of the PWR event classification and acceptance criteria developed or adopted by several organizations in USA and Europe, and a draft of the proposed standard. The proposed standard will affect all the relevant stakeholders such as reactor designers, vendors, suppliers, utilities, regulatory bodies, and publics of the leading countries in the area of nuclear industry as well as utilities, regulatory bodies, and publics of the newly entering (starting) countries. It is expected that all of the stakeholders will benefit from the proposed deliverable which provides an internationally harmonized standard for classifying the PWR events as follows: The reactor design bases for assuring safety and related technical information can be effectively communicated and shared among them resulting in enhancement of the global nuclear safety and fosterage of the global nuclear trade. The countries starting

  15. Design optimization of general arrangement in Korean next generation reactor

    International Nuclear Information System (INIS)

    Kim, S. H.; Jung, D. W.; Choi, Y. B.; Cho, S. J.

    1999-01-01

    In order to optimize the general arrangement(GA) of Korean Next Generation Reactor (KNGR), field opinions in domestic nuclear power plants have been collected, and the bench marking on UCN No.1,2 which were estimated to be the most excellent in view of operability and maintenance has been accomplished. Through this work, design optimization items for GA were reviewed. Major items to be selected for optimization are summarized as follows; 'Expanding the compound building function and the mezzanine floor concept in the auxiliary building', 'Including the diesel generator building to the auxiliary building', 'Change of the equipment removal method in the auxiliary building'. With these GA design optimization, the auxiliary building boundary will be improved as a complete rectangular type. The power block volume except the changing effect to the single containment structure will be reduced to about 10% in comparison with that of in KNGR phase II

  16. General engineering ethics and multiple stress of atomic energy engineering

    International Nuclear Information System (INIS)

    Takeda, Kunihiko

    1999-01-01

    The factors, by which the modern engineering ethics has been profoundly affected, were classified to three categories, namely mental blow, the destruction of human function and environment damage. The role of atomic energy engineering in the ethic field has been shown in the first place. It is pointed out that it has brought about the mental blow by the elucidation of universal truth and discipline and the functional disorder by the power supply. However, the direct effect of radiation to the human kinds is only a part of the stresses comparing to the accumulation of the social stress which should be taken into account of by the possibility of disaster and the suspicion of the atomic energy politics. An increase in the multiple stresses as well as the restriction of criticism will place obstacles on the promotion of atomic energy. (author)

  17. General engineering ethics and multiple stress of atomic energy engineering

    Energy Technology Data Exchange (ETDEWEB)

    Takeda, Kunihiko [Shibaura Inst. of Tech., Tokyo (Japan)

    1999-08-01

    The factors, by which the modern engineering ethics has been profoundly affected, were classified to three categories, namely mental blow, the destruction of human function and environment damage. The role of atomic energy engineering in the ethic field has been shown in the first place. It is pointed out that it has brought about the mental blow by the elucidation of universal truth and discipline and the functional disorder by the power supply. However, the direct effect of radiation to the human kinds is only a part of the stresses comparing to the accumulation of the social stress which should be taken into account of by the possibility of disaster and the suspicion of the atomic energy politics. An increase in the multiple stresses as well as the restriction of criticism will place obstacles on the promotion of atomic energy. (author)

  18. Description of the general properties of pionic atoms

    International Nuclear Information System (INIS)

    Cervantes S, B.R.

    1979-01-01

    The effects of the finite dimension of the nucleus and the strong interaction effects in the energy levels ls and 2p of pionic atoms are considered. The energy transition between these two levels are calculated using the method of perturbations of first order, considering a uniform distribution of nuclear charge, for the effect of the finite dimension of the nucleus and one local optical potential as a model for the strong interaction. The calculations were realized for 13 elements and the results were compared with the experimentally obtained, founding a relative difference around of 4%. In conclusion the author observed that the effects of the finite dimension and the strong interaction can be considered as first order perturbations in light atoms, and for heavy atoms this effects can be considered as of second order or higher perturbations. (author)

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

    International Nuclear Information System (INIS)

    Miyasaka, Yasuhiko

    1979-01-01

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

  20. Selective methods for the maintainability and standardization of the engineering of a research reactor

    International Nuclear Information System (INIS)

    Rico, N.

    1999-01-01

    Full text: The main point of this work consists of a selective method for the engineering of a research reactor based on parameters, which determine a safer design, installation, operation and maintenance. The variety of tasks in a research reactor are: research, development, production of radioisotopes, etc. They are developed within the installation and the different specialties gathered for these activities. It is necessary to count on an intrinsically safe environment, from the point of view of the investigator, the operator and the maintenance personnel. In general, in both nuclear and conventional installation, independent of its size, certain investment necessities prevail, starting from its design, such as: Nuclear Security, Engineering, Versatility, Production (both for investigation and development)), Conventional Security and Physical Protection, Profitability, etc. The concepts which help us accentuate a greater benefit for the research are not found within these parameters, purpose for which this facility was created. When obtaining a simple engineering the results show an increase in security, decrease in maintenance and operative costs, less ageing and an easy operation. The plant engineering of research reactors could be titled, from the engineering and maintenance point of view, as a technological chaos. Not only for its aspect but for its physiognomy too: inaccessible to certain areas; impassable in its circulation aisles; hard to check and measure; disassemble; clean its components; thus increasing unnecessarily the personnel's exposure time. The facilities of research reactors have different disciplines used as rules for the development of the design, such as nuclear, mechanical, thermodynamical, electronic, chemical, electrical, etc. Common guide lines - from design to operation - are non-existent. This is why different manufacturers and models are found within instruments, pumps, electrical engines, illumination, etc. even when they perform

  1. Fusion reactor design studies: standard unit costs and cost scaling rules

    International Nuclear Information System (INIS)

    Schulte, S.C.; Bickford, W.E.; Willingham, C.E.; Ghose, S.K.; Walker, M.G.

    1979-09-01

    This report establishes standard unit costs and scaling rules for estimating costs of material, equipment, land, and labor components used in magnetic confinement fusion reactor plant construction and operation. Use of the standard unit costs and scaling rules will add uniformity to cost estimates, and thus allow valid comparison of the economic characteristics of various reactor concepts

  2. Atomic and magnetic configurational energetics by the generalized perturbation method

    DEFF Research Database (Denmark)

    Ruban, Andrei V.; Shallcross, Sam; Simak, S.I.

    2004-01-01

    in the framework of the Korringa-Kohn-Rostoker method within the atomic sphere and coherent potential approximations. This is demonstrated with calculations of ordering energies, short-range order parameters, and transition temperatures in the CuZn, CuAu, CuPd, and PtCo systems. Furthermore, we show that the GPM...

  3. Dynamical generalization of a solvable family of two-electron model atoms with general interparticle repulsion

    International Nuclear Information System (INIS)

    Niehaus, T A; Suhai, S; March, N H

    2008-01-01

    Holas, Howard and March (2003 Phys. Lett. A 310 451) have obtained analytic solutions for ground-state properties of a whole family of two-electron spin-compensated harmonically confined model atoms whose different members are characterized by a specific interparticle potential energy u(r 12 ). Here, we make a start on the dynamic generalization of the harmonic external potential, the motivation being the serious criticism levelled recently against the foundations of time-dependent density-functional theory (e.g., Schirmer and Dreuw 2007 Phys. Rev. A 75 022513). In this context, we derive a simplified expression for the time-dependent electron density for arbitrary interparticle interaction, which is fully determined by a one-dimensional non-interacting Hamiltonian. Moreover, a closed solution for the momentum space density in the Moshinsky model is obtained

  4. 36 CFR 905.735-201 - General standards of conduct.

    Science.gov (United States)

    2010-07-01

    ... economy; (4) Losing complete independence or impartiality; (5) Making a Government decision outside... CORPORATION STANDARDS OF CONDUCT Conduct and Responsibilities of Employees § 905.735-201 General standards of conduct. (a) All employees shall conduct themselves on the job so as to efficiently discharge the work of...

  5. Review of fast reactor operating experience gained in 1998 in Russia. General trends of future fast reactor development

    International Nuclear Information System (INIS)

    Poplavski, V.M.; Ashurko, Y.M.; Zverev, K.V.; Sarayev, O.M.; Oshkanov, N.N.; Korol'kov, A.S.

    1999-01-01

    Review of the general state of nuclear power in Russia as for 1998 is given in brief in the paper. Results of operation of BR-10, BOR-60 and BN-600 fast reactors are presented as well as of scientific and technological escort of the BN-350 reactor. The paper outlines the current status and prospects of South-Urals and Beloyarskaya power unit projects with the BN-800 reactors. The main planned development trends on fast reactors are described concerning both new projects and R and D works. (author)

  6. Correspondence between the President of the Atomic Energy Organization of Iran and the Director General

    International Nuclear Information System (INIS)

    1984-09-01

    The document includes 5 attachments: the letter of 9 April 1984 from the President of the Atomic Energy Organization of Iran to the Director General, the letter of 11 May 1984 from the Director General to the President of the Atomic Energy Organization of Iran, the text of the telex of 27 May 1984 from the President of the Atomic Energy Organization of Iran to the Director General, the text of the telex of 28 May 1984 from the Director General to the President of the Atomic Energy Organization of Iran and the text of the telex from 30 May 1984 from the President of the Atomic Energy Organization of Iran to the Director General refering to ''Military attack on Iran's Bushehr Nuclear Power Plant''

  7. A subsurface add-on for standard atomic force microscopes

    Energy Technology Data Exchange (ETDEWEB)

    Verbiest, G. J., E-mail: Verbiest@physik.rwth-aachen.de [JARA-FIT and II. Institute of Physics, RWTH Aachen University, 52074 Aachen (Germany); Zalm, D. J. van der; Oosterkamp, T. H.; Rost, M. J., E-mail: Rost@physics.leidenuniv.nl [Kamerlingh Onnes Laboratory, Leiden University, P.O. Box 9504, 2300 RA Leiden (Netherlands)

    2015-03-15

    The application of ultrasound in an Atomic Force Microscope (AFM) gives access to subsurface information. However, no commercially AFM exists that is equipped with this technique. The main problems are the electronic crosstalk in the AFM setup and the insufficiently strong excitation of the cantilever at ultrasonic (MHz) frequencies. In this paper, we describe the development of an add-on that provides a solution to these problems by using a special piezo element with a lowest resonance frequency of 2.5 MHz and by separating the electronic connection for this high frequency piezo element from all other connections. In this sense, we support researches with the possibility to perform subsurface measurements with their existing AFMs and hopefully pave also the way for the development of a commercial AFM that is capable of imaging subsurface features with nanometer resolution.

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  9. General areas needing chemical competence to support reactor operation

    International Nuclear Information System (INIS)

    Proksch, E.; Bildstein, H.

    1963-01-01

    Chemical competence is needed not only for the development of new types of reactors but also for the start-up and safe operation of reactors. The activities of chemistry and chemical engineering cover a number of fields, namely chemical analysis, radiochemical analysis, corrosion research, radiolysis of water and water purification. The author reviews fields in reactor operation and maintenance in which chemical competence is needed. (author). 9 refs

  10. Standard Test Method for Application and Analysis of Helium Accumulation Fluence Monitors for Reactor Vessel Surveillance, E706 (IIIC)

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2007-01-01

    1.1 This test method describes the concept and use of helium accumulation for neutron fluence dosimetry for reactor vessel surveillance. Although this test method is directed toward applications in vessel surveillance, the concepts and techniques are equally applicable to the general field of neutron dosimetry. The various applications of this test method for reactor vessel surveillance are as follows: 1.1.1 Helium accumulation fluence monitor (HAFM) capsules, 1.1.2 Unencapsulated, or cadmium or gadolinium covered, radiometric monitors (RM) and HAFM wires for helium analysis, 1.1.3 Charpy test block samples for helium accumulation, and 1.1.4 Reactor vessel (RV) wall samples for helium accumulation. This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

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

    International Nuclear Information System (INIS)

    Virgilio, M.J.

    1981-11-01

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

  12. The generalized sturmian method for calculating spectra of atoms and ions

    DEFF Research Database (Denmark)

    Avery, James Emil; Avery, John Scales

    2003-01-01

    The properties of generalized Sturmian basis sets are reviewed, and functions of this type are used to perform direct configuration interaction calculations on the spectra of atoms and ions. Singlet excited states calculated in this way show good agreement with experimentally measured spectra. When...... the generalized Sturmian method is applied to atoms, the configurations are constructed from hydrogenlike atomic orbitals with an effective charge which is characteristic of the configuration. Thus, orthonormality between the orbitals of different configurations cannot be assumed, and the generalized Slater...

  13. Rotary reactor for atomic layer deposition on large quantities of nanoparticles

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  14. VALIDATION OF SIMBAT-PWR USING STANDARD CODE OF COBRA-EN ON REACTOR TRANSIENT CONDITION

    Directory of Open Access Journals (Sweden)

    Muhammad Darwis Isnaini

    2016-03-01

    Full Text Available The validation of Pressurized Water Reactor typed Nuclear Power Plant simulator developed by BATAN (SIMBAT-PWR using standard code of COBRA-EN on reactor transient condition has been done. The development of SIMBAT-PWR has accomplished several neutronics and thermal-hydraulic calculation modules. Therefore, the validation of the simulator is needed, especially in transient reactor operation condition. The research purpose is for characterizing the thermal-hydraulic parameters of PWR1000 core, which be able to be applied or as a comparison in developing the SIMBAT-PWR. The validation involves the calculation of the thermal-hydraulic parameters using COBRA-EN code. Furthermore, the calculation schemes are based on COBRA-EN with fixed material properties and dynamic properties that calculated by MATPRO subroutine (COBRA-EN+MATPRO for reactor condition of startup, power rise and power fluctuation from nominal to over power. The comparison of the temperature distribution at nominal 100% power shows that the fuel centerline temperature calculated by SIMBAT-PWR has 8.76% higher result than COBRA-EN result and 7.70% lower result than COBRA-EN+MATPRO. In general, SIMBAT-PWR calculation results on fuel temperature distribution are mostly between COBRA-EN and COBRA-EN+MATPRO results. The deviations of the fuel centerline, fuel surface, inner and outer cladding as well as coolant bulk temperature in the SIMBAT-PWR and the COBRA-EN calculation, are due to the value difference of the gap heat transfer coefficient and the cladding thermal conductivity.

  15. 78 FR 59981 - Proposed Revision to Physical Security-Standard Design Certification and Operating Reactors

    Science.gov (United States)

    2013-09-30

    ... Design Certification and Operating Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Standard... Design Certification and Operating Reactors.'' The NRC seeks comments on the proposed revised section of... subject): Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2013...

  16. General formulation of standard model the standard model is in need of new concepts

    International Nuclear Information System (INIS)

    Khodjaev, L.Sh.

    2001-01-01

    The phenomenological basis for formulation of the Standard Model has been reviewed. The Standard Model based on the fundamental postulates has been formulated. The concept of the fundamental symmetries has been introduced: To look for not fundamental particles but fundamental symmetries. By searching of more general theory it is natural to search first of all global symmetries and than to learn consequence connected with the localisation of this global symmetries like wise of the standard Model

  17. Maintenance experience on reactor recirculation pumps at Tarapur Atomic Power Station

    International Nuclear Information System (INIS)

    Singh, A.K.

    1995-01-01

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

  18. THE GENERAL ATOMICS FUSION THEORY PROGRAM ANNUAL REPORT FOR FISCAL YEAR 2002

    International Nuclear Information System (INIS)

    PROJECT STAFF

    2002-01-01

    OAK B202 THE GENERAL ATOMICS FUSION THEORY PROGRAM ANNUAL REPORT FOR FISCAL YEAR 2002. The dual objective of the fusion theory program at General Atomics (GA) is to significantly advance the scientific understanding of the physics of fusion plasmas and to support the DIII-D and other tokamak experiments. The program plan is aimed at contributing significantly to the Fusion Energy Science and the Tokamak Concept Improvement goals of the Office of Fusion Energy Sciences (OFES)

  19. IAEA Workshop (Training Course) on Codes and Standards for Sodium Cooled Fast Reactors. Working Material

    International Nuclear Information System (INIS)

    2010-01-01

    The training course consisted of lectures and Q&A sessions. The lectures dealt with the history of the development of Design Codes and Standards for Sodium Cooled Fast Reactors (SFRs) in the respective country, the detailed description of the current design Codes and Standards for SFRs and their application to ongoing Fast Reactor design projects, as well as the ongoing development work and plans for the future in this area. Annex 1 contains the detailed Workshop program

  20. International Atomic Energy Agency thirty-third general conference

    International Nuclear Information System (INIS)

    Anon.

    1989-01-01

    A brief account is given of the 33rd session of IAEA General Conference held in Vienna from 25 to 29 September 1989. The Minister for Primary Industry and Energy, Mr John Kerin, led the Australian delegation. His statement to the General Conference highlighted Australia's role as a major uranium exporter, its committment to the Non-Proliferation of Nuclear Weapon Treaty and support for the role which the Agency plays in it. The major Australian policy initiatives in environmental areas were also outlined. Australia continues to make a substantial extra-budgetary contribution to the Regional Cooperation Agreement for Asia and Pacific (RCA), and it has recently expanded its committment by agreeing to support a series of new RCA projects. In this context, the role played by the Australian Nuclear Science and Technology Organization for the provision of technical assistance and cooperation in many areas of the Agency's activities is discussed. ills

  1. Generalized series method in the theory of atomic nucleus

    International Nuclear Information System (INIS)

    Gorbatov, A.M.

    1991-01-01

    On a hypersphere of a prescribed radius the so-called genealogical basis has been constructed. By making use of this basis, the many-body Schroedinger equation has been obtained for bound states of various physical systems. The genealogical series method, being in general outline the extension of the angular potential functions method, deals with the potential harmonics of any generation needed. The new approach provides an exact numerical description of the hadron systems with two-body higher interaction

  2. ANS shielding standards for light-water reactors

    International Nuclear Information System (INIS)

    Trubey, D.K.

    1982-01-01

    The purpose of the American Nuclear Society Standards Subcommittee, ANS-6, Radiation Protection and Shielding, is to develop standards for radiation protection and shield design, to provide shielding information to other standards-writing groups, and to develop standard reference shielding data and test problems. A total of seven published ANS-6 standards are now current. Additional projects of the subcommittee, now composed of nine working groups, include: standard reference data for multigroup cross sections, gamma-ray absorption coefficients and buildup factors, additional benchwork problems for shielding problems and energy spectrum unfolding, power plant zoning design for normal and accident conditions, process radiation monitors, and design for postaccident radiological conditions

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

    International Nuclear Information System (INIS)

    Ramana, M.V.; Saikawa, Eri

    2011-01-01

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

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

    International Nuclear Information System (INIS)

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

    1994-01-01

    The Yankee Atomic Electric Company test irradiation program was implemented to characterize the irradiation response of representative Yankee Rowe reactor vessel beltline plate materials and to remove uncertainties in the analysis of existing irradiation data on the Yankee Rowe reactor vessel steel. Plate materials each containing 0.24 w/o copper, but different nickel contents at 0.63 w/o and 0.19 w/o, were heat treated to simulate the Yankee vessel heat treatment (austenitized at 982 o C (1800 o F)) and to simulate Regulatory Guide 1.99 database materials (austenitized at 871 o C (1600 o F)). These heat treatments produced different microstructures so the effect of microstructure on irradiation damage sensitivity could be tested. Because the nickel content of the test plates varied and the copper level was constant, the effect of nickel on irradiation embrittlement was also tested. Correlation monitor material, HSST-02, was included in the program to benchmark the Ford Nuclear Reactor (University of Michigan Test Reactor) which had never been used before for this type of irradiation program. Materials taken from plate surface locations (versus 1/4 T) were included to test whether or not the improved toughness properties of the plate surface layer, resulting from the rapid quench, are maintained after irradiation. If the improved properties are maintained, pressurized thermal shock calculations could utilize this margin. Finally, for one experiment, irradiations were conducted at two irradiation temperatures (260 o C and 288 o C) to determine the effect of irradiation temperature on embrittlement. (Author)

  5. Automation of the CHARMM General Force Field (CGenFF) I: bond perception and atom typing.

    Science.gov (United States)

    Vanommeslaeghe, K; MacKerell, A D

    2012-12-21

    Molecular mechanics force fields are widely used in computer-aided drug design for the study of drug-like molecules alone or interacting with biological systems. In simulations involving biological macromolecules, the biological part is typically represented by a specialized biomolecular force field, while the drug is represented by a matching general (organic) force field. In order to apply these general force fields to an arbitrary drug-like molecule, functionality for assignment of atom types, parameters, and charges is required. In the present article, which is part I of a series of two, we present the algorithms for bond perception and atom typing for the CHARMM General Force Field (CGenFF). The CGenFF atom typer first associates attributes to the atoms and bonds in a molecule, such as valence, bond order, and ring membership among others. Of note are a number of features that are specifically required for CGenFF. This information is then used by the atom typing routine to assign CGenFF atom types based on a programmable decision tree. This allows for straightforward implementation of CGenFF's complicated atom typing rules and for equally straightforward updating of the atom typing scheme as the force field grows. The presented atom typer was validated by assigning correct atom types on 477 model compounds including in the training set as well as 126 test-set molecules that were constructed to specifically verify its different components. The program may be utilized via an online implementation at https://www.paramchem.org/ .

  6. Safety of power transformers, power supplies, reactors and similar products - Part 1: General requirements and tests

    CERN Document Server

    International Electrotechnical Commission. Geneva

    1998-01-01

    This International Standard deals with safety aspects of power transformers, power supplies, reactors and similar products such as electrical, thermal and mechanical safety. This standard covers the following types of dry-type transformers, power supplies, including switch mode power supplies, and reactors, the windings of which may be encapsulated or non-encapsulated. It has the status of a group safety publication in accordance with IEC Guide 104.

  7. NEPTUNIX, a general program of simulation applied to nuclear reactors

    International Nuclear Information System (INIS)

    Bonnemay, A.; Dansac Bon, V.

    1978-01-01

    Most simulation languages admit an incremental description and involve explicit integration algorithms. NEPTUNIX is a simulation language directly admitting algebraic differential equations under an implicit form, and it involves a very efficient implicit integration method with variable step and order. NEPTUNIX is a tool used for building large systems models in the field of nuclear reactors [fr

  8. Standardized reactors for the study of medical biofilms: a review of the principles and latest modifications.

    Science.gov (United States)

    Gomes, Inês B; Meireles, Ana; Gonçalves, Ana L; Goeres, Darla M; Sjollema, Jelmer; Simões, Lúcia C; Simões, Manuel

    2018-08-01

    Biofilms can cause severe problems to human health due to the high tolerance to antimicrobials; consequently, biofilm science and technology constitutes an important research field. Growing a relevant biofilm in the laboratory provides insights into the basic understanding of the biofilm life cycle including responses to antibiotic therapies. Therefore, the selection of an appropriate biofilm reactor is a critical decision, necessary to obtain reproducible and reliable in vitro results. A reactor should be chosen based upon the study goals and a balance between the pros and cons associated with its use and operational conditions that are as similar as possible to the clinical setting. However, standardization in biofilm studies is rare. This review will focus on the four reactors (Calgary biofilm device, Center for Disease Control biofilm reactor, drip flow biofilm reactor, and rotating disk reactor) approved by a standard setting organization (ASTM International) for biofilm experiments and how researchers have modified these standardized reactors and associated protocols to improve the study and understanding of medical biofilms.

  9. Generalized Rashba-Dresselhaus spin-orbit coupling for cold atoms

    International Nuclear Information System (INIS)

    Juzeliunas, Gediminas; Ruseckas, Julius; Dalibard, Jean

    2010-01-01

    We study the possibility for generating a new type of spin-orbit coupling for the center-of-mass motion of cold atoms, using laser beams that resonantly couple N atomic internal ground states to an extra state. After a general analysis of the scheme, we concentrate on the tetrapod setup (N=4) where the atomic state can be described by a three-component spinor, evolving under the action of a Rashba-Dresselhaus-type spin-orbit coupling for a spin 1 particle. We illustrate a consequence of this coupling by studying the negative refraction of atoms at a potential step and show that the amplitude of the refracted beam is significantly increased in comparison to the known case of spin 1/2 Rashba-Dresselhaus coupling. Finally, we explore a possible implementation of this tetrapod setup, using stimulated Raman couplings between Zeeman sublevels of the ground state of alkali-metal atoms.

  10. The Closed-Orbit Theory for General Rydberg Atoms in External Fields

    International Nuclear Information System (INIS)

    Carboni, R.

    1997-01-01

    The photoabsorption spectra of hydrogen Rydberg atoms, as well of model Rydberg atoms in pure magnetic or electric fields have been successfully calculated using the semiclassical closed-orbit theory. The theory relates the resonances of the spectra to closed classical orbits of the excited electron. The dynamics of multielectron atoms is more complicated than the hydrogenic one; additionally, when the atoms are in the presence of perpendicular magnetic and electric fields becomes more complex than when they are in pure fields, due to the fact that the Hamiltonian is non-separable in three degrees of freedom, instead of two non-separable degrees of freedom. In this work, I present an extension of the closed-orbit theory to three degrees of freedom, considering arbitrary quantum defects, i.e., general atoms. (Author) [es

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

    International Nuclear Information System (INIS)

    Giraudel, B.; Langlois, G.

    1979-01-01

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

  12. Design and analysis of reactor headers for Narora Atomic Power Project

    International Nuclear Information System (INIS)

    Danak, M.R.

    1975-01-01

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

  13. The Atomic energy basic law

    International Nuclear Information System (INIS)

    1979-01-01

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

  14. Neutron importance and the generalized Green function for the conventionally critical reactor with normalized neutron distribution

    International Nuclear Information System (INIS)

    Khromov, V.V.

    1978-01-01

    The notion of neutron importance when applied to nuclear reactor statics problems described by time-independent homogeneous equations of neutron transport with provision for normalization of neutron distribution is considered. An equation has been obtained for the function of neutron importance in a conditionally critical reactor with respect to an arbitrary nons linear functional determined for the normalized neutron distribution. Relation between this function and the generalized Green function of the selfconjugated operator of the reactor equation is determined and the formula of small perturbations for the functionals of a conditionally critical reactor is deduced

  15. General scheme of research reactor mainly for production of fission 99Mo

    International Nuclear Information System (INIS)

    Shen Feng; Liu Xingmin; Wu Xiaochun; Sun Zheng; Guo Chunqiu; Yi Dayong

    2012-01-01

    On the basis of the analysis for current circumstance and development tendency of research reactor mainly for 99 Mo production in the world, the design idea of this sort of research reactor was proposed. By the optimization and basic design, the general design parameters of the reactor were analyzed and testified. The evaluation of output activities of 99 Mo and the analysis of economics were conducted on the basically assumption. It is argued that the economics of this reactor is improved dramatically while the safety is ensured by the analysis. (authors)

  16. Spatial atomic layer deposition on flexible substrates using a modular rotating cylinder reactor

    International Nuclear Information System (INIS)

    Sharma, Kashish; Hall, Robert A.; George, Steven M.

    2015-01-01

    Spatial atomic layer deposition (ALD) is a new version of ALD based on the separation of reactant gases in space instead of time. In this paper, the authors present results for spatial ALD on flexible substrates using a modular rotating cylinder reactor. The design for this reactor is based on two concentric cylinders. The outer cylinder remains fixed and contains a series of slits. These slits can accept a wide range of modules that attach from the outside. The modules can easily move between the various slit positions and perform precursor dosing, purging, or pumping. The inner cylinder rotates with the flexible substrate and passes underneath the various spatially separated slits in the outer cylinder. Trimethyl aluminum and ozone were used to grow Al 2 O 3 ALD films at 40 °C on metallized polyethylene terephthalate (PET) substrates to characterize this spatial ALD reactor. Spectroscopic ellipsometry measurements revealed a constant Al 2 O 3 ALD growth rate of 1.03 Å/cycle with rotation speeds from 40 to 100 RPM with the outer cylinder configured for one Al 2 O 3 ALD cycle per rotation. The Al 2 O 3 ALD growth rate then decreased at higher rotation rates for reactant residence times < 5 ms. The Al 2 O 3 ALD films were also uniform to within <1% across the central portion of metallized PET substrate. Fixed deposition time experiments revealed that Al 2 O 3 ALD films could be deposited at 2.08 Å/s at higher rotation speeds of 175 RPM. Even faster deposition rates are possible by adding more modules for additional Al 2 O 3 ALD cycles for every one rotation of the inner cylinder

  17. Mirror reactor studies

    International Nuclear Information System (INIS)

    Moir, R.W.; Barr, W.L.; Bender, D.J.

    1977-01-01

    Design studies of a fusion mirror reactor, a fusion-fission mirror reactor, and two small mirror reactors are summarized. The fusion reactor uses 150-keV neutral-beam injectors based on the acceleration of negative ions. The injectors provide over 1 GW of continuous power at an efficiency greater than 80%. The fusion reactor has three-stage, modularized, Venetian blind, plasma direct converter with a predicted efficiency of 59% and a new concept for removal of the lune-shaped blanket: a crane is brought between the two halves of the Yin-Yang magnet, which are separated by a float. The design has desirable features such as steady-state operation, minimal impurity problems, and low first-wall thermal stress. The major disadvantage is low Q resulting in high re-circulating power and hence high cost of electrical power. However, the direct capital cost per unit of gross electrical power is reasonable [$1000/kW(e)]. By contrast, the fusion-fission reactor design is not penalized by re-circulating power and uses relatively near-term fusion technology being developed for the fusion power program. New results are presented on the Th- 233 U and the U- 239 Pu fuel cycles. The purpose of this hybrid is fuel production, with projected costs at $55/g of Pu or $127/g of 233 U. Blanket and cooling system designs, including an emergency cooling system, by General Atomic Company, lead us to the opinion that the reactor can meet expected safety standards for licensing. The smallest mirror reactor having only a shield between the plasma and the coil is the 4.2-m long fusion engineering research facility (FERF) designed for material irradiation. The smallest mirror reactor having both a blanket and shield is the 7.5-m long experimental power reactor (EPR), which has both a fusion and a fusion-fission version. (author)

  18. Generalized saddle point condition for ignition in a tokamak reactor with temperature and density profiles

    International Nuclear Information System (INIS)

    Mitari, O.; Hirose, A.; Skarsgard, H.M.

    1989-01-01

    In this paper, the concept of a generalized ignition contour map, is extended to the realistic case of a plasma with temperature and density profiles in order to study access to ignition in a tokamak reactor. The generalized saddle point is found to lie between the Lawson and ignition conditions. If the height of the operation path with Goldston L-mode scaling is higher than the generalized saddle point, a reactor can reach ignition with this scaling for the case with no confinement degradation effect due to alpha-particle heating. In this sense, the saddle point given in a general form is a new criterion for reaching ignition. Peaking the profiles for the plasma temperature and density can lower the height of the generalized saddle point and help a reactor to reach ignition. With this in mind, the authors can judge whether next-generation tokamaks, such as Compact Ignition Tokamak, Tokamak Ignition/Burn Experimental Reactor, Next European Torus, Fusion Experimental Reactor, International Tokamak Reactor, and AC Tokamak Reactor, can reach ignition with realistic profile parameters and an L-mode scaling law

  19. General considerations in fuel management for thermal reactors

    International Nuclear Information System (INIS)

    Tyror, J.G.; Fayers, F.J.

    1971-07-01

    By fuel management we mean the strategy for fuelling and refuelling a reactor together with any associated absorber movements. It incorporates (a) decisions made about the timing of fuel loading operations; (b) choice of enrichments to be loaded; (c) selection of sites at which reloading occurs; (d) programming of control rods and any other reactivity control facilities such as soluble or burnable poisons; and (e) evaluation of the resulting fuel element performance consequences. The topic of fuel management is thus a vast and vital one. It embraces most of the various aspects of core performance and determines many of a reactor's design characteristics. In this paper we review what to us appear to be some of the important issues in this important field

  20. General considerations for neutron capture therapy at a reactor facility

    International Nuclear Information System (INIS)

    Binney, S.E.

    2001-01-01

    In addition to neutron beam intensity and quality, there are also a number of other significant criteria related to a nuclear reactor that contribute to a successful neutron capture therapy (NCT) facility. These criteria are classified into four main categories: Nuclear design factors, facility management and operations factors, facility resources, and non-technical factors. Important factors to consider are given for each of these categories. In addition to an adequate neutron beam intensity and quality, key requirements for a successful neutron capture therapy facility include necessary finances to construct or convert a facility for NCT, a capable medical staff to perform the NCT, and the administrative support for the facility. The absence of any one of these four factors seriously jeopardizes the overall probability of success of the facility. Thus nuclear reactor facility management considering becoming involved in neutron capture therapy, should it be proven clinically successful, should take all these factors into consideration. (author)

  1. Feasibility of ISO 14000 certification for the RA-6 reactor of the Bariloche Atomic Center

    International Nuclear Information System (INIS)

    Gho, Carlos J.

    2003-01-01

    The strengths and weaknesses of a proposal to set up a System for the Environmental Management of RA-6 reactor are reviewed. With this aim: The facility, its surroundings, its institutional frame, and the links of both the National Atomic Energy Commission and the Bariloche Atomic Center with the environmental issues are described. The RA-6 past and present regarding the environment is analyzed. The existence of an abundant documentation on procedures and manuals for administrative and technical management, as well as records of environmental behavior is determined. A summary is made of the work done with the staff in order to assess their motivation to develop this type of initiative, and their degree of knowledge and awareness of environmental issues. It is worthy to point out the high professional training of the RA-6 staff. This information is analyzed under the point of view of ISO 14001 both text and philosophy. It is shown that there exist important strengths that enable to face a concrete project, and that the weaknesses are few, most of them can be overcome somehow easily. The opportunity is remarkable. Conclusion is made that RA-6 is not only in optimal conditions to face successfully the implementation of an Environmental Management System, but it has already a rudimentary one

  2. Proposed Advanced Reactor Adaptation of the Standard Review Plan NUREG-0800 Chapter 4 (Reactor) for Sodium-Cooled Fast Reactors and Modular High-Temperature Gas-Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Belles, Randy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Poore, III, Willis P. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Flanagan, George F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Holbrook, Mark [Idaho National Lab. (INL), Idaho Falls, ID (United States); Moe, Wayne [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sofu, Tanju [Argonne National Lab. (ANL), Argonne, IL (United States)

    2017-03-01

    This report proposes adaptation of the previous regulatory gap analysis in Chapter 4 (Reactor) of NUREG 0800, Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light Water Reactor] Edition. The proposed adaptation would result in a Chapter 4 review plan applicable to certain advanced reactors. This report addresses two technologies: the sodium-cooled fast reactor (SFR) and the modular high temperature gas-cooled reactor (mHTGR). SRP Chapter 4, which addresses reactor components, was selected for adaptation because of the possible significant differences in advanced non-light water reactor (non-LWR) technologies compared with the current LWR-based description in Chapter 4. SFR and mHTGR technologies were chosen for this gap analysis because of their diverse designs and the availability of significant historical design detail.

  3. Relativistic Spin-Orbit Heavy Atom on the Light Atom NMR Chemical Shifts: General Trends Across the Periodic Table Explained.

    Science.gov (United States)

    Vícha, Jan; Komorovsky, Stanislav; Repisky, Michal; Marek, Radek; Straka, Michal

    2018-05-10

    The importance of relativistic effects on the NMR parameters in heavy-atom (HA) compounds, particularly the SO-HALA (Spin-Orbit Heavy Atom on the Light Atom) effect on NMR chemical shifts, has been known for about 40 years. Yet, a general correlation between the electronic structure and SO-HALA effect has been missing. By analyzing 1 H NMR chemical shifts of the sixth-period hydrides (Cs-At), we discovered general electronic-structure principles and mechanisms that dictate the size and sign of the SO-HALA NMR chemical shifts. In brief, partially occupied HA valence shells induce relativistic shielding at the light atom (LA) nuclei, while empty HA valence shells induce relativistic deshielding. In particular, the LA nucleus is relativistically shielded in 5d 2 -5d 8 and 6p 4 HA hydrides and deshielded in 4f 0 , 5d 0 , 6s 0 , and 6p 0 HA hydrides. This general and intuitive concept explains periodic trends in the 1 H NMR chemical shifts along the sixth-period hydrides (Cs-At) studied in this work. We present substantial evidence that the introduced principles have a general validity across the periodic table and can be extended to nonhydride LAs. The decades-old question of why compounds with occupied frontier π molecular orbitals (MOs) cause SO-HALA shielding at the LA nuclei, while the frontier σ MOs cause deshielding is answered. We further derive connection between the SO-HALA NMR chemical shifts and Spin-Orbit-induced Electron Deformation Density (SO-EDD), a property that can be obtained easily from differential electron densities and can be represented graphically. SO-EDD provides an intuitive understanding of the SO-HALA effect in terms of the depletion/concentration of the electron density at LA nuclei caused by spin-orbit coupling due to HA in the presence of a magnetic field. Using an analogy between the SO-EDD concept and arguments from classic NMR theory, the complex question of the SO-HALA NMR chemical shifts becomes easily understandable for a wide

  4. Operator licensing examination standards for power reactors. Interim revision 8

    International Nuclear Information System (INIS)

    1997-01-01

    These examination standards are intended to assist NRC examiners and facility licensees to better understand the processes associated with initial and requalification examinations. The standards also ensure the equitable and consistent administration of examinations for all applicants. These standards are for guidance purposes and are not a substitute for the operator licensing regulations (i.e., 10 CFR Part 55), and they are subject to revision or other changes in internal operator licensing policy. This interim revision permits facility licensees to prepare their initial operator licensing examinations on a voluntary basis pending an amendment to 10 CFR Part 55 that will require facility participation. The NRC intends to solicit comments on this revision during the rulemaking process and to issue a final Revision 8 in conjunction with the final rule

  5. Testing general relativity and alternative theories of gravity with space-based atomic clocks and atom interferometers

    Directory of Open Access Journals (Sweden)

    Bondarescu Ruxandra

    2015-01-01

    Full Text Available The successful miniaturisation of extremely accurate atomic clocks and atom interferometers invites prospects for satellite missions to perform precision experiments. We discuss the effects predicted by general relativity and alternative theories of gravity that can be detected by a clock, which orbits the Earth. Our experiment relies on the precise tracking of the spacecraft using its observed tick-rate. The spacecraft’s reconstructed four-dimensional trajectory will reveal the nature of gravitational perturbations in Earth’s gravitational field, potentially differentiating between different theories of gravity. This mission can measure multiple relativistic effects all during the course of a single experiment, and constrain the Parametrized Post-Newtonian Parameters around the Earth. A satellite carrying a clock of fractional timing inaccuracy of Δ f / f ∼ 10−16 in an elliptic orbit around the Earth would constrain the PPN parameters |β − 1|, |γ − 1| ≲ 10−6. We also briefly review potential constraints by atom interferometers on scalar tensor theories and in particular on Chameleon and dilaton models.

  6. Conversion and standardization of university reactor fuels using low-enrichment uranium - Options and costs

    International Nuclear Information System (INIS)

    Harris, D.R.; Matos, J.E.; Young, H.H.

    1985-01-01

    The highly-enriched uranium (HEU) fuel used in twenty United States university reactors can be viewed as contributing to the risk of theft or diversion of weapons-useable material. The U.S. Nuclear Regulatory Commission has issued a policy statement expressing its concern and has published a proposed rule on limiting the use of HEU in NRC-licensed non-power reactors. The fuel options, functional impacts, licensing, and scheduling of conversion and standardization of these reactor fuels to use of low-enrichment uranium (LEU) have been assessed. The university reactors span a wide range in form and function, from medium-power intense neutron sources where HEU fuel may be required, to low-power training and research facilities where HEU fuel is unnecessary. Conversion provides an opportunity to standardize university reactor fuels and improve reactor utilization in some cases. The entire program is estimated to cost about $10 million and to last about five years. Planning for conversion and standardization is facilitated by the U.S. Department of Energy. (author)

  7. Conversion and standardization of university reactor fuels using low-enrichment uranium - options and costs

    International Nuclear Information System (INIS)

    Harris, D.R.; Matos, J.E.; Young, H.H.

    1985-01-01

    The highly-enriched uranium (HEU) fuel used in twenty United States university reactors can be viewed as contributing to the risk of theft or diversion of weapons-useable material. The US Nuclear Regulatory Commission has issued a policy statement expressing its concern and has published a proposed rule on limiting the use of HEU in NRC-licensed non-power reactors. The fuel options, functional impacts, licensing, and scheduling of conversion and standardization of these reactor fuels to use of low-enrichment uranium (LEU) have been assessed. The university reactors span a wide range in form and function, from medium-power intense neutron sources where HEU fuel may be required, to low-power training and research facilities where HEU fuel is unnecessary. Conversion provides an opportunity to standardize university reactor fuels and improve reactor utilization in some cases. The entire program is estimated to cost about $10 million and to last about five years. Planning for conversion and standardization is facilitated by the US Department of Energy. 20 refs., 1 tab

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

    International Nuclear Information System (INIS)

    1995-01-01

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

  9. Present status and improvement approach of atomic energy laws and safety standards

    International Nuclear Information System (INIS)

    Oh, B. J.; An, H. J.; Kim, S. W.; Kim, C. B.; Kang, S. C.; Lee, J. I.

    2000-01-01

    Major revision to the atomic energy act, which is currently undergoing are introduced: increase of members of nuclear safety commission, adoption of standard design certification, periodic safety review, production license system of radioactive isotope facilities, preparation for implementation of IAEA convention. Improvement of the notice of ministry of science and technology are discussed in accordance with the new atomic energy act, enforcement detect, and enforcement regulations, whose revision were completed in May 2000. Allocation of the code number to the notice, development procedures for the safety and regulatory guides are also introduced

  10. Colombeau's generalized functions and non-standard analysis

    International Nuclear Information System (INIS)

    Todorov, T.D.

    1987-10-01

    Using some methods of the Non-Standard Analysis we modify one of Colombeau's classes of generalized functions. As a result we define a class ε-circumflex of the so-called meta-functions which possesses all good properties of Colombeau's generalized functions, i.e. (i) ε-circumflex is an associative and commutative algebra over the system of the so-called complex meta-numbers C-circumflex; (ii) Every meta-function has partial derivatives of any order (which are meta-functions again); (iii) Every meta-function is integrable on any compact set of R n and the integral is a number from C-circumflex; (iv) ε-circumflex contains all tempered distributions S', i.e. S' is contained in ε' isomorphically with respect to all linear operations (including the differentiation). Thus, within the class ε-circumflex the problem of multiplication of the tempered distributions is satisfactorily solved (every two distributions in S' have a well-defined product in ε-circumflex). The crucial point is that C-circumflex is a field in contrast to the system of Colombeau's generalized numbers C-bar which is a ring only (C-bar is the counterpart of C-circumflex in Colombeau's theory). In this way we simplify and improve slightly the properties of the integral and notion of ''values of the meta-functions'' as well as the properties of the whole class ε-circumflex itself if compared with the original Colombeau theory. And, what is maybe more important, we clarify the connection between the Non-Standard Analysis and Colombeau's theory of new generalized functions in the framework of which the problem of multiplication of distributions was recently solved. (author). 14 refs

  11. Generalized oscillator strengths for some higher valence-shell excitations of krypton atom

    Institute of Scientific and Technical Information of China (English)

    2007-01-01

    The valence-shell excitations of krypton atom have been investigated by fast electron impact with an angle-resolved electron-energy-loss spectrometer. The generalized oscillator strengths for some higher mixed valence-shell excitations in 4d, 4f, 5p, 5d, 6s, 6p, 7s ← 4p of krypton atom have been determined. Their profiles are discussed, and the generalized oscillator strengths for the electric monopole and quadrupole excitations in 5p ← 4p are compared with the calculations of Amusia et al. (Phys. Rev. A 67 022703 (2003)). The differences between the experimental results and theoretical calculations show that more studies are needed.

  12. Evolution of general design requirements for french pressurized water reactors

    International Nuclear Information System (INIS)

    Gros, G.; Jalouneix, J.; Rollinger, F.

    1988-10-01

    The design of French pressurized water reactors is based first on deterministic principles, using the well-known defense in depth concept. This safety approach, basically reflected current American practice at that time, which consisted notably in designing engineered safeguard systems capable of limiting the consequences of accidents assumed to be credible despite the preventive measures taken. Further reflections have led to complete this approach, resulting in modifications to regulatory practice, mainly related to better practical assimilation of the problems arising during plant unit operation and reactor control after an accident and to the determination to enhance the overall consistency of the safety approach. As regards system redundancy, it should be noted that common cause failures can result in the total loss of a redundant system. System redundancy aspects will be dealt with in Chapter 2. As regards study of design basis accidents, attention was focused on the human intervention stage following automatic activation of protection and safeguard systems. This resulted, for all plant units, in the revision of operating procedures, accompanied by examination of the means required for their implementation. These subjects will be discussed in Chapter 3. Finally, as regards equipment classification, the range of equipment subjected to particular requirements, formerly limited to design basis safety classified equipment, was enlarged to include important for safety equipment. This subject will be dealt with in Chapter 5

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

    International Nuclear Information System (INIS)

    1987-01-01

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

  14. Generalized atomic processes for interaction of intense femtosecond XUV- and X-ray radiation with solids

    International Nuclear Information System (INIS)

    Deschaud, B.; Peyrusse, O.; Rosmej, F.B.

    2014-01-01

    Generalized atomic processes are proposed to establish a consistent description from the free-atom approach to the heated and even up to the cold solid. It is based on a rigorous introduction of the Fermi-Dirac statistics, Pauli blocking factors and on the respect of the principle of detailed balance via the introduction of direct and inverse processes. A probability formalism driven by the degeneracy of the free electrons enables to establish a link of atomic rates valid from the heated atom up to the cold solid. This allows to describe photoionization processes in atomic population kinetics and subsequent solid matter heating on a femtosecond time scale. The Auger effect is linked to the 3-body recombination via a generalized 3-body recombination that is identified as a key mechanism, along with the collisional ionization, that follows energy deposition by photoionization of inner shells when short, intense and high-energy radiation interacts with matter. Detailed simulations are carried out for aluminum that highlight the importance of the generalized approach. (authors)

  15. Assessment of United States industry structural codes and standards for application to advanced nuclear power reactors: Appendices. Volume 2

    International Nuclear Information System (INIS)

    Adams, T.M.; Stevenson, J.D.

    1995-10-01

    Throughout its history, the USNRC has remained committed to the use of industry consensus standards for the design, construction, and licensing of commercial nuclear power facilities. The existing industry standards are based on the current class of light water reactors and as such may not adequately address design and construction features of the next generation of Advanced Light Water Reactors and other types of Advanced Reactors. As part of their on-going commitment to industry standards, the USNRC commissioned this study to evaluate US industry structural standards for application to Advanced Light Water Reactors and Advanced Reactors. The initial review effort included (1) the review and study of the relevant reactor design basis documentation for eight Advanced Light Water Reactors and Advanced Reactor Designs, (2) the review of the USNRCs design requirements for advanced reactors, (3) the review of the latest revisions of the relevant industry consensus structural standards, and (4) the identification of the need for changes to these standards. The results of these studies were used to develop recommended changes to industry consensus structural standards which will be used in the construction of Advanced Light Water Reactors and Advanced Reactors. Over seventy sets of proposed standard changes were recommended and the need for the development of four new structural standards was identified. In addition to the recommended standard changes, several other sets of information and data were extracted for use by USNRC in other on-going programs. This information included (1) detailed observations on the response of structures and distribution system supports to the recent Northridge, California (1994) and Kobe, Japan (1995) earthquakes, (2) comparison of versions of certain standards cited in the standard review plan to the most current versions, and (3) comparison of the seismic and wind design basis for all the subject reactor designs

  16. Nuclear reactors' construction costs: The role of lead-time, standardization and technological progress

    International Nuclear Information System (INIS)

    Berthelemy, Michel; Escobar Rangel, Lina

    2013-01-01

    This paper provides the first comparative analysis of nuclear reactor construction costs in France and the United States. Studying the cost of nuclear power has often been a challenge, owing to the lack of reliable data sources and heterogeneity between countries, as well as the long time horizon which requires controlling for input prices and structural changes. We build a simultaneous system of equations for overnight costs and construction time (lead-time) to control for endogeneity, using expected demand variation as an instrument. We argue that benefits from nuclear reactor program standardization can arise through short term coordination gains, when the diversity of nuclear reactors' technologies under construction is low, or through long term benefits from learning spillovers from past reactor construction experience, if those spillovers are limited to similar reactors. We find that overnight construction costs benefit directly from learning spillovers but that these spillovers are only significant for nuclear models built by the same Architect-Engineer (A- E). In addition, we show that the standardization of nuclear reactors under construction has an indirect and positive effect on construction costs through a reduction in lead-time, the latter being one of the main drivers of construction costs. Conversely, we also explore the possibility of learning by searching and find that, contrary to other energy technologies, innovation leads to construction costs increases. (authors)

  17. The search For Closed Orbits Of General Rydberg Atoms in External Fields And Their Classification

    International Nuclear Information System (INIS)

    Carboni, R.

    1997-01-01

    A program of high precision that find closed orbits for the classical motion of the electron of general Rydberg atoms in crossed magnetic and electric fields is explained. Investigations of the influence of the ionic core on the electronic trajectories using a phenomenological model potential were done. Additional closed orbits that are not present in hydrogen atoms and that seem to be composed of hydrogenic orbits were found. The stability and formation of orbits are explained. Using the generalized closed-orbit theory, the scaled recurrence spectra for rubidium Rydberg atoms were calculated. The results are in good agreement with reported experiments. Two important features of the expectra can be explained by classical core scattering: The additional non-hydrogenic resonances associated to composite orbits and the vanishing of hydrogenic resonances related to closed or whose trajectories approach the core. (Author) [es

  18. General Atomic reprocessing pilot plant: description and results of initial testing

    International Nuclear Information System (INIS)

    1977-12-01

    In June 1976 General Atomic completed the construction of a reprocessing head-end cold pilot plant. In the year since then, each system within the head end has been used for experiments which have qualified the designs. This report describes the equipment in the plant and summarizes the results of the initial phase of reprocessing testing

  19. Atoms-First Curriculum: A Comparison of Student Success in General Chemistry

    Science.gov (United States)

    Esterling, Kevin M.; Bartels, Ludwig

    2013-01-01

    We present an evaluation of the impact of an atoms-first curriculum on student success in introductory chemistry classes and find that initially a lower fraction of students obtain passing grades in the first and second quarters of the general chemistry series. This effect is more than reversed for first-quarter students after one year of…

  20. Atomic Densities, Polarizabilities, and Natural Orbitals Derived from Generalized Sturmian Calculations

    DEFF Research Database (Denmark)

    Avery, John Scales; Avery, James Emil; Aquilanti, Vincenzo

    2004-01-01

    The generalized Sturmian method for atomic and molecular electronic structure calculations is a direct configuration interaction method in which the configurations are chosen to be isoenergetic solutions of an approximate N-electron Schrödinger equation with a weighted potential, $\\beta_\

  1. THE GENERAL ATOMICS FUSION THEORY PROGRAM ANNUAL REPORT FOR GRANT YEAR 2004

    International Nuclear Information System (INIS)

    PROJECT STAFF

    2004-01-01

    The dual objective of the fusion theory program at General Atomics (GA) is to significantly advance our scientific understanding of the physics of fusion plasmas and to support the DIII-D and other tokamak experiments. The program plan is aimed at contributing significantly to the Fusion Energy Science and the Tokamak Concept Improvement goals of the Office of Fusion Energy Sciences (OFES)

  2. Telex message to the Director General from the President of the Atomic Energy Organization of Iran

    International Nuclear Information System (INIS)

    1988-08-01

    The document reproduces the text of a telex message dated 19 July 1988 from the Deputy Prime Minister of the Islamic Republic of Iran and President of the Atomic Energy Organization of Iran to the Director General of the IAEA and the Director General's reply dated 29 July 1988, in connection with an Iraqi military attack of the Busher Nuclear Power Plant on 18 July 1988

  3. Standard technical specifications for Babcock and Wilcox pressurized water reactors

    International Nuclear Information System (INIS)

    Virgilio, M.

    1979-07-01

    This Standard Technical Specification (STS) has been structured for the broadest possible use on B and W NSSS plants currently being reviewed for an Operating License. Two separate and discrete containment specification sections are provided for each of the following containment types: Atmospheric and Dual. Optional specifications are provided for those features and systems which may be included in individual plant designs but are not generic in their scope of application. Alternate specifications are provided in a limited number of cases to cover situations where alternate specification requirements are necessary on a generic basis because of design differences. This revision of STS does not typically include requirements which may be added or revised as a result of the NRC staff's further review of the Three Mile Island incident

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

    International Nuclear Information System (INIS)

    1978-06-01

    The Standard Technical Specification (STS) has been structured for the broadest possible use on B and W NSSS plants currently being reviewed for an Operating License. Two separate and discrete containment specification sections are provided for each of the following containment types: Atmospheric, and Dual. Optional specifications are provided for those features and systems which may be included in individual plant designs but are not generic in their scope of application. Alternate specifications are provided in a limited number of cases to cover situations where alternate specification requirements are necessary on a generic basis because of design differences. The format of the STS addresses the categories required by 10 CFR 50 and consists of six sections covering the areas of: Definitions, Safety Limits and Limiting Safety System Settings, Limiting Conditions for Operation, Surveillance Requirements, Design Features, and Administrative Controls

  5. Standard technical specifications for combustion engineering pressurized water reactors

    International Nuclear Information System (INIS)

    1979-08-01

    This Standard Technical Specification (STS) has been structured for the broadest possible use on Combustion Engineering plants currently being reviewed for an Operating License. Two separate and discrete containment specification sections are provided for each of the following containment types: Atmospheric and Dual. Optional specifications are provided for those features and systems which may be included in individual plant designs but are not generic in their scope of application. Alternate specifications are provided in a limited number of cases to cover situations where alternate specification requirements are necessary on a generic basis because of design differences. This revision of STS does not typically include requirements which may be added or revised as a result of the NRC staff's further review of the Three Mile Island incident

  6. Standard technical specifications for Westinghouse pressurized water reactors

    International Nuclear Information System (INIS)

    Wagner, P.C.

    1979-07-01

    This Standard Technical Specification (STS) has been structured for the broadest possible use on Westinghouse plants currently being reviewed for an Operating License. Accordingly, the document contains specifications applicable to plants with (1) either 3 or 4 loops and (2) with and without loop stop valves. In addition, four separate and discrete containment specification sections are provided for each of the following containment types: Atmospheric, Ice Condenser, Sub-Atmospheric, and Dual. Optional specifications are provided for those features and systems which may be included in individual plant designs but are not generic in their scope of application. Alternate specifications are provided in a limited number of cases to cover situations where alternate specification requirements are necessary on a generic basis because of design differences. This revision of the STS does not typically include requirements which may be added or revised as a result of the NRC staff's further review of the Three Mile Island incident

  7. On the Power Dependence of Extraneous Microwave Fields in Atomic Frequency Standards

    Science.gov (United States)

    2005-01-01

    uncertainty”, Metrologia 35 (1998) pp. 829-845. [6] K. Dorenwendt and A. Bauch, “Spurious Microwave Fields in Caesium Atomic Beam Standards...Cesium Beam Clocks Induced by Microwave Leakages”, IEEE Trans. UFFC 45 (1998)728-738. [8] M. Abgrall, “Evaluation des Performances de la Fontaine...Proc of the EFTF 2005 – in press. [12] A. DeMarchi, “The Optically Pumped Caesium Fountain: 10-15 Frequency Accuracy?”, Metrologia 18 (1982) pp

  8. Atom

    International Nuclear Information System (INIS)

    Auffray, J.P.

    1997-01-01

    The atom through centuries, has been imagined, described, explored, then accelerated, combined...But what happens truly inside the atom? And what are mechanisms who allow its stability? Physicist and historian of sciences, Jean-Paul Auffray explains that these questions are to the heart of the modern physics and it brings them a new lighting. (N.C.)

  9. The French Atomic Energy Commission program in the field of reactor instrumentation and control

    International Nuclear Information System (INIS)

    Golinelli, C.; Bernard, P.; Thomas, J.B.

    1992-01-01

    The worldwide slowing-down of the nuclear reactor construction must not lead to decrease the Research and Development effort. Particularly, in the field of the Nuclear Instrumentation and Control, new technologies are quickly changing: sensors, electronics, optronics, computer science... The nuclear industry is reluctant to the introduction of new concepts and of sophisticated technologies. Safety requires highly qualified systems. The development process must respect each step: - interest evaluation of the new idea, - designing and manufacturing of a prototype equipment, - qualification using an experimental facility or with a simulator, - qualification in operational condition (reliability, ageing, accidental standards...). We present an overview of the French CEA program covering the IC domain from the sensors to the operator screen

  10. Plan and procedures for rapid inventory taking at the Research Institute for Atomic Reactors

    International Nuclear Information System (INIS)

    Kalygin, V.; Gorobets, A.; Karlov, S.; Suda, S.C.; Bonner, M.F.; Satkowiak, L.

    1997-01-01

    A major element of a system for nuclear material protection, control, and accounting (MPC and A) is to take the physical inventory of the nuclear material periodically. Physical inventory taking (PIT) includes ensuring that all nuclear material on inventory is included in the facility records and that the measured content of items or containers corresponds to the recorded values. A preliminary step to the conduct of the PIT is application of rapid inventory procedures that serve to provide the benchmark for the inventory, e.g., by identifying if any items are missing and also, if any unrecorded items are present. The Rapid Inventory approach is being implemented by the Research Institute for Atomic Reactors (RIAR) in Dimitrovgrad, Russia, as one of the first steps in the program to enhance nuclear materials safeguards at the site. This effort is being conducted under the US-Russian Cooperative Program on Nuclear Materials Protection. Control And Accounting (MPC and A), with assistance provided by specialists from US Department of Energy National Laboratories. This paper summarizes the features of the existing physical inventory system at RIAR, discusses the upgrades being introduced, and provides some observations on the technology transfer process with regard to the safeguards program

  11. Fluidized bed coupled rotary reactor for nanoparticles coating via atomic layer deposition

    Energy Technology Data Exchange (ETDEWEB)

    Duan, Chen-Long; Liu, Xiao; Chen, Rong, E-mail: rongchen@mail.hust.edu.cn, E-mail: bshan@mail.hust.edu.cn [State Key Laboratory of Digital Manufacturing Equipment and Technology, School of Mechanical Science and Engineering, Huazhong University of Science and Technology, 1037 Luoyu Road, Wuhan, Hubei 430074 (China); Shan, Bin, E-mail: rongchen@mail.hust.edu.cn, E-mail: bshan@mail.hust.edu.cn [State Key Laboratory of Material Processing and Die & Mould Technology, School of Materials Science and Engineering, Huazhong University of Science and Technology, 1037 Luoyu Road, Wuhan, Hubei 430074 (China)

    2015-07-15

    A fluidized bed coupled rotary reactor has been designed for coating on nanoparticles (NPs) via atomic layer deposition. It consists of five major parts: reaction chamber, dosing and fluidizing section, pumping section, rotary manipulator components, as well as a double-layer cartridge for the storage of particles. In the deposition procedure, continuous fluidization of particles enlarges and homogenizes the void fraction in the particle bed, while rotation enhances the gas-solid interactions to stabilize fluidization. The particle cartridge presented here enables both the fluidization and rotation acting on the particle bed, demonstrated by the analysis of pressure drop. Moreover, enlarged interstitials and intense gas–solid contact under sufficient fluidizing velocity and proper rotation speed facilitate the precursor delivery throughout the particle bed and consequently provide a fast coating process. The cartridge can ensure precursors flowing through the particle bed exclusively to achieve high utilization without static exposure operation. By optimizing superficial gas velocities and rotation speeds, minimum pulse time for complete coating has been shortened in experiment, and in situ mass spectrometry showed the precursor usage can reach 90%. Inductively coupled plasma-optical emission spectroscopy results suggested a saturated growth of nanoscale Al{sub 2}O{sub 3} films on spherical SiO{sub 2} NPs. Finally, the uniformity and composition of the shells were characterized by high angle annular dark field-transmission electron microscopy and energy dispersive X-ray spectroscopy.

  12. Plan and procedures for rapid inventory taking at the research institute for atomic reactors

    International Nuclear Information System (INIS)

    Kalygin, V.; Gorobets, A.; Karlov, S.; Suda, S.C.; Bonner, F.M.

    1999-01-01

    A major element of a system for nuclear material protection, control, and accounting (MPC and A) is to take the physical inventory of the nuclear material periodically. Physical inventory taking (PIT) includes ensuring that all nuclear material on inventory is included in the facility records and that the measured content of items or containers (or at least a suitable random sample thereof) corresponds to the recorded values. A preliminary step to the conduct of the PIT is application of rapid inventory procedures that serve to provide the benchmark for the inventory, e.g., by identifying if any items are missing and also, if any unrecorded items are present. The Rapid Inventory approach is being implemented by the Research Institute for Atomic Reactors (RIAR) in Dimitrovgrad, Russia, as one of the first steps in the program to enhance nuclear materials safeguards at the site. This effort is being conducted under the US-Russian Cooperative Program on Nuclear Materials Protection, Control And Accounting (MPC and A), with assistance provided by specialists from U.S. Department of Energy National Laboratories. This paper summarizes the features of the existing physical inventory system at RIAR, discusses the upgrades being introduced, and provides some observations on the technology transfer process with regard to the safeguards program. (author)

  13. Physical characteristics of GE [General Electric] BWR [boiling-water reactor] fuel assemblies

    International Nuclear Information System (INIS)

    Moore, R.S.; Notz, K.J.

    1989-06-01

    The physical characteristics of fuel assemblies manufactured by the General Electric Company for boiling-water reactors are classified and described. The classification into assembly types is based on the GE reactor product line, the Characteristics Data Base (CDB) assembly class, and the GE fuel design. Thirty production assembly types are identified. Detailed physical data are presented for each assembly type in an appendix. Descriptions of special (nonstandard) fuels are also reported. 52 refs., 1 fig., 6 tabs

  14. Single failure effects of reactor coolant system large bore hydraulic snubbers for Korean Standard Nuclear Power Plant

    International Nuclear Information System (INIS)

    Choi, T.S.; Park, S.H.; Sung, K.K.; Kim, T.W.; Jheon, J.H.

    1996-01-01

    A potential snubber single failure is one of the safety significances identified in General Safety Issue 113 for Large Bore Hydraulic Snubber (LBHS) dynamic qualification. This paper investigates dynamic structural effects of single failures of the steam generator and reactor coolant pump snubbers in Korean Standard Nuclear Power Plant by performing the time history dynamic analyses for the reactor coolant system under seismic and postulated pipe break events. The seismic input motions considered are the synthesized ground time histories conforming to SRP 3.7.1, and he postulated pipe break input loadings result from steam generator main seam line and feedwater line pipe breaks which govern pipe breaks remaining after applying LBB to the main coolant line and primary side ranch lines equal to and greater than 12 inch nominal pipe size

  15. Nuclear reactors' construction costs: The role of lead-time, standardization and technological progress

    International Nuclear Information System (INIS)

    Berthélemy, Michel; Escobar Rangel, Lina

    2015-01-01

    This paper provides an econometric analysis of nuclear reactor construction costs in France and the United States based on overnight costs data. We build a simultaneous system of equations for overnight costs and construction time (lead-time) to control for endogeneity, using change in expected electricity demand as instrument. We argue that the construction of nuclear reactors can benefit from standardization gains through two channels. First, short term coordination benefits can arise when the diversity of nuclear reactors' designs under construction is low. Second, long term benefits can occur due to learning spillovers from past constructions of similar reactors. We find that construction costs benefit directly from learning spillovers but that these spillovers are only significant for nuclear models built by the same Architect–Engineer. In addition, we show that the standardization of nuclear reactors under construction has an indirect and positive effect on construction costs through a reduction in lead-time, the latter being one of the main drivers of construction costs. Conversely, we also explore the possibility of learning by searching and find that, contrary to other energy technologies, innovation leads to construction costs increases. -- Highlights: •This paper analyses the determinants of nuclear reactors construction costs and lead-time. •We study short term (coordination gains) and long term (learning by doing) benefits of standardization in France and the US. •Results show that standardization of nuclear programs is a key factor for reducing construction costs. •We also suggest that technological progress has contributed to construction costs escalation

  16. Piping benchmark problems for the General Electric Advanced Boiling Water Reactor

    International Nuclear Information System (INIS)

    Bezler, P.; DeGrassi, G.; Braverman, J.; Wang, Y.K.

    1993-08-01

    To satisfy the need for verification of the computer programs and modeling techniques that will be used to perform the final piping analyses for an advanced boiling water reactor standard design, three benchmark problems were developed. The problems are representative piping systems subjected to representative dynamic loads with solutions developed using the methods being proposed for analysis for the advanced reactor standard design. It will be required that the combined license holders demonstrate that their solutions to these problems are in agreement with the benchmark problem set

  17. Twenty years of radioisotope production from Institute of Atomic Energy reactors

    International Nuclear Information System (INIS)

    Lun, Xiao

    1980-01-01

    The heavy water reactor in People's Republic of China went critical in November, 1958, and the radioisotope development work began since then. The thermal power of the reactor was 7 MW, and the maximum thermal neutron flux was 1.2 x 10 14 n/cm 2 /sec. Since 1967, it was operated at 10 MW. The first radioisotope product was 24 Na, using Na 2 CO 3 as a target, while the first chemically processed product was an electroplated reference source of 60 Co. The first processed radiochemical was the carrier-free H 2 SO 4 of 35 S. Since then, 131 I and 32 P for medical uses, colloidal 198 Au, colloidal Cr 32 PO 4 , chemicals containing 203 Hg, organic compounds labelled with 125 I, 131 I, 3 H and 14 C and smoke detectors have been produced to date. In addition, 22 Na, 54 Mn, 57 Co, 88 Y, 109 Cd have been prepared from a cyclotron. Now about 140 kinds of products can be supplied, and 60% of the users are the hospitals with nuclear medicine department. The present status of the kinds and production figures of nuclear medicines, radiopharmaceuticals, labelled compounds, radiation sources, and some works in progress are reported. General aspects of the application of radioisotopes in China are also described. Radioisotopes have been applied to agriculture, industry, medicine, and sciences such as physics, chemistry, biology and geography. (Wakatsuki, Y.)

  18. 76 FR 14437 - Economic Simplified Boiling Water Reactor Standard Design: GE Hitachi Nuclear Energy; Issuance of...

    Science.gov (United States)

    2011-03-16

    ... NUCLEAR REGULATORY COMMISSION [NRC-2011-0055] Economic Simplified Boiling Water Reactor Standard Design: GE Hitachi Nuclear Energy; Issuance of Final Design Approval The U.S. Nuclear Regulatory Commission has issued a final design approval (FDA) to GE Hitachi Nuclear Energy (GEH) for the economic...

  19. Standard review plan for the review and evaluation of emergency plans for research and test reactors

    International Nuclear Information System (INIS)

    1983-10-01

    This document provides a Standard Review Plan to assure that complete and uniform reviews are made of research and test reactor radiological emergency plans. The report is organized under ten planning standards which correspond to the guidance criteria in American National Standard ANSI/ANS 15.16 - 1982 as endorsed by Revision 1 to Regulatory Guide 2.6. The applicability of the items under each planning standard is indicated by subdivisions of the steady-state thermal power levels at which the reactors are licensed to operate. Standard emergency classes and example action levels for research and test reactors which should initiate these classes are given in an Appendix. The content of the emergency plan is as follows: the emergency plan addresses the necessary provisions for coping with radiological emergencies. Activation of the emergency plan is in response to the emergency action levels. In addition to addressing those severe emergencies that will fall within one of the standard emergency classes, the plan also discusses the necessary provisions to deal with radiological emergencies of lesser severity that can occur within the operations boundary. The emergency plan allows for emergency personnel to deviate from actions described in the plan for unusual or unanticipated conditions

  20. Safety of reactors built according to earlier standards (WWER 440/V230 type)

    International Nuclear Information System (INIS)

    Misak, J.; Rohar, S.

    1995-01-01

    The problems of safety of WWER-440/V-230 type reactors are discussed, and the following conclusions are made. (1) The reactors have a very good operational record. (2) The reactors have serious design shortcomings, which should be eliminated by safety upgrading. Core damage frequency should be further reduced. (3) PSA methods constitute an appropriate tool for assessment of plant vulnerability to some initiating events and malfunctions, for prioritization of upgrading measures and for tolerability of deviations from current safety standards. (4) The most important safety merits, such as a large thermal inertia and low rupture probability, should be properly taken into account in the analysis. (5) Extensive safety upgrading is feasible and can lead to a considerable risk reduction. In certain circumstances such upgrading is the least expensive option even though the total cost is much higher than the initial plant construction cost. (6) Properly upgraded, the reactor units may be operable until better power resources are available within the country. (7) The existing gap between the technological and political judgements of nuclear safety should be reduced continuously by information exchange improvements. (8) A unified approach to nuclear safety should be adopted for all nuclear reactors (not just WWERs) built to earlier standards. 5 tabs., 1 fig

  1. Effect of atomic spontaneous decay on entanglement in the generalized Jaynes-Cummings model

    International Nuclear Information System (INIS)

    Hessian, H.A.; Obada, A.-S.F.; Mohamed, A.-B.A.

    2010-01-01

    Some aspects of the irreversible dynamics of a generalized Jaynes-Cummings model are addressed. By working in the dressed-state representation, it is possible to split the dynamics of the entanglement and coherence. The exact solution of the master equation in the case of a high-Q cavity with atomic decay is found. Effects of the atomic spontaneous decay on the temporal evolution of partial entropies of the atom or the field and the total entropy as a quantitative measure entanglement are elucidated. The degree of entanglement, through the sum of the negative eigenvalues of the partially transposed density matrix and the negative mutual information has been studied and compared with other measures.

  2. Standard Guide for In-Service Annealing of Light-Water Moderated Nuclear Reactor Vessels

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2003-01-01

    1.1 This guide covers the general procedures to be considered for conducting an in-service thermal anneal of a light-water moderated nuclear reactor vessel and demonstrating the effectiveness of the procedure. The purpose of this in-service annealing (heat treatment) is to improve the mechanical properties, especially fracture toughness, of the reactor vessel materials previously degraded by neutron embrittlement. The improvement in mechanical properties generally is assessed using Charpy V-notch impact test results, or alternatively, fracture toughness test results or inferred toughness property changes from tensile, hardness, indentation, or other miniature specimen testing (1). 1.2 This guide is designed to accommodate the variable response of reactor-vessel materials in post-irradiation annealing at various temperatures and different time periods. Certain inherent limiting factors must be considered in developing an annealing procedure. These factors include system-design limitations; physical constrain...

  3. Report on the specialists' meeting on passive and active safety features of liquid-metal fast breeder reactors organized by the international atomic energy agency at Oarai Engineering Centre of power reactor and nuclear development corporation, Japan, November 5-7, 1991

    International Nuclear Information System (INIS)

    Paranjpe, S.R.

    1992-01-01

    As recommended by the International Working Group on Fast Reactors (IWGFR), the International Atomic Energy Agency organized a specialists' meeting on passive and active safety features of liquid-metal fast breeder reactors. Specialists from all member countries of IWGFR-China, France, Germany, India, Italy, Japan, Russia, the United Kingdom, and the United States-participated in the meeting and made presentations as listed in Table 1. The Commission of European Communities also sent representatives to the meeting. Table 2 contains a list of participants. The meeting consisted of five sessions: (1) an overview, (2) safety characteristics of decay heal removal systems, (3) safely characteristics of reactor protection systems and reactor shutdown systems, (4) safely characteristics of reactor cores, and (5) general discussions antiformulation of recommendations

  4. Nuclear reactor can detonate like an atomic bomb: yes or no

    International Nuclear Information System (INIS)

    Martin, D.

    1974-01-01

    The fission process in a nuclear weapon and in power reactor are compared. The operation of a power reactor is described and the various protective systems and shielding devices are mentioned. It is shown that it is not possible for a nuclear explosion to occur in a power reactor

  5. The International Atomic Energy Agency (IAEA) standards and recommendations on radioactive waste and transport safety

    International Nuclear Information System (INIS)

    Warnecke, E.; Rawl, R.

    1996-01-01

    The International Atomic Energy Agency (IAEA) publishes standards and recommendations on all aspects of nuclear safety in its Safety Series, which includes radioactive waste management and transport of radioactive materials. Safety Series documents may be adopted by a State into its national legal framework. Most of the States used the IAEA transport regulations (Safety Series No. 6) as a basis for their national regulation. The two highest ranking documents of the Radioactive Waste Safety Standards (RADWASS) programme, the Safety Fundamentals and the Safety Standard on the national waste management system, have been published. Both provide impetus into the waste management safety convention, a legally binding document for signatory states, which is being drafted. The already existing Convention on Nuclear Safety covers the management of radioactive waste at land-based civil nuclear power plants. (author) 1 fig., 18 refs

  6. Technical regulations on the general design and safety criteria for design and construction of nuclear reactors of May 1975

    International Nuclear Information System (INIS)

    1975-05-01

    These Technical Regulations published on 5th September 1975 were made in implementation of Section 33 of Decree No 7/9141 on the procedure for the licensing of nuclear installations. They serve as a guide to licensing authorities, project designers and operators in the nuclear field and therefore provide general criteria for safety standards, engineering codes, siting considerations, design bases for overall environmental radiation protection, and also deal with reactor core design, instrumentation, control, alarm systems, including an emergency core cooling system. Finally, the safe design of fuel elements must be ensured and fuel storage and handling techniques complied with. (NEA) [fr

  7. Generic risk insights for General Electric boiling water reactors

    International Nuclear Information System (INIS)

    Travis, R.; Taylor, J.; Chung, J.

    1991-05-01

    A methodology has been developed to extract generic risk-based information from probabilistic risk assessments (PRAs) of General Electric boiling water rectors and applying the insights gained to plants that have not been subjected to a PRA. The available risk assessments (six plants) were examined to identify the most probable, i.e., dominant accident sequences at each plants. The goal was to include all sequences which represented at least 80% of core damage frequency. If the same plant specific dominant accident sequence appeared within this boundary in at least two plant PRAs, the sequence was considered to be a representative sequence. Eight sequences met this definition. From these sequences, the most important component failures and human error that contributed to each sequence have been prioritized. Guidance is provided to prioritize the representative sequences and modify selected basic events that have been shown to be sensitive to the plant specific design or operating variations of the contributing PRAs. This risk-based guidance can be used for utility and NRC activities including operator training, maintenance, design review, and inspections. 13 refs., 6 tabs

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

    International Nuclear Information System (INIS)

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

    1998-02-01

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

  9. The general expression for the transition amplitude of two-photon ionization of atomic hydrogen

    Energy Technology Data Exchange (ETDEWEB)

    Karule, E [Institute of Atomic Physics and Spectroscopy, University of Latvia, Raina Boulevard 19, Riga, LV-1586 (Latvia); Moine, B [Universite Paris Sud, 91405 Orsay Cedex (France)

    2003-05-28

    Two-photon ionization of atomic hydrogen with an excess photon is revisited. The non-relativistic dipole approximation and Coulomb Green function (CGF) formalism are applied. Using the CGF Sturmian expansion straightforwardly, one gets the radial transition amplitude in the form of an infinite sum over Gauss hypergeometric functions which are polynomials. It is convergent if all intermediate states are in the discrete spectrum. In the case of two-photon ionization with an excess photon, when photoionization is also possible, intermediate states are in the continuum. We performed the explicit summation over intermediate states and got a simple general expression for the radial transition amplitude in the form of a finite sum over Appell hypergeometric functions, which are not polynomials. An Appell function may be expressed as an infinite sum over Gauss functions. In the case of ionization by an excess photon, Gauss functions are transformed to give a convergent radial transition amplitude for the whole region. The generalized cross sections for two-photon above-threshold ionization of atomic hydrogen in the ground state and excited states calculated by us agree very well with results of previous calculations. Generalized cross sections for two-photon ionization of positronium in the ground state are obtained by scaling those for atomic hydrogen.

  10. Standardization of specifications and inspection procedures for LEU plate-type research reactor fuels

    International Nuclear Information System (INIS)

    1988-06-01

    With the transition to high density uranium LEU fuel, fabrication costs of research reactor fuel elements have a tendency to increase because of two reasons. First, the amount of the powder of the uranium compound required increases by more than a factor of five. Second, fabrication requirements are in many cases nearer the fabrication limits. Therefore, it is important that measures be undertaken to eliminate or reduce unnecessary requirements in the specification or inspection procedures of research reactor fuel elements utilizing LEU. An additional stimulus for standardizing specifications and inspection procedures at this time is provided by the fact that most LEU conversions will occur within a short time span, and that nearly all of them will require preparation of new specifications and inspection procedures. In this sense, the LEU conversions offer an opportunity for improving the rationality and efficiency of the fuel fabrication and inspection processes. This report focuses on the standardization of specifications and inspection processes of high uranium density LEU fuels for research reactors. However, in many cases the results can also be extended directly to other research reactor fuels. 15 refs, 1 fig., 3 tabs

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

    International Nuclear Information System (INIS)

    Forlerer, Elena; Palacios, Tulio A.

    1998-01-01

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

  12. The IPR-R1 TRIGA Mark I Reactor in 39 years: Operations and general improvements

    International Nuclear Information System (INIS)

    Maretti Junior, Fausto; Prado Fernandes, Marcio; Oliveira, Paulo Fernando; Alves de Amorim, Valter

    1999-01-01

    The nuclear IPR-R1 TRIGA Mark I Reactor operating in the Nuclear Technology Development Center, originally Institute for Radioactive Research in Minas Gerais, Brazil, was dedicated in November 11, 1960. Initially operating for the production of radioisotopes for different uses, it started later to be used in large scale for neutron activation analysis and training of operators for nuclear power plants. Many improvements have been made throughout these years to provide a better performance in its operation and safety conditions. A new cooling system to operate until 300 kW, a new control rod mechanism, an aluminum tank for the reactor pool, an optimization in the pneumatic system, a new reactor control console and a general remodeling of the reactor laboratory were some of the improvements added. To prevent and mitigate the ageing effects, the reactor operation personnel is starting a program to minimize future operation problems. This paper describes the improvements made, the results obtained during the past 39 years, and the precautions taken to ensure future safe operation of the reactor to give operators better conditions of safe work. (author)

  13. Processing of FRG mixed oxide fuel elements at General Atomic under the US/FRG cooperative agreement for spent fuel elements

    International Nuclear Information System (INIS)

    Holder, N.D.; Strand, J.B.; Schwarz, F.A.; Tischer, H.E.

    1980-11-01

    The Federal Republic of Germany (FRG) and the United States (US) are cooperating on certain aspects gas-cooled reactor technology under an umbrella agreement. Under the spent fuel treatment section of the agreement, FRG fuel spheres were recently sent for processing in the Department of Energy sponsored cold pilot plant for High-Temperature Gas-Cooled Reactor (HTGR) fuel processing at General Atomic Company in San Diego, California. The FRG fuel spheres were crushed and burned to recover coated fuel particles. These particles were in turn crushed and burned to recover the fuel-bearing kernels for further treatment for uranium recovery. Successful completion of the tests described in this paper demonstrated the applicability of the US HTGR fuel treatment flowsheet to FRG fuel processing. 10 figures

  14. 10 CFR 110.26 - General license for the export of nuclear reactor components.

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false General license for the export of nuclear reactor components. 110.26 Section 110.26 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) EXPORT AND IMPORT OF... Denmark Finland France Germany Greece Indonesia Ireland Italy Japan Latvia Lithuania Luxembourg...

  15. SACRD: a data base for fast reactor safety computer codes, general description

    International Nuclear Information System (INIS)

    Greene, N.M.; Forsberg, V.M.; Raiford, G.B.; Arwood, J.W.; Simpson, D.B.; Flanagan, G.F.

    1979-01-01

    SACRD is a data base of material properties and other handbook data needed in computer codes used for fast reactor safety studies. Data are available in the thermodynamics, heat transfer, fluid mechanics, structural mechanics, aerosol transport, meteorology, neutronics, and dosimetry areas. Tabular, graphical and parameterized data are provided in many cases. A general description of the SACRD system is presented in the report

  16. Shifts of neutrino oscillation parameters in reactor antineutrino experiments with non-standard interactions

    Directory of Open Access Journals (Sweden)

    Yu-Feng Li

    2014-11-01

    Full Text Available We discuss reactor antineutrino oscillations with non-standard interactions (NSIs at the neutrino production and detection processes. The neutrino oscillation probability is calculated with a parametrization of the NSI parameters by splitting them into the averages and differences of the production and detection processes respectively. The average parts induce constant shifts of the neutrino mixing angles from their true values, and the difference parts can generate the energy (and baseline dependent corrections to the initial mass-squared differences. We stress that only the shifts of mass-squared differences are measurable in reactor antineutrino experiments. Taking Jiangmen Underground Neutrino Observatory (JUNO as an example, we analyze how NSIs influence the standard neutrino measurements and to what extent we can constrain the NSI parameters.

  17. Standard Guide for Use of Melt Wire Temperature Monitors for Reactor Vessel Surveillance, E 706 (IIIE)

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2006-01-01

    1.1 This guide describes the application of melt wire temperature monitors and their use for reactor vessel surveillance of light-water power reactors as called for in Practice E 185. 1.2 The purpose of this guide is to recommend the selection and use of the common melt wire technique where the correspondence between melting temperature and composition of different alloys is used as a passive temperature monitor. Guidelines are provided for the selection and calibration of monitor materials; design, fabrication, and assembly of monitor and container; post-irradiation examinations; interpretation of the results; and estimation of uncertainties. 1.3 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use. (See Note 1.)

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

    International Nuclear Information System (INIS)

    Muranaka, R.G.

    1985-01-01

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

  19. The uses and implications of standards in general practice consultations

    DEFF Research Database (Denmark)

    Lippert, Maria Laura; Reventlow, Susanne; Kousgaard, Marius Brostrøm

    2017-01-01

    Quality standards play an increasingly important role in primary care through their inscription in various technologies for improving professional practice. While ‘hard’ biomedical standards have been the most common and debated, current quality development initiatives increasingly seek to include...... as manifestations of an inherent conflict between principles of patient-centredness and formal biomedical quality standards. However, this study suggests that standards on the ‘softer’ aspects of care may just as well interfere with a clinical approach relying on situated and attentive interactions with patients....

  20. Progress report on the IAEA programme on the standardization of reactor dosimetry measurements

    International Nuclear Information System (INIS)

    Ertek, C.; Cross, B.; Chernyshev, V.

    1979-01-01

    This report briefly summarizes present activities, current status and procedures associated with neutron spectrum unfolding by activation technique within the IAEA programme on standardization of reactor radiation measurements. Experimental efforts and calculations related to unfolding are critically analyzed including the most recent techniques, interlaboratory cooperation, direct influence of recently measured cross-sections on the unfolded neutron flux density spectrum, re-evaluation of some cross-sections, neutron self-shielding factors and scattering effects. (author)

  1. High flux materials testing reactor HFR Petten. Characteristics of facilities and standard irradiation devices

    International Nuclear Information System (INIS)

    Roettger, H.; Hardt, P. von der; Tas, A.; Voorbraak, W.P.

    1981-01-01

    For the materials testing reactor HFR some characteristic information is presented. Besides the nuclear data for the experiment positions short descriptions are given of the most important standard facilities for material irradiation and radionuclide production. One paragraph deals with the experimental set-ups for solid state and nuclear structure investigations. The information in this report refers to a core type, which is operational since March 1977. The numerical data compiled have been up-dated to January 1981

  2. Comparison of the General Electric BWR/6 standard plant design to the IAEA NUSS codes and guides

    International Nuclear Information System (INIS)

    D'Ardenne, W.H.; Sherwood, G.G.

    1985-01-01

    The General Electric BWR/6 Mark III standard plant design meets or exceeds current requirements of published International Atomic Energy Agency (IAEA) Nuclear Safety Standards (NUSS) codes and guides. This conclusion is based on a review of the NUSS codes and guides by General Electric and by the co-ordinated US review of the NUSS codes and guides during their development. General Electric compared the published IAEA NUSS codes and guides with the General Electric design. The applicability of each code and guide to the BWR/6 Mark III standard plant design was determined. Each code or guide was reviewed by a General Electric engineer knowledgeable about the structures, systems and components addressed and the technical area covered by that code or guide. The results of this review show that the BWR/6 Mark III standard plant design meets or exceeds the applicable requirements of the published IAEA NUSS codes and guides. The co-ordinated US review of the IAEA NUSS codes and guides corroborates the General Electric review. In the co-ordinated US review, the USNRC and US industry organizations (including General Electric) review the NUSS codes and guides during their development. This review ensures that the NUSS codes and guides are consistent with the current US government regulations, guidance and regulatory practices, US voluntary industry codes and standards, and accepted US industry design, construction and operational practices. If any inconsistencies are identified, comments are submitted to the IAEA by the USNRC. All US concerns submitted to the IAEA have been resolved. General Electric design reviews and the Final Design Approval (FDA) issued by the USNRC have verified that the General Electric BWR/6 Mark III standard plant design meets or exceeds the current US requirements, guidance and practices. Since these requirements, guidance and practices meet or exceed those of the NUSS codes and guides, so does the General Electric design. (author)

  3. 40 CFR 94.7 - General standards and requirements.

    Science.gov (United States)

    2010-07-01

    ... to the emission standards of this part are equipped with a connection in the engine exhaust system that is located downstream of the engine and before any point at which the exhaust contacts water (or... be internally threaded with standard pipe threads of a size not larger than one-half inch, and shall...

  4. Licensing topical report: interpretation of general design criteria for high-temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Orvis, D.D.; Raabe, P.H.

    1980-01-01

    This Licensing Topical Report presents a set of General Design Criteria (GDC) which is proposed for applicability to licensing of graphite-moderated, high-temperature gas-cooled reactors (HTGRs). Modifications as necessary to reflect HTGR characteristics and design practices have been made to the GDC derived for applicability to light-water-cooled reactors and presented in Appendix A of Part 50, Title 10, Code of Federal Regulations, including the Introduction, Definitions, and Criteria. It is concluded that the proposed set of GDC affords a better basis for design and licensing of HTGRs

  5. General plan for the partial dismantling of the IRT-Sofia research reactor

    Directory of Open Access Journals (Sweden)

    Apostolov Tihomir G.

    2006-01-01

    Full Text Available After the decision of the Bulgarian Government to reconstruct it, the strategy concerning the IRT-Sofia Research Reactor is to partially dismantle the old systems and equipment. The removal of the reactor core and replacement of old equipment will not pose any significant problems. For a more efficient use of existing resources, there is a need for an engineering project which has been already prepared under the title "General Plan for the Partial Dismantling of Equipment at the IRT-Sofia as a Part of the Reconstruction into a Low Power RR".

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-08-15

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

  7. Simulation of the preliminary General Electric SP-100 space reactor concept using the ATHENA computer code

    International Nuclear Information System (INIS)

    Fletcher, C.D.

    1986-01-01

    The capability to perform thermal-hydraulic analyses of a space reactor using the ATHENA computer code is demonstrated. The fast reactor, liquid-lithium coolant loops, and lithium-filled heat pipes of the preliminary General electric SP-100 design were modeled with ATHENA. Two demonstration transient calculations were performed simulating accident conditions. Calculated results are available for display using the Nuclear Plant Analyzer color graphics analysis tool in addition to traditional plots. ATHENA-calculated results appear reasonable, both for steady state full power conditions, and for the two transients. This analysis represents the first known transient thermal-hydraulic simulation using an integral space reactor system model incorporating heat pipes. 6 refs., 17 figs., 1 tab

  8. 29 CFR 1607.5 - General standards for validity studies.

    Science.gov (United States)

    2010-07-01

    ...) (hereinafter “A.P.A. Standards”) and standard textbooks and journals in the field of personnel selection. D... market and the job should be considered in the determination of when a validity study is outdated. ...

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

    Directory of Open Access Journals (Sweden)

    Robert L. Z. Hoye

    2015-04-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-04-01

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

  11. Standard practice for analysis and interpretation of physics dosimetry results for test reactors

    International Nuclear Information System (INIS)

    Anon.

    1984-01-01

    This practice describes the methodology summarized in Annex Al to be used in the analysis and interpretation of physics-dosimetry results from test reactors. This practice relies on, and ties together, the application of several supporting ASTM standard practices, guides, and methods that are in various stages of completion (see Fig. 1). Support subject areas that are discussed include reactor physics calculations, dosimeter selection and analysis, exposure units, and neutron spectrum adjustment methods. This practice is directed towards the development and application of physics-dosimetrymetallurgical data obtained from test reactor irradiation experiments that are performed in support of the operation, licensing, and regulation of LWR nuclear power plants. It specifically addresses the physics-dosimetry aspects of the problem. Procedures related to the analysis, interpretation, and application of both test and power reactor physics-dosimetry-metallurgy results are addressed in Practice E 853, Practice E 560, Matrix E 706(IE), Practice E 185, Matrix E 706(IG), Guide E 900, and Method E 646

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

    International Nuclear Information System (INIS)

    Clark, R.E.H.

    2006-01-01

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

  13. General Atomic Reprocessing Pilot Plant: engineering-scale dissolution system description

    International Nuclear Information System (INIS)

    Yip, H.H.

    1979-04-01

    In February 1978, a dissolver-centrifuge system was added to the cold reprocessing pilot plant at General Atomic Company, which completed the installation of an HTGR fuel head-end reprocessing pilot plant. This report describes the engineering-scale equipment in the pilot plant and summarizes the design features derived from development work performed in the last few years. The dissolver operating cycles for both thorium containing BISO and uranium containinng WAR fissile fuels are included. A continuous vertical centrifuge is used to clarify the resultant dissolver product solution. Process instrumentation and controls for the system reflect design philosophy suitable for remote operation

  14. Semiclassical series solution of the generalized phase shift atom--diatom scattering equations

    International Nuclear Information System (INIS)

    Squire, K.R.; Curtiss, C.F.

    1980-01-01

    A semiclassical series solution of the previously developed operator form of the generalized phase shift equations describing atom--diatom scattering is presented. This development is based on earlier work which led to a double series in powers of Planck's constant and a scaling parameter of the anisotropic portion of the intermolecular potential. The present solution is similar in that it is a double power series in Planck's constant and in the difference between the spherical radial momentum and a first order approximation. The present series solution avoids difficulties of the previous series associated with the classical turning point

  15. The world must build two atomic reactors each day the next hundred years

    International Nuclear Information System (INIS)

    Anon.

    1975-01-01

    In summarising and commenting on the ideas presented in Mesarovic and Pestel's book 'Mankind at the turning point' it is pointed out that the global energy crisis makes comprehensive long term planning a necessity. Assuming, optimistically, that nuclear power alone is able to supply the total projected energy demand in 100 years, it is stated that this will require 3000 nuclear power stations, each with 8 fast breeder reactors, totally 100GW(t). This means a net rate of construction of four reactors per week, which again means, allowing for a 30 year life, two reactors per day, every day, for the next hundred years. Fuelling these reactors will require the production and transport of 15 x 10 6 kg of Pu239 per year. It is therefore obvious that the energy crisis is not only a technological, but also a political, social and even psychological problem. (JIW)

  16. Quantum electrodynamics tests and X-rays standards using pionic atoms and highly charged ions

    International Nuclear Information System (INIS)

    Martino, Trassinelli

    2005-12-01

    The object of this thesis is to present a new measurement of the pion mass using pionic nitrogen X-ray spectroscopy and results on helium-like argon and sulphur spectroscopy. The new pion mass has been measured with an accuracy of 1.7 ppm, 30% better that the present world average value, and it is obtained from Bragg spectroscopy of 5 ->4 pionic nitrogen transitions using the theoretical predictions provided by quantum electrodynamics. We have got: m(π - ) = (139.571042 ± 0.000210 ± 0.000110) where the first error is due to the statistics and the second is the systematic error. I present the calculation of the hyperfine structure and recoil corrections for pionic atoms using a new perturbation method for the Klein-Gordon equation. The spectrometer used for this measurement has been characterized with the relativistic M1 transitions from helium-like ions produced with a new device, the Electron-Cyclotron-Resonance Ion Trap. High statistics spectra from these ions have enabled us to measure transition energies with an accuracy of some ppm which has allowed us to compare theoretical predictions with experiment data. X-ray emission from pionic atoms and multicharged ions can be used to define new types of X-ray standards for energies of a few keV

  17. Activity of the Atomic Energy Society of Japan for compiling the consensus standard on nuclear criticality safety control

    International Nuclear Information System (INIS)

    Yamane, Yoshihiro; Matsumoto, Tadakuni

    2003-01-01

    Activity of the Atomic Energy Society of Japan for compiling the consensus standard on nuclear criticality safety control is presented. The standard recommends an enhancement of nuclear criticality safety throughout a life cycle of facility in terms of a concept of 'barriers against criticality'. (author)

  18. Standard Guide for Application of Neutron Transport Methods for Reactor Vessel Surveillance, E706 (IID)

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2011-01-01

    1.1 Need for Neutronics Calculations—An accurate calculation of the neutron fluence and fluence rate at several locations is essential for the analysis of integral dosimetry measurements and for predicting irradiation damage exposure parameter values in the pressure vessel. Exposure parameter values may be obtained directly from calculations or indirectly from calculations that are adjusted with dosimetry measurements; Guide E944 and Practice E853 define appropriate computational procedures. 1.2 Methodology—Neutronics calculations for application to reactor vessel surveillance encompass three essential areas: (1) validation of methods by comparison of calculations with dosimetry measurements in a benchmark experiment, (2) determination of the neutron source distribution in the reactor core, and (3) calculation of neutron fluence rate at the surveillance position and in the pressure vessel. 1.3 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is th...

  19. Development of standards and investigation of safety examination items for advancement of safety regulation of fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    The purposes of this study are to prepare the fuel technical standard and the structure and materials standard of fast breeder reactors (FBRs), and to develop the requirements in a reactor establishment permission. The objects of this study are mainly the Monju high performance core and a demonstration FBR. In JFY 2012, the following results were obtained. As for the fuel technical standard, the fuel technical standard adapting the examination of integrity of the FBR fuels was prepared based on the information and data obtained in this study. As for the structure and material standard, the investigation of the revised parts of the standard was carried out. And as for the examination of the safety requirements, safety evaluation items for the future FBR plant and the fission products to be considered in a reactor establishment permission were investigated and examined. (author)

  20. Thai research reactor

    International Nuclear Information System (INIS)

    Aramrattana, M.

    1987-01-01

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

  1. Why has the bohr-sommerfeld model of the atom been ignoredby general chemistry textbooks?

    Science.gov (United States)

    Niaz, Mansoor; Cardellini, Liberato

    2011-12-01

    Bohr's model of the atom is considered to be important by general chemistry textbooks. A major shortcoming of this model was that it could not explain the spectra of atoms containing more than one electron. In order to increase the explanatory power of the model, Sommerfeld hypothesized the existence of elliptical orbits. This study has the following objectives: 1) Formulation of criteria based on a history and philosophy of science framework; and 2) Evaluation of university-level general chemistry textbooks based on the criteria, published in Italy and U.S.A. Presentation of a textbook was considered to be "satisfactory" if it included a description of the Bohr-Sommerfeld model along with diagrams of the elliptical orbits. Of the 28 textbooks published in Italy that were analyzed, only five were classified as "satisfactory". Of the 46 textbooks published in U.S.A., only three were classified as "satisfactory". This study has the following educational implications: a) Sommerfeld's innovation (auxiliary hypothesis) by introducing elliptical orbits, helped to restore the viability of Bohr's model; b) Bohr-Sommerfeld's model went no further than the alkali metals, which led scientists to look for other models; c) This clearly shows that scientific models are tentative in nature; d) Textbook authors and chemistry teachers do not consider the tentative nature of scientific knowledge to be important; e) Inclusion of the Bohr-Sommerfeld model in textbooks can help our students to understand how science progresses.

  2. Regulations for RA reactor operation

    International Nuclear Information System (INIS)

    1980-09-01

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

  3. Nuclear reactor pressure vessel surveillance capsule examinations. Application of American Society for Testing and Materials Standards

    International Nuclear Information System (INIS)

    Perrin, J.S.

    1978-01-01

    A series of pressure vessel surveillance capsules is installed in each commercial nuclear power plant in the United States. A capsule typically contains neutron dose meters, thermal monitors, tensile specimens, and Charpy V-notch impact specimens. In order to determine property changes of the pressure vessel resulting from irradiation, surveillance capsules are periodically removed during the life of a reactor and examined. There are numerous standards, regulations, and codes governing US pressure vessel surveillance capsule programmes. These are put out by the US Nuclear Regulatory Commission, the Boiler and Pressure Vessel Committee of the American Society of Mechanical Engineers, and the American Society for Testing and Materials (ASTM). A majority of the pertinent ASTM standards are under the jurisdiction of ASTM Committee E-10 on Nuclear Applications and Measurements of Radiation Effects. The standards, regulations, and codes pertaining to pressure vessel surveillance play an important role in ensuring reliability of the nuclear pressure vessels. ASTM E 185-73 is the Standard Recommended Practice for Surveillance Tests for Nuclear Reactors. This standard recommends procedures for both the irradiation and subsequent testing of surveillance capsules. ASTM E 185-73 references many additional specialized ASTM standards to be followed in specific areas of a surveillance capsule examination. A key element of surveillance capsule programmes is the Charpy V-notch impact test, used to define curves of fracture behaviour over a range of temperatures. The data from these tests are used to define the adjusted reference temperature used in determining pressure-temperature operating curves for a nuclear power plant. (author)

  4. Guide to power reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1959-07-15

    The IAEA's major first scientific publication is the Directory of Power Reactors now in operation or under construction in various parts of the world. The purpose of the directory is to present important details of various power projects in such a way as to provide a source of easy reference for anyone interested in the development of the peaceful uses of atomic energy, either at the technical or management level. Six pages have been devoted to each reactor the first of which contains general information, reactor physics data and information about the core. The second and third contain sketches of the fuel element or of the fuel element assembly, and of the horizontal and vertical sections of the reactor. On the fourth page information is grouped under the following heads: fuel element, core heat transfer, control, reactor vessel and over-all dimensions, and fluid flow. The fifth page shows a simplified flow diagram, while the sixth provides information on reflector and shielding, containment and turbo generator. Some information has also been given, when available, on cost estimates and operating staff requirements. Remarks and a bibliography constitute the last part of the description of each reactor. Reactor projects included in this directory are pressurized light water cooled power reactors. Boiling light water cooled power reactors, heavy water cooled power reactors, gas cooled power reactors, organic cooled power reactors liquid metal cooled power reactors and liquid metal cooled power reactors

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

    International Nuclear Information System (INIS)

    Virgilio, M.J.

    1980-10-01

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

  6. Flow injection analysis-flame atomic absorption spectrometry system for indirect determination of sulfite after on-line reduction of solid-phase manganese (IV) dioxide reactor.

    Science.gov (United States)

    Zare-Dorabei, Rouholah; Boroun, Shokoufeh; Noroozifar, Meissam

    2018-02-01

    A new and simple flow injection method followed by atomic absorption spectrometry was developed for indirect determination of sulfite. The proposed method is based on the oxidation of sulfite to sulphate ion using solid-phase manganese dioxide (30% W/W suspended on silica gel beads) reactor. MnO 2 will be reduced to Mn(II) by sample injection in to the column under acidic carrier stream of HNO 3 (pH 2) with flow rate of 3.5mLmin -1 at room temperature. Absorption measurement of Mn(II) which is proportional to the concentration of sulfite in the sample was carried out by atomic absorption spectrometry. The calibration curve was linear up to 25mgL -1 with a detection limit (DL) of 0.08mgL -1 for 400µL injection sample volume. The presented method is efficient toward sulfite determination in sugar and water samples with a relative standard deviation (RSD) less than 1.2% and a sampling rate of about 60h -1 . Copyright © 2017 Elsevier B.V. All rights reserved.

  7. Column preconcentration and electrothermal atomic absorption spectrometric determination of rhodium in some food and standard samples.

    Science.gov (United States)

    Taher, Mohammad Ali; Pourmohammad, Fatemeh; Fazelirad, Hamid

    2015-12-01

    In the present work, an electrothermal atomic absorption spectrometric method has been developed for the determination of ultra-trace amounts of rhodium after adsorption of its 2-(5-bromo-2-pyridylazo)-5-diethylaminophenol/tetraphenylborate ion associated complex at the surface of alumina. Several factors affecting the extraction efficiency such as the pH, type of eluent, sample and eluent flow rates, sorption capacity of alumina and sample volume were investigated and optimized. The relative standard deviation for eight measurements of 0.1 ng/mL of rhodium was ±6.3%. In this method, the detection limit was 0.003 ng/mL in the original solution. The sorption capacity of alumina and the linear range for Rh(III) were evaluated as 0.8 mg/g and 0.015-0.45 ng/mL in the original solution, respectively. The proposed method was successfully applied for the extraction and determination of rhodium content in some food and standard samples with high recovery values. © 2015 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  8. Adjustment of the Brazilian radioprotection standards to the safety principles of the International Atomic Energy Agency

    International Nuclear Information System (INIS)

    Pereira, Wagner de S.; Py Junior, Delcy de A.

    2013-01-01

    The International Atomic Energy Agency (IAEA) has a recommendation with 10 basic safety principles (Fundamental Safety Principles Safety Fundamentals series, number SF-1), which are: 1) Responsibility for safety; 2) Role for government; 3) Leadership and management for safety; 4) Justification of facilities and activities; 5) Optimization of protection; 6) Limitation of risk to individuals; 7) Protection of present and futures generations; 8) Prevention of accidents; 9) Emergency preparedness and response and 10) Protection actions to reduce existing or unregulated radiations risk. The aim of this study is to verify that the Brazilian standards of radiation protection meet the principles described above and how well suited to them. The analysis of the national radiation protection regulatory system, developed and deployed by the National Nuclear Energy Commission (CNEN), showed that out of the ten items, two are covered partially, the number 2 and 10. The others are fully met. The item 2 the fact that the regulatory body (CNEN) be stock controller of a large company in the sector put in check its independence as a regulatory body. In item 10 the Brazilian standard of radiation protection does not provide explicit resolution of environmental liabilities

  9. Design and performance of General Electric boiling water reactor main steam line isolation valves

    International Nuclear Information System (INIS)

    Rockwell, D.A.; van Zylstra, E.H.

    1976-08-01

    An extensive test program has been completed by the General Electric Company in cooperation with the Commonwealth Edison Company on the basic design type of large main steam line isolation valves used on General Electric Boiling Water Reactors. Based on a total of 40 tests under simulated accident conditions covering a wide range of mass flows, mixture qualities, and closing times, it was concluded that the commercially available valves of this basic type will close completely and reliably as required. Analytical methods to predict transient effects in the steam line and valve after postulated breaks were refined and confirmed by the test program

  10. A Standardized Generalized Dimensionality Discrepancy Measure and a Standardized Model-Based Covariance for Dimensionality Assessment for Multidimensional Models

    Science.gov (United States)

    Levy, Roy; Xu, Yuning; Yel, Nedim; Svetina, Dubravka

    2015-01-01

    The standardized generalized dimensionality discrepancy measure and the standardized model-based covariance are introduced as tools to critique dimensionality assumptions in multidimensional item response models. These tools are grounded in a covariance theory perspective and associated connections between dimensionality and local independence.…

  11. Activated carbon-modified knotted reactor coupled to electrothermal atomic absorption spectrometry for sensitive determination of arsenic species in medicinal herbs and tea infusions

    Energy Technology Data Exchange (ETDEWEB)

    Grijalba, Alexander Castro; Martinis, Estefanía M. [Laboratory of Analytical Chemistry for Research and Development (QUIANID), Facultad de Ciencias Exactas y Naturales, Universidad Nacional de Cuyo, Padre J. Contreras 1300, (5500) Mendoza (Argentina); Consejo Nacional de Investigaciones Científicas y Técnicas (CONICET), Buenos Aires (Argentina); Lascalea, Gustavo E. [Consejo Nacional de Investigaciones Científicas y Técnicas (CONICET), Buenos Aires (Argentina); Wuilloud, Rodolfo G., E-mail: rwuilloud@mendoza-conicet.gob.ar [Laboratory of Analytical Chemistry for Research and Development (QUIANID), Facultad de Ciencias Exactas y Naturales, Universidad Nacional de Cuyo, Padre J. Contreras 1300, (5500) Mendoza (Argentina); Consejo Nacional de Investigaciones Científicas y Técnicas (CONICET), Buenos Aires (Argentina)

    2015-01-01

    A flow injection system based on a modified polytetrafluoroethylene (PTFE) knotted reactor (KR) was developed for arsenite [As(III)] and arsenate [As(V)] species preconcentration and determination by electrothermal atomic absorption spectrometry (ETAAS). Activated carbon (AC) was immobilized on the inner walls of a PTFE KR by a thermal treatment. A significant increase in analyte retention was obtained with the AC-modified KR (100%) as compared to the regular PTFE KR (25%). The preconcentration method involved the on-line formation of As(III)-ammonium pyrrolidinedithiocarbamate (As-APDC) complex, followed by its adsorption onto the inner walls of the AC-modified KR. After analyte retention, the complex was eluted with acetone directly into the graphite furnace of ETAAS. The parameters affecting the flow injection system were evaluated with a full central composite face centered design with three center points. Under optimum conditions, a preconcentration factor of 200 was obtained with 10 ml of sample. The detection limit was 4 ng L{sup −1} and the relative standard deviation (RSD) for six replicate measurements at 0.2 μg L{sup −1} of As were 4.3% and 4.7% for As(III) and As(V), respectively. The developed methodology was highly selective towards As(III), while As(V), monomethylarsonic acid [MMA(V)] and dimethylarsinic [DMA(V)] were not retained in the AC-modified KR. The proposed method was successfully applied for As speciation analysis in infusions originated from medicinal herbs and tea. - Highlights: • We report an efficient method for As speciation. • We have modified a knotted reactor with activated carbon for high sorption capacity. • We provide a simple procedure for surface modification of a PTFE knotted reactor. • We have selectively separated inorganic As species from complex matrix samples. • We have implemented a modified KR in a flow injection system coupled to ETAAS.

  12. Activated carbon-modified knotted reactor coupled to electrothermal atomic absorption spectrometry for sensitive determination of arsenic species in medicinal herbs and tea infusions

    International Nuclear Information System (INIS)

    Grijalba, Alexander Castro; Martinis, Estefanía M.; Lascalea, Gustavo E.; Wuilloud, Rodolfo G.

    2015-01-01

    A flow injection system based on a modified polytetrafluoroethylene (PTFE) knotted reactor (KR) was developed for arsenite [As(III)] and arsenate [As(V)] species preconcentration and determination by electrothermal atomic absorption spectrometry (ETAAS). Activated carbon (AC) was immobilized on the inner walls of a PTFE KR by a thermal treatment. A significant increase in analyte retention was obtained with the AC-modified KR (100%) as compared to the regular PTFE KR (25%). The preconcentration method involved the on-line formation of As(III)-ammonium pyrrolidinedithiocarbamate (As-APDC) complex, followed by its adsorption onto the inner walls of the AC-modified KR. After analyte retention, the complex was eluted with acetone directly into the graphite furnace of ETAAS. The parameters affecting the flow injection system were evaluated with a full central composite face centered design with three center points. Under optimum conditions, a preconcentration factor of 200 was obtained with 10 ml of sample. The detection limit was 4 ng L −1 and the relative standard deviation (RSD) for six replicate measurements at 0.2 μg L −1 of As were 4.3% and 4.7% for As(III) and As(V), respectively. The developed methodology was highly selective towards As(III), while As(V), monomethylarsonic acid [MMA(V)] and dimethylarsinic [DMA(V)] were not retained in the AC-modified KR. The proposed method was successfully applied for As speciation analysis in infusions originated from medicinal herbs and tea. - Highlights: • We report an efficient method for As speciation. • We have modified a knotted reactor with activated carbon for high sorption capacity. • We provide a simple procedure for surface modification of a PTFE knotted reactor. • We have selectively separated inorganic As species from complex matrix samples. • We have implemented a modified KR in a flow injection system coupled to ETAAS

  13. Standard technical specifications General Electric plants, BWR/6. Volume 1, Revision 1

    International Nuclear Information System (INIS)

    1995-04-01

    This report documents the results of the combined effort of the NRC and the industry to produce improved Standard Technical Specifications (STS), Revision 1 for General Electric BWR/6 Plants. The changes reflected in Revision 1 resulted from the experience gained from license amendment applications to convert to these improved STS or to adopt partial improvements to existing technical specifications. This NUREG is the result of extensive public technical meetings and discussions between the Nuclear Regulatory Commission (NRC) staff and various nuclear power plant licensees, Nuclear Steam Supply System (NSSS) Owners Groups, NSSS vendors, and the Nuclear Energy Institute (NEI). The improved STS were developed based on the criteria in the Final Commission Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, dated July 22, 1993. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume 1 contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. Volume 3 contains the Bases for Sections 3.4--3.10 of the improved STS

  14. Standard technical specifications: General Electric plants, BWR/4. Volume 1, Revision 1: Specifications

    International Nuclear Information System (INIS)

    1995-04-01

    This report documents the results of the combined effort of the NRC and the industry to produce improved Standard Technical Specifications (STS), Revision 1 for General Electric BWR/4 Plants. The changes reflected in Revision 1 resulted from the experience gained from license amendment applications to convert to these improved STS or to adopt partial improvements to existing technical specifications. This NUREG is the result of extensive public technical meetings and discussions between the Nuclear Regulatory Commission (NRC) staff and various nuclear power plant licensees, Nuclear Steam Supply System (NSSS) Owners Groups, NSSS vendors, and the Nuclear Energy Institute (NEI). The improved STS were developed based on the criteria in the Final Commission Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, dated July 22, 1993. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume 1 contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. Volume 3 contains the Bases for Sections 3.4--3.10 of the improved STS

  15. LASIP-III, a generalized processor for standard interface files

    International Nuclear Information System (INIS)

    Bosler, G.E.; O'Dell, R.D.; Resnik, W.M.

    1976-03-01

    The LASIP-III code was developed for processing Version III standard interface data files which have been specified by the Committee on Computer Code Coordination. This processor performs two distinct tasks, namely, transforming free-field format, BCD data into well-defined binary files and providing for printing and punching data in the binary files. While LASIP-III is exported as a complete free-standing code package, techniques are described for easily separating the processor into two modules, viz., one for creating the binary files and one for printing the files. The two modules can be separated into free-standing codes or they can be incorporated into other codes. Also, the LASIP-III code can be easily expanded for processing additional files, and procedures are described for such an expansion. 2 figures, 8 tables

  16. A probabilistic safety assessment of the standard French 900MWe pressurized water reactor. Main report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1990-04-15

    To situate the probabilistic safety assessment of standardized 900 MWe units made by the Institute for Nuclear Safety and Protection (IPSN), it is necessary to consider the importance and possible utilization of a study of this type. At the present time, the safety of nuclear installations essentially depends on the application of the defence in-depth approach. The design arrangements adopted are justified by the operating organization on the basis of deterministic studies of a limited number of conventional situations with corresponding safety margins. These conventional situations are grouped in categories by frequency, it being accepted that the greater the consequences the lesser the frequency must be. However in the framework of the analysis performed under the control of the French safety authority, the importance was rapidly recognized of setting an overall reference objective. By 1977, on the occasion of appraisal of the fundamental safety options of the standardized 1300 MWe units, the Central Service for the Safety of Nuclear Installations (SCSIN) set the following global probabilistic objective: 'Generally speaking, the design of installations including a pressurized water nuclear reactor must be such that the global probability of the nuclear unit being the origin of unacceptable consequences does not exceed 10{sup -6} per year...' Probabilistic analyses making reference to this global objective gradually began to supplement the deterministic approach, both for examining external hazards to be considered in the design basis and for examining the possible need for additional means of countering the failure of doubled systems in application of the deterministic single-failure criterion. A new step has been taken in France by carrying out two level 1 probabilistic safety assessments (calculation of the annual probability of core meltdown), one for the 900 MWe series by the IPSN and the other for the 1300 MWe series by Electricite de France. The objective

  17. A probabilistic safety assessment of the standard French 900MWe pressurized water reactor. Main report

    International Nuclear Information System (INIS)

    1990-04-01

    To situate the probabilistic safety assessment of standardized 900 MWe units made by the Institute for Nuclear Safety and Protection (IPSN), it is necessary to consider the importance and possible utilization of a study of this type. At the present time, the safety of nuclear installations essentially depends on the application of the defence in-depth approach. The design arrangements adopted are justified by the operating organization on the basis of deterministic studies of a limited number of conventional situations with corresponding safety margins. These conventional situations are grouped in categories by frequency, it being accepted that the greater the consequences the lesser the frequency must be. However in the framework of the analysis performed under the control of the French safety authority, the importance was rapidly recognized of setting an overall reference objective. By 1977, on the occasion of appraisal of the fundamental safety options of the standardized 1300 MWe units, the Central Service for the Safety of Nuclear Installations (SCSIN) set the following global probabilistic objective: 'Generally speaking, the design of installations including a pressurized water nuclear reactor must be such that the global probability of the nuclear unit being the origin of unacceptable consequences does not exceed 10 -6 per year...' Probabilistic analyses making reference to this global objective gradually began to supplement the deterministic approach, both for examining external hazards to be considered in the design basis and for examining the possible need for additional means of countering the failure of doubled systems in application of the deterministic single-failure criterion. A new step has been taken in France by carrying out two level 1 probabilistic safety assessments (calculation of the annual probability of core meltdown), one for the 900 MWe series by the IPSN and the other for the 1300 MWe series by Electricite de France. The objective of

  18. Numerical modeling of carrier gas flow in atomic layer deposition vacuum reactor: A comparative study of lattice Boltzmann models

    International Nuclear Information System (INIS)

    Pan, Dongqing; Chien Jen, Tien; Li, Tao; Yuan, Chris

    2014-01-01

    This paper characterizes the carrier gas flow in the atomic layer deposition (ALD) vacuum reactor by introducing Lattice Boltzmann Method (LBM) to the ALD simulation through a comparative study of two LBM models. Numerical models of gas flow are constructed and implemented in two-dimensional geometry based on lattice Bhatnagar–Gross–Krook (LBGK)-D2Q9 model and two-relaxation-time (TRT) model. Both incompressible and compressible scenarios are simulated and the two models are compared in the aspects of flow features, stability, and efficiency. Our simulation outcome reveals that, for our specific ALD vacuum reactor, TRT model generates better steady laminar flow features all over the domain with better stability and reliability than LBGK-D2Q9 model especially when considering the compressible effects of the gas flow. The LBM-TRT is verified indirectly by comparing the numerical result with conventional continuum-based computational fluid dynamics solvers, and it shows very good agreement with these conventional methods. The velocity field of carrier gas flow through ALD vacuum reactor was characterized by LBM-TRT model finally. The flow in ALD is in a laminar steady state with velocity concentrated at the corners and around the wafer. The effects of flow fields on precursor distributions, surface absorptions, and surface reactions are discussed in detail. Steady and evenly distributed velocity field contribute to higher precursor concentration near the wafer and relatively lower particle velocities help to achieve better surface adsorption and deposition. The ALD reactor geometry needs to be considered carefully if a steady and laminar flow field around the wafer and better surface deposition are desired

  19. Numerical modeling of carrier gas flow in atomic layer deposition vacuum reactor: A comparative study of lattice Boltzmann models

    Energy Technology Data Exchange (ETDEWEB)

    Pan, Dongqing; Chien Jen, Tien [Department of Mechanical Engineering, University of Wisconsin-Milwaukee, Milwaukee, Wisconsin 53201 (United States); Li, Tao [School of Mechanical Engineering, Dalian University of Technology, Dalian 116024 (China); Yuan, Chris, E-mail: cyuan@uwm.edu [Department of Mechanical Engineering, University of Wisconsin-Milwaukee, 3200 North Cramer Street, Milwaukee, Wisconsin 53211 (United States)

    2014-01-15

    This paper characterizes the carrier gas flow in the atomic layer deposition (ALD) vacuum reactor by introducing Lattice Boltzmann Method (LBM) to the ALD simulation through a comparative study of two LBM models. Numerical models of gas flow are constructed and implemented in two-dimensional geometry based on lattice Bhatnagar–Gross–Krook (LBGK)-D2Q9 model and two-relaxation-time (TRT) model. Both incompressible and compressible scenarios are simulated and the two models are compared in the aspects of flow features, stability, and efficiency. Our simulation outcome reveals that, for our specific ALD vacuum reactor, TRT model generates better steady laminar flow features all over the domain with better stability and reliability than LBGK-D2Q9 model especially when considering the compressible effects of the gas flow. The LBM-TRT is verified indirectly by comparing the numerical result with conventional continuum-based computational fluid dynamics solvers, and it shows very good agreement with these conventional methods. The velocity field of carrier gas flow through ALD vacuum reactor was characterized by LBM-TRT model finally. The flow in ALD is in a laminar steady state with velocity concentrated at the corners and around the wafer. The effects of flow fields on precursor distributions, surface absorptions, and surface reactions are discussed in detail. Steady and evenly distributed velocity field contribute to higher precursor concentration near the wafer and relatively lower particle velocities help to achieve better surface adsorption and deposition. The ALD reactor geometry needs to be considered carefully if a steady and laminar flow field around the wafer and better surface deposition are desired.

  20. Bonding analysis of planar hypercoordinate atoms via the generalized BLW-LOL.

    Science.gov (United States)

    Bomble, Laetitia; Steinmann, Stephan N; Perez-Peralta, Nancy; Merino, Gabriel; Corminboeuf, Clemence

    2013-10-05

    The multicenter bonding pattern of the intriguing hexa-, hepta-, and octacoordinate boron wheel series (e.g., CB62-, CB7-, B82-, and SiB8 as well as the experimentally detected CB7- isomer) is revised using the block-localized wave function analyzed by the localized orbital locator (BLW-LOL). The more general implementation of BLW combined with the LOL scalar field is not restricted to the analysis of the out-of-plane π-system but can also provide an intuitive picture of the σ-radial delocalization and of the role of the central atom. The results confirm the presence of a π-ring current pattern similar to that of benzene. In addition, the LOLπ isosurfaces along with the maximum intensity in the ΔLOL profiles located above and below the ring suggest that the central atom plays a minor role in the π-delocalized bonding pattern. Finally, the analysis of the σ-framework in these boron wheels is in line with a moderated inner cyclic rather than disk-type delocalization. Copyright © 2013 Wiley Periodicals, Inc.

  1. A General Catalytic Method for Highly Cost- and Atom-Efficient Nucleophilic Substitutions.

    Science.gov (United States)

    Huy, Peter H; Filbrich, Isabel

    2018-05-23

    A general formamide-catalyzed protocol for the efficient transformation of alcohols into alkyl chlorides, which is promoted by substoichiometric amounts (down to 34 mol %) of inexpensive trichlorotriazine (TCT), is introduced. This is the first example of a TCT-mediated dihydroxychlorination of an OH-containing substrate (e.g., alcohols and carboxylic acids) in which all three chlorine atoms of TCT are transferred to the starting material. The consequently enhanced atom economy facilitates a significantly improved waste balance (E-factors down to 4), cost efficiency, and scalability (>50 g). Furthermore, the current procedure is distinguished by high levels of functional-group compatibility and stereoselectivity, as only weakly acidic cyanuric acid is released as exclusive byproduct. Finally, a one-pot protocol for the preparation of amines, azides, ethers, and sulfides enabled the synthesis of the drug rivastigmine with twofold S N 2 inversion, which demonstrates the high practical value of the presented method. © 2018 Wiley-VCH Verlag GmbH & Co. KGaA, Weinheim.

  2. General Practice Messaging Standard Version 3.0, May 2014

    LENUS (Irish Health Repository)

    2018-01-25

    We addressed the research question "what factors enable the successful development and implementation of a frail older person\\'s pathway within the acute setting". A rapid realist review (RRR) was conducted by adopting the RAMESES standards. We began with a sample of 232 articles via database searches supplemented with 94 additional records including inputs from a twitter chat and a hospital site visit. Our final sample consisted of 18 documents. Following review and consensus by an expert panel we identified a conceptual model of context-mechanism-(resources)-outcomes. There was overall agreement frailty should be identified at the front door of the acute hospital. Significant challenges identified related to organisational boundaries both within the acute setting and externally, the need to shift outcomes to patient orientated ones, to support staff to sustain the pathway by providing ongoing education and by providing role clarity. RRRs can support research such as the systematic approach to improving care for frail older adults (SAFE) study by producing accounts of what works based on a wide range of sources and innovative engagement with stakeholders. It is evident from our provisional model that numerous factors need to combine and interact to enable and sustain a successful frail older person\\'s pathway.

  3. Development of a Superconducting Magnet System for the ONR/General Atomics Homopolar Motor

    Science.gov (United States)

    Schaubel, K. M.; Langhorn, A. R.; Creedon, W. P.; Johanson, N. W.; Sheynin, S.; Thome, R. J.

    2006-04-01

    This paper describes the design, testing and operational experience of a superconducting magnet system presently in use on the Homopolar Motor Program. The homopolar motor is presently being tested at General Atomics in San Diego, California for the U.S Navy Office of Naval Research. The magnet system consists of two identical superconducting solenoid coils housed in two cryostats mounted integrally within the homopolar motor housing. The coils provide the static magnetic field required for motor operation and are wound using NbTi superconductor in a copper matrix. Each magnet is conduction cooled using a Gifford McMahon cryocooler. The coils are in close proximity to the iron motor housing requiring a cold to warm support structure with high stiffness and strength. The design of the coils, cold to warm support structure, cryogenic system, and the overall magnet system design will be described. The test results and operational experience will also be described.

  4. Atomization of U3Si2/U3Si for research reactor fuel

    International Nuclear Information System (INIS)

    Kuk, Il Hiun

    2004-01-01

    Instead of comminuting, U 3 Si 2 /U 3 Si powders are produced by atomizating directly from the molten alloys. Many benefits are introduced by applying the atomization technique: reduction of the process, homogeneous alloy composition within a particle and between particles, increase of the thermal conductivity and decrease of the chemical reactivity with aluminium due to particle's spherical shape. (author)

  5. Reactor

    International Nuclear Information System (INIS)

    Toyama, Masahiro; Kasai, Shigeo.

    1978-01-01

    Purpose: To provide a lmfbr type reactor wherein effusion of coolants through a loop contact portion is reduced even when fuel assemblies float up, and misloading of reactor core constituting elements is prevented thereby improving the reactor safety. Constitution: The reactor core constituents are secured in the reactor by utilizing the differential pressure between the high-pressure cooling chamber and low-pressure cooling chamber. A resistance port is formed at the upper part of a connecting pipe, and which is connect the low-pressure cooling chamber and the lower surface of the reactor core constituent. This resistance part is formed such that the internal sectional area of the connecting pipe is made larger stepwise toward the upper part, and the cylinder is formed larger so that it profiles the inner surface of the connecting pipe. (Aizawa, K.)

  6. Generalized quantum mean-field systems and their application to ultracold atoms

    International Nuclear Information System (INIS)

    Trimborn-Witthaut, Friederike Annemarie

    2011-01-01

    Strongly interacting many-body systems consisting of a large number of indistinguishable particles play an important role in many areas of physics. Though such systems are hard to deal with theoretically since the dimension of the respective Hilbert space increases exponentially both in the particle number and in the number of system modes. Therefore, approximations are of considerable interest. The mean-field approximation describes the behaviour in the macroscopic limit N→∞, which leads to an effective nonlinear single-particle problem. Although this approximation is widely used, rigorous results on the applicability and especially on finite size corrections are extremely rare. One prominent example of strongly interacting many-body systems are ultracold atoms in optical lattices, which are a major subject of this thesis. Typically these systems consist of a large but well-defined number of particles, such that corrections to the mean-field limit can be systematically studied. This thesis is divided into two parts: In the first part we study generalized quantum mean-field systems in a C * -algebraic framework. These systems are characterized by their intrinsic permutation symmetry. In the limit of infinite system size, N→∞, the intensive observables converge to the commutative algebra of weak * -continuous functions on the single particle state space. To quantify the deviations from the meanfield prediction for large but finite N, we establish a differential calculus for state space functions and provide a generalized Taylor expansion around the mean-field limit. Furthermore, we introduce the algebra of macroscopic fluctuations around the mean-field limit and prove a quantum version of the central limit theorem. On the basis of these results, we give a detailed study of the finite size corrections to the ground state energy and establish bounds, for both the quantum and the classical case. Finally, we restrict ourselves to the subspace of Bose

  7. 76 FR 23630 - Office of New Reactors; Proposed Revision 2 to Standard Review Plan, Section 1.0 on Introduction...

    Science.gov (United States)

    2011-04-27

    ... Standard Review Plan, Section 1.0 on Introduction and Interfaces AGENCY: Nuclear Regulatory Commission (NRC... Revision 2 to Standard Review Plan (SRP), Section 1.0, ``Introduction and Interfaces'' (Agencywide Documents Access and Management System (ADAMS) Accession No. ML110110573). The Office of New Reactors (NRO...

  8. Standard review plan for the review and evaluation of emergency plans for research and test reactors. Technical report

    International Nuclear Information System (INIS)

    Bates, E.F.; Grimes, B.K.; Ramos, S.L.

    1982-05-01

    This document provides a Standard Review Plan for the guidance of the NRC staff to assure that complete and uniform reviews are made of research and test reactor emergency plans. The report is organized under ten planning standards which correspond to the guidance criteria in Draft II of ANSI/ANS 15.16 as endorsed by Revision 1 to Regulatory Guide 2.6. The applicability of the items under each planning standard is indicated by subdivisions of the steady state thermal power levels at which the reactors are licensed to operate. Standard emergency classes and example action levels for research and test reactors which should initiate these classes are given in an Appendix

  9. Very high temperature gas-cooled reactor critical facility for Japan Atomic Energy Research Institute

    International Nuclear Information System (INIS)

    Ishihara, Noriyuki

    1985-01-01

    The outline of the critical facility, its construction, the results of the basic studies and experiments on the graphite material, and the results obtained from the test conducted on the overall functions of the critical facility were reported. With the completion of the critical facility, it has been made possible to demonstrate the establishment of the manufacturing techniques and product-quality guarantee for extremely pure isotropic graphite in addition to the reliability of the structural design and analytical techniques for the main unit of the critical facility. It is expected that the present facility will prove instrumental in the verification of the nuclear safety of the very high temperature gas-cooled nuclear reactor and in the acquisition of experimental data on the reactor physics pertaining to the improvement of the reactor characteristics. The tasks which remain to be accomplished hereafter are the improvements of the performance and quality features with regard to the oxidization of graphite, the heat-resisting structural materials, and the welded structures. (Kubozono, M.)

  10. The design of a fuel element for the RA-3 reactor (Ezeiza Atomic Center)

    International Nuclear Information System (INIS)

    Agueda, Horacio C.; Estevez, Esteban; Gerding, Jose R.; Markiewicz, Mario E.

    2003-01-01

    Some features of the mechanical design of the low enrichment fuel element for the RA-3 reactor are described, with emphasis in those aspects of the original design that have been modified considering the experience acquired in the design of other fuel elements. The proposed modification is based fundamentally on the replacement of all welded joints by screwed joints, which facilitates the manufacture of the fuel element, avoiding the distortions produced by the welds used at present and contributing to the fulfillment of the foreseen tolerances. A basic characteristic of this design is a careful manufacture of the fuel element's structural components in order to assure an assembling of the fuel element that fulfills the tolerances intrinsically required. The fuel is designed for the RA-3 reactor and uses U 3 O 8 or U 3 Si 2 as carrying phase of the fissile material with an enrichment of 19.70% of 235 U. The design verification was performed by analytical and numerical methods, and is supported by testing of materials in laboratory, hydrodynamics tests and performance evaluations of the fuel elements in the RA-3 reactor. (author)

  11. Reactor

    International Nuclear Information System (INIS)

    Ikeda, Masaomi; Kashimura, Kazuo; Inoue, Kazuyuki; Nishioka, Kazuya.

    1979-01-01

    Purpose: To facilitate the construction of a reactor containment building, whereby the inspections of the outer wall of a reactor container after the completion of the construction of the reactor building can be easily carried out. Constitution: In a reactor accommodated in a container encircled by a building wall, a space is provided between the container and the building wall encircling the container, and a metal wall is provided in the space so that it is fitted in the building wall in an attachable or detatchable manner. (Aizawa, K.)

  12. 76 FR 70166 - Electrical Standards for Construction and General Industry; Extension of the Office of Management...

    Science.gov (United States)

    2011-11-10

    ...] Electrical Standards for Construction and General Industry; Extension of the Office of Management and Budget... contained in the Electrical Standards for Construction (29 CFR part 1926, Subpart K) and for General... maintenance of electric utilization equipment that prevent death and serious injuries among construction and...

  13. Development of a general learning algorithm with applications in nuclear reactor systems

    International Nuclear Information System (INIS)

    Brittain, C.R.; Otaduy, P.J.; Perez, R.B.

    1989-12-01

    The objective of this study was development of a generalized learning algorithm that can learn to predict a particular feature of a process by observation of a set of representative input examples. The algorithm uses pattern matching and statistical analysis techniques to find a functional relationship between descriptive attributes of the input examples and the feature to be predicted. The algorithm was tested by applying it to a set of examples consisting of performance descriptions for 277 fuel cycles of Oak Ridge National Laboratory's High Flux Isotope Reactor (HFIR). The program learned to predict the critical rod position for the HFIR from core configuration data prior to reactor startup. The functional relationship bases its predictions on initial core reactivity, the number of certain targets placed in the center of the reactor, and the total exposure of the control plates. Twelve characteristic fuel cycle clusters were identified. Nine fuel cycles were diagnosed as having noisy data, and one could not be predicted by the functional relationship. 13 refs., 6 figs

  14. Development of a general learning algorithm with applications in nuclear reactor systems

    Energy Technology Data Exchange (ETDEWEB)

    Brittain, C.R.; Otaduy, P.J.; Perez, R.B.

    1989-12-01

    The objective of this study was development of a generalized learning algorithm that can learn to predict a particular feature of a process by observation of a set of representative input examples. The algorithm uses pattern matching and statistical analysis techniques to find a functional relationship between descriptive attributes of the input examples and the feature to be predicted. The algorithm was tested by applying it to a set of examples consisting of performance descriptions for 277 fuel cycles of Oak Ridge National Laboratory's High Flux Isotope Reactor (HFIR). The program learned to predict the critical rod position for the HFIR from core configuration data prior to reactor startup. The functional relationship bases its predictions on initial core reactivity, the number of certain targets placed in the center of the reactor, and the total exposure of the control plates. Twelve characteristic fuel cycle clusters were identified. Nine fuel cycles were diagnosed as having noisy data, and one could not be predicted by the functional relationship. 13 refs., 6 figs.

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

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  16. A Deeper Glimpse into the National Core Arts Standards for General Music

    Science.gov (United States)

    Zaffini, Erin Dineen

    2018-01-01

    The National Core Arts Standards in general music provide some exciting possibilities for music growth and understanding among our students. For those of us who are still unsure of how to read the standards or implement them in our classrooms, the standards also present some challenges for music educators. This article provides a deeper look into…

  17. General Atomic's 1500-MWe high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Waage, J.M.; Goodjohn, A.J.; Crimmins, D.H.; Wessman, G.L.

    1974-01-01

    Information is presented concerning the specifications for the plant layout, nuclear steam supply system, core and major components, prestressed concrete pressure vessel, control system, and safety systems. (U.S.)

  18. The Use Of High Speed Photography In Reactor Safety Studies At The Atomic Energy Establishment, Winfrith

    Science.gov (United States)

    Maddison, R. J.

    1985-02-01

    The investigation of certain areas of nuclear reactor safety involves the study of high speed phenomena with timescales ranging from microseconds to a few hundreds of milliseconds. Examples which have been extensively studied at Winfrith are firstly, the thermal interaction of molten fuel and reactor coolant which can generate high pressures on the 100 msec timescale, and which involves phenomena such as vapour film collapse which takes place on the microsecond timescale. Secondly, there is the response of reactor structures to such pressures, and finally there is the response of structural materials such as metals and concrete to the impulsive loading arising from the impact of heavy, high velocity missiles. A wide range of experimental techniques is used in these studies, many of which have been developed specially for this type of work which ranges from small laboratory scale to large field scale experiments. There are two important features which characterise many of these experiments:- i) a long period of meticulous preparation of very heavily instrumented, short duration experiments and; ii) the destructive nature of the experiments. Various forms of High Speed photography are included in the inventory of experimental techniques. These include the use of single and double exposure, short duration, spark photography; the use of an Image Convertor Camera (IMACON 790); and a number of rotating prism cine cameras. High Speed Photography is used both in a primary experimental role in the studies, and in a supportive role for other instrumentation. Because of the sometimes violent nature of these experiments, cameras are often heavily protected and operated remotely; lighting systems are sometimes destroyed. This has led to the development of unconventional techniques for camera operation and subject lighting. This paper will describe some of the experiments and the way in which High Speed Photography has been applied as an essential experimental tool. It will be

  19. Standard deviation of local tallies in global Monte Carlo calculation of nuclear reactor core

    International Nuclear Information System (INIS)

    Ueki, Taro

    2010-01-01

    Time series methodology has been studied to assess the feasibility of statistical error estimation in the continuous space and energy Monte Carlo calculation of the three-dimensional whole reactor core. The noise propagation was examined and the fluctuation of track length tallies for local fission rate and power has been formally shown to be represented by the autoregressive moving average process of orders p and p-1 [ARMA(p,p-1)], where p is an integer larger than or equal to two. Therefore, ARMA(p,p-1) fitting was applied to the real standard deviation estimation of the power of fuel assemblies at particular heights. Numerical results indicate that straightforward ARMA(3,2) fitting is promising, but a stability issue must be resolved toward the incorporation in the distributed version of production Monte Carlo codes. The same numerical results reveal that the average performance of ARMA(3,2) fitting is equivalent to that of the batch method with a batch size larger than 100 and smaller than 200 cycles for a 1,100 MWe pressurized water reactor. (author)

  20. A computer program for accident calculations of a standard pressurized water reactor

    International Nuclear Information System (INIS)

    Keutner, H.

    1979-01-01

    In this computer program the dynamic of a standard pressurized water reactor should be realized by both circulation loops with all important components. All important phenomena are taken into consideration, which appear for calculation of disturbances in order to state a realistic process for some minutes after a disturbance or a desired change of condition. In order to optimize the computer time simplifications are introduced in the statement of a differential-algebraic equalization system such that all important effects are taken into consideration. The model analysis starts from the heat production of the fuel rod via cladding material to the cooling medium water and considers the delay time from the core to the steam generator. Alternations of the cooling medium pressure as well as the different temperatures in the primary loop influence the pressuring system - the pressurizer - which is realized by a water and a steam zone with saturated and superheated steam respectively saturated and undercooled water with injection, heating and blow-down devices. The bilance of the steam generator to the secondary loop realizes the process engineering devices. Thereby the control regulation of the steam pressure and the reactor performance is realized. (orig.) [de

  1. Standardization of advanced light water reactors and progress on achieving utility requirements

    International Nuclear Information System (INIS)

    Marston, T.U.; Layman, W.H.; Bockhold, G. Jr.

    1992-01-01

    This paper reports that for a number of years, the U.S. utilities had led an industry-wide effort to establish a technical foundation for the design of the next generation of light water reactors in the United States. Since 1985, this utility initiative has been effected through a major technical program managed by the Electric Power Research Institute (EPRI); the U.S. Advanced Light Water Reactor (ALWR) Program. In addition to the U.S. utility leadership and sponsorship, the ALWR Program also has the participation and sponsorship of a number of international utility companies and close cooperation with the U.S. Department of Energy (DOE). The NPOC Strategic Plan for Building New Nuclear Plants creates a framework within which new standardized nuclear plants may be built. The Strategic Plan is an expression of the nuclear energy industry's serious intent to create the necessary conditions for new plant construction and operation. The industry has assembled a comprehensive, integrated list of actions that must be taken before new plants will be built and assigns responsibility for managing the various issues and sets time-tables and milestones against which we must measure progress

  2. Process for separately recovering uranium, transuranium elements, and fission products of uranium from atomic reactor fuel

    International Nuclear Information System (INIS)

    Balal, A.L.; Metscher, K.; Muehlig, B.; Reichmuth, C.; Schwarz, B.; Zimen, K.E.

    1976-01-01

    Spent reactor fuel elements are dissolved in dilute nitric acid. After addition of acetic acid as a complexing agent, the nitric acid is partly decomposed and the mixture subjected to electrolysis while a carrier liquid, which may be dilute acetic acid or a dilute mixture of acetic acid and nitric acid is caused to flow in the electric field between the electrodes either against the direction of ion migration or transversely thereto. The ions of uranium, plutonium, and other transuranium elements, and of fission products accumulate in discrete portions of the electrolyte and are separately withdrawn as at least three fractions after one or more stages of electrolysis

  3. Variegated operation of MAPS reactors after enmasse' coolant channel replacement: a tale-tell signature of high standard fuel bundle production quality

    International Nuclear Information System (INIS)

    Jena, J.K.; Sahu, J.K.; Arularasan, V.; Sivagurnathan, D.; Rathakrishnan, S.; Ramamurthy, K.

    2009-01-01

    After the Enmasse' Coolant Channel Replacement (EMCCR) of both the reactors of Madras Atomic Power Station (MAPS), they have put up a good performance, as far as core integrity is considered. This is a tale-tell signature of the high quality of the fuel bundles manufactured by Nuclear Fuel Complex (NFC), Hyderabad. Both the reactor cores have been loaded with various types of fuel bundles viz. Natural Uranium (NU), Depleted Uranium (DU), and Deeply Depleted Uranium (DDU) and were operated at different power level with different flux configuration at different stages of operation. Even around 1026 low burn up bundle (<2500 MWD/TeU) were transferred from MAPS-1 to MAPS-2, first time in the history of PHWRS. During all such variegated operations, the Primary Heat Transport (PHT) system 131 I activity, which is synonymous with the core integrity, was maintaining low for most of the reactor operation period. However, recently a low burn up fuel bundle failure has been observed in MAPS-1. Even though the overall failure rate is very low, the cause of such failure needs to be ascertained for taking appropriate action to maintain the high standards of quality in the manufacturing process of the fuel bundles. (author)

  4. Nuclear power reactors

    International Nuclear Information System (INIS)

    1982-11-01

    After an introduction and general explanation of nuclear power the following reactor types are described: magnox thermal reactor; advanced gas-cooled reactor (AGR); pressurised water reactor (PWR); fast reactors (sodium cooled); boiling water reactor (BWR); CANDU thermal reactor; steam generating heavy water reactor (SGHWR); high temperature reactor (HTR); Leningrad (RMBK) type water-cooled graphite moderated reactor. (U.K.)

  5. Experiences with interpretation and application of the German Atomic Energy Act for the German research reactors in Geesthacht FRG-1 and FRG-2

    International Nuclear Information System (INIS)

    Krull, W.

    1980-01-01

    The German research reactors FRG-1 and FRG-2 have passed different types of licensing procedures in the past years. It is reported about the experiences we have got in the interpretation and application of section 7 of the German Atomic Energy Act. Following these experiences an estimation is done for the licensing procedure for the reduction of the uranium enrichment. (orig.) [de

  6. General principles of nuclear safety management related to research reactor decommissioning

    International Nuclear Information System (INIS)

    Banciu, Ortenzia; Vladescu, Gabriela

    2003-01-01

    The paper contents the general principles applicable to the decommissioning of research reactors to ensure a proper nuclear safety management, during both decommissioning activities and post decommissioning period. The main objective of decommissioning is to ensure the protection of workers, population and environment against all radiological and non-radiological hazards that could result after a reactor shutdown and dismantling. In the same time, it is necessary, by some proper provisions, to limit the effect of decommissioning for the future generation, according to the new Romanian, IAEA and EU Norms and Regulations. Assurance of nuclear safety during decommissioning process involves, in the first step, to establish of some safety principles and requirements to be taken into account during whole process. In the same time, it is necessary to perform a series of analyses to ensure that the whole process is conducted in a planned and safe manner. The general principles proposed for a proper management of safety during research reactor decommissioning are as follows: - Set-up of all operations included in a Decommissioning Plan; - Set-up and qualitative evaluation of safety problems, which could appear during normal decommissioning process, both radiological and nonradiological risks for workers and public; - Set-up of accident list related to decommissioning process the events that could appear both due to some abnormal working conditions and to some on-site and off-site events like fires, explosions, flooding, earthquake, etc.); - Development and qualitative/ quantitative evaluation of scenarios for each incidents; - Development (and evaluation) of safety indicator system. The safety indicators are the most important tools used to assess the level of nuclear safety during decommissioning process, to discover the weak points and to establish safety measures. The paper contains also, a safety case evaluation (description of facility according to the decommissioning

  7. Validation and standardization of the Generalized Anxiety Disorder Screener (GAD-7) in the general population.

    Science.gov (United States)

    Löwe, Bernd; Decker, Oliver; Müller, Stefanie; Brähler, Elmar; Schellberg, Dieter; Herzog, Wolfgang; Herzberg, Philipp Yorck

    2008-03-01

    The 7-item Generalized Anxiety Disorder Scale (GAD-7) is a practical self-report anxiety questionnaire that proved valid in primary care. However, the GAD-7 was not yet validated in the general population and thus far, normative data are not available. To investigate reliability, construct validity, and factorial validity of the GAD-7 in the general population and to generate normative data. Nationally representative face-to-face household survey conducted in Germany between May 5 and June 8, 2006. Five thousand thirty subjects (53.6% female) with a mean age (SD) of 48.4 (18.0) years. The survey questionnaire included the GAD-7, the 2-item depression module from the Patient Health Questionnaire (PHQ-2), the Rosenberg Self-Esteem Scale, and demographic characteristics. Confirmatory factor analyses substantiated the 1-dimensional structure of the GAD-7 and its factorial invariance for gender and age. Internal consistency was identical across all subgroups (alpha = 0.89). Intercorrelations with the PHQ-2 and the Rosenberg Self-Esteem Scale were r = 0.64 (P < 0.001) and r = -0.43 (P < 0.001), respectively. As expected, women had significantly higher mean (SD) GAD-7 anxiety scores compared with men [3.2 (3.5) vs. 2.7 (3.2); P < 0.001]. Normative data for the GAD-7 were generated for both genders and different age levels. Approximately 5% of subjects had GAD-7 scores of 10 or greater, and 1% had GAD-7 scores of 15 or greater. Evidence supports reliability and validity of the GAD-7 as a measure of anxiety in the general population. The normative data provided in this study can be used to compare a subject's GAD-7 score with those determined from a general population reference group.

  8. The development of professional practice standards for Australian general practice nurses.

    Science.gov (United States)

    Halcomb, Elizabeth; Stephens, Moira; Bryce, Julianne; Foley, Elizabeth; Ashley, Christine

    2017-08-01

    The aim of this study was to explore the current role of general practice nurses and the scope of nursing practice to inform the development of national professional practice standards for Australian general practice nurses. Increasing numbers of nurses have been employed in Australian general practice to meet the growing demand for primary care services. This has brought significant changes to the nursing role. Competency standards for nurses working in general practice were first developed in Australia in 2005, but limited attention has been placed on articulating the contemporary scope of practice for nurses in this setting. Concurrent mixed methods design. Data collection was conducted during 2013-2014 and involved two online surveys of Registered and Enrolled Nurses currently working in general practice, a series of 14 focus groups across Australia and a series of consultations with key experts. Data collection enabled the development of 22 Practice Standards separated into four domains: (i) Professional Practice; (ii) Nursing Care; (iii) General Practice Environment and (iv) Collaborative Practice. To differentiate the variations in enacting these Standards, performance indicators for the Enrolled Nurse, Registered Nurse and Registered Nurse Advanced Practice are provided under each Standard. The development of national professional practice standards for nurses working in Australian general practice will support ongoing workforce development. These Standards are also an important means of articulating the role and scope of the nurses' practice for both consumers and other health professionals, as well as being a guide for curriculum development and measurement of performance. © 2017 John Wiley & Sons Ltd.

  9. Novel pathways for elimination of chlorine atoms from growing Si(100) surfaces in CVD reactors

    Science.gov (United States)

    Kunioshi, Nílson; Hagino, Sho; Fuwa, Akio; Yamaguchi, Katsunori

    2018-05-01

    Reactions leading to elimination of chlorine atoms from growing Si(100) surfaces were simulated using clusters of silicon atoms of different sizes and shapes, and at the UB3LYP/6-31 g(d,p) level of theory. The reactions of type SiCl2(s) + 2 H2(g), where (s) indicates an adsorbed species at the surface and (g) a gas-phase species, were found to proceed in two steps: SiCl2(s) + H2(g) → SiHCl(s) + HCl(g) and SiHCl(s) + H2(g) → SiH2(s) + HCl(g), each having activation energies around 55 kcal/mol, a value which is comparable to experimental values published in the literature. In addition, the results suggested that H-passivation of Si(100) surfaces support reactions leading to canonical epitaxial growth, providing a plausible explanation for the convenience of passivating the surfaces prior to silicon deposition. The reactions analyzed here can therefore be seen as important steps in the mechanism of epitaxial growth of Si(100) surfaces.

  10. Void nucleation by the helium atoms during lifetime of reactor pressure vessel

    International Nuclear Information System (INIS)

    Rahman, F.A.

    1984-01-01

    Void formation and growth has a great influence on the reactor pressure vessel steels during its lifetime and during post-irradiation annealing to increase its life. The present investigation aimed at the fact that if one can prevent void nucleation, accordingly one would not wary about void formation and growth. From that concept a model for helium production by transmutation reaction and corresponding swelling under irradiation conditions for several number of steels have been developed. This was done for recommending a steel type that can oppose such a phenomena. In the same time the present investigation gives a procedure utilizing such phenomena for checking the validity of pressure vessel steel used in the NPP

  11. Interim report on construction of data base for atomic energy science documents (concerning Kyoto University Reactor)

    International Nuclear Information System (INIS)

    Takeuchi, Takayuki

    1984-01-01

    The Kyoto University Research Reactor Institute was established in 1963 as a research institute for all universities in Japan utilizing the facilities in common. The construction of a document data base has been undertaken in commemoration of the 20th anniversary of the institute. The data base concerns the research works performed at the institute and also the publications and reports on the research made by the personnel belonging to the institute. Input data are gathered from concerned researchers. In this interim report, the structure and contents of this data base are shortly described. One of the features of this data base is that it handles data with both Japanese and English at the same time. (Aoki, K.)

  12. The application of the k0-standardization method at the TRIGA Mark II reactor, Ljubljana, Slovenia

    International Nuclear Information System (INIS)

    Jacimovic, Radojko; Benedik, Ljudmila; Stegnar, Peter; Smodis, Borut

    2002-01-01

    The k 0 -standardization method of neutron activation analysis (k 0 -NAA) was launched in the 1970s and since then continuously developed. Nowadays, k 0 -NAA became widespread as a practical analytical tool used to analyse different sample matrices. At the Jozef Stefan Institute (IJS), the KAYZERO/SOLCOI software package has been introduced for data processing after extensive testing and comparison with other available programs. In the process of validation of the software a suite of natural matrix reference materials (RMs) were used. Five certified reference materials (CRMs) from the Institute for Reference Materials and Measurements (IRMM), two standard reference materials (SRMs) from the National Institute of Standards and Technology (NIST), three RMs from the International Atomic Energy Agency (IAEA) and one RM from IJS were analysed. Altogether, results for ten elements in inorganic matrices and twenty-one elements in organic matrices, obtained by k 0 -instrumental neutron activation analysis (k 0 -INAA), were compared to certified values. The results obtained show good agreement with certified or assigned values except for Fe and U in inorganic matrices, and Al and Cr in organic matrices. (author)

  13. Standard reference material certification: contribution of NAA with a TRIGA reactor

    International Nuclear Information System (INIS)

    Orvini, E.; Speziali, M.; Salvini, A.; Herborg, C.

    2002-01-01

    Pavia has cooperative links with the major international agencies devoted to the certification of SRMs or CRMs as the Bureau Communautaire de Reference (BCR), the European Institute for Reference Materials and Measurement (IRMM), the USA National Institute of Standards and Technology (NIST) and the International Atomic Energy Agency (IAEA). During these cooperative works, a large amount of analytical data obtained with NAA has been compared, and meaningful methodological information achieved with respect to accuracy and precision in the analysis of several elements at different concentrations in various matrices. Analytical data on As, Cd, Cr, Co, Cu, Cs, Fe, Zn, K, Sc, U, Th, Al, Sb, Mn, V, Hg, Sr, Rb, Se,Pt, all the Rare Earths and halogens Br, Cl, I, have been obtained and contributed for the final certification

  14. Improvement of the homogeneity of atomized particles dispersed in high uranium density research reactor fuels

    International Nuclear Information System (INIS)

    Kim, Chang-Kyu; Kim, Ki-Hwan; Park, Jong-Man; Lee, Yoon-Sang; Lee, Don-Bae; Sohn, Woong-Hee; Hong, Soon-Hyung

    1998-01-01

    A study on improving the homogeneous dispersion of atomized spherical particles in fuel meats has been performed in connection with the development of high uranium density fuel. In comparing various mixing methods, the better homogeneity of the mixture could be obtained as in order of Spex mill, V-shape tumbler mixer, and off-axis rotating drum mixer. The Spex mill mixer required some laborious work because of its small capacity per batch. Trough optimizing the rotating speed parameter for the V-shape tumbler mixer, almost the same homogeneity as with the Spex mill could be obtained. The homogeneity of the extruded fuel meats appeared to improve through extrusion. All extruded fuel meats with U 3 Si powder of 50-volume % had fairly smooth surfaces. The homogeneity of fuel meats by V-shaped tumbler mixer revealed to be fairly good on micrographs. (author)

  15. Standard Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Results, E706(IA)

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2001-01-01

    1.1 This practice covers the methodology, summarized in Annex A1, to be used in the analysis and interpretation of neutron exposure data obtained from LWR pressure vessel surveillance programs; and, based on the results of that analysis, establishes a formalism to be used to evaluate present and future condition of the pressure vessel and its support structures (1-70). 1.2 This practice relies on, and ties together, the application of several supporting ASTM standard practices, guides, and methods (see Master Matrix E 706) (1, 5, 13, 48, 49). In order to make this practice at least partially self-contained, a moderate amount of discussion is provided in areas relating to ASTM and other documents. Support subject areas that are discussed include reactor physics calculations, dosimeter selection and analysis, and exposure units. Note 1—(Figure 1 is deleted in the latest update. The user is refered to Master Matrix E 706 for the latest figure of the standards interconnectivity). 1.3 This practice is restri...

  16. Prediction of enthalpy and standard Gibbs energy of vaporization of haloaromatics from atomic properties.

    Science.gov (United States)

    Monte, M J S; Almeida, A R R P; Liebman, J F

    2015-11-01

    Halogenated benzenes form a class of pollutants with a huge number of members - 1504 distinct benzene compounds, where one or more hydrogen atoms are replaced by halogens, may exist theoretically. This study presents a user friendly method for accurate prediction of vapor pressures and enthalpies of vaporization, at 298.15 K, of any mono or poly halobenzene compound. The derived equations for the prediction of those vaporization properties depend just on the number of each constituent halogen atom. This is a consequence of the absence of intramolecular interactions between the halogen atoms, revealed after examining vaporization results of ca. 40 halogenated benzenes. In order to rationalize the estimation equations, the contribution of the halogen atoms for the referred to above properties of vaporization was decomposed into two atomic properties - the volume and electron affinity. Extension of the applicability of the estimation method to substituted benzenes containing other substituent groups beyond halogen atoms as well as to some polycyclic aromatic species was tested with success. Copyright © 2015 Elsevier Ltd. All rights reserved.

  17. General views about specimen irradiations in reactors; Considerations generales sur'les irradiations d'echantillons dans les reacteurs

    Energy Technology Data Exchange (ETDEWEB)

    Seguin, M [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

    1965-07-01

    Specimen irradiation of fissile or non-fissile materials, carried out under circumstances becoming more and more severe and in reactor of increasing flux bas led to an evolution of irradiation rigs. A survey of the problems arising from irradiating under these various circumstances leads to conclude that it is possible to devise one capsule type suitable to every particular case, and that in a wide temperature range. Consequently, once the various irradiation-parameters known, a general method of calculation can be followed so as to determine the various sizes of the parts constituting the capsule. These theoretical calculations might sometimes be corrected through benefits gained from previous irradiations. Similarly, practical experimentation might allow to foresee more handy assembling of the capsule, specimen loading-and unloading being easier at the same time. (author) [French] L'irradiation d'echantillons, fissiles ou non fissiles, dans des conditions imposees de plus en plus strictes et dans des reacteurs a flux de plus en plus eleve, a eu pour consequence une evolution dans la conception des dispositifs d'irradiation. Lorsqu'on examine les problemes souleves par ces differentes irradiations, on en conclut qu'il est possible de concevoir un type de capsule capable de donner satisfaction dans chaque cas particulier, et ce, dans une tres large gamme de temperature. Par consequent, les differents parametres de l'irradiation etant connus, une methode generale de calcul peut etre suivie pour determiner les differentes cotes des pieces constitutives de la capsule. Ces calculs theoriques devront quelquefois etre corriges grace aux enseignements tires d'irradiations precedentes. De meme, l'experience acquise permettra d'envisager un montage plus aise de la capsule, tout en facilitant l'enfournement et le defournement des echantillons.

  18. The Atomic Spectroscopy Data Center at the National Institute of Standards and Technology (NIST). Activities 1999-2001

    International Nuclear Information System (INIS)

    Wiese, W.L.

    2001-01-01

    Dr. Wiese discussed activities and trends at the NIST Data Centers in the last two years. He reviewed priorities covered in data work and reviewed the bibliographic and numerical databases now on their website. The Atomic Spectra Database (ASD) is their main atomic physics web database and this is a reference data, e.g., the wavelength data is generally accurate to six significant figures and transition probability data is certain to with less than ±50%. Dr. Wiese also reported about recent work on the compilation and evaluation of data for wavelengths and energy levels of elements Cu, Kr and Mo (and several others), which are fusion relevant

  19. The Atomic Spectroscopy Data Center at the National Institute of Standards and Technology (NIST). Activities 1999-2001

    Energy Technology Data Exchange (ETDEWEB)

    Wiese, W L

    2001-12-01

    Dr. Wiese discussed activities and trends at the NIST Data Centers in the last two years. He reviewed priorities covered in data work and reviewed the bibliographic and numerical databases now on their website. The Atomic Spectra Database (ASD) is their main atomic physics web database and this is a reference data, e.g., the wavelength data is generally accurate to six significant figures and transition probability data is certain to with less than {+-}50%. Dr. Wiese also reported about recent work on the compilation and evaluation of data for wavelengths and energy levels of elements Cu, Kr and Mo (and several others), which are fusion relevant.

  20. Steel fibre concrete, a safer material for reactor construction. A general theory for rupture prediction

    International Nuclear Information System (INIS)

    Rammant, J.P.; Van Laethem, L.; Backx, E.

    1977-01-01

    The effect of steel fibre reinforcement on the mechanical behavior of concrete reactor structures is studied. It is shown that this material leads to a higher safety factor for highly stressed concrete structures like prestressed concrete pressure vessels. The reinforcement of concrete with short steel fibres results clearly in a fundamental change of the material properties. The study comprises basic experiments, the elaboration of an expression of the material laws, the development of a general computer program and the comparison of computational results with more elaborate experiments. Basic experimental work is conducted to determine the material characteristics of the fibre reinforced concrete. It is shown how the fibre reinforcement mechanism is translated into mathematical formulae by expressing the principal characteristics as matrix relationships. These relationships describe the elasto-plastic behavior and the cracked behavior. Probabilistic principles are used to express to fibre efficiency, such that a general stress-strain relationship is incorporated in a subsequent computer program. A general finite element program is developed which includes the new matrix relationships, the pull-out of fibres and the general stress-strain equations. A nonlinear calculation method gives the propagation of the distributed cracks with increasing load untill failure of the structure. Similarly, thermal cycling conditions are accounted for. For example the crack propagation in a fibre reinforced beam was measured by the photostress coating technique: the comparison with the computed crack propagation reveals an excellent agreement. Other comparative studies on simple structural parts are also reported

  1. Reference and standard benchmark field consensus fission yields for U.S. reactor dosimetry programs

    International Nuclear Information System (INIS)

    Gilliam, D.M.; Helmer, R.G.; Greenwood, R.C.; Rogers, J.W.; Heinrich, R.R.; Popek, R.J.; Kellogg, L.S.; Lippincott, E.P.; Hansen, G.E.; Zimmer, W.H.

    1977-01-01

    Measured fission product yields are reported for three benchmark neutron fields--the BIG-10 fast critical assembly at Los Alamos, the CFRMF fast neutron cavity at INEL, and the thermal column of the NBS Research Reactor. These measurements were carried out by participants in the Interlaboratory LMFBR Reaction Rates (ILRR) program. Fission product generation rates were determined by post-irradiation analysis of gamma-ray emission from fission activation foils. The gamma counting was performed by Ge(Li) spectrometry at INEL, ANL, and HEDL; the sample sent to INEL was also analyzed by NaI(Tl) spectrometry for Ba-140 content. The fission rates were determined by means of the NBS Double Fission Ionization Chamber using thin deposits of each of the fissionable isotopes. Four fissionable isotopes were included in the fast neutron field measurements; these were U-235, U-238, Pu-239, and Np-237. Only U-235 was included in the thermal neutron yield measurements. For the fast neutron fields, consensus yields were determined for three fission product isotopes--Zr-95, Ru-103, and Ba-140. For these fission product isotopes, a separately activated foil was analyzed by each of the three gamma counting laboratories. The experimental standard deviation of the three independent results was typically +- 1.5%. For the thermal neutron field, a consensus value for the Cs-137 yield was also obtained. Subsidiary fission yields are also reported for other isotopes which were studied less intensively (usually by only one of the participating laboratories). Comparisons with EBR-II fast reactor yields from destructive analysis and with ENDF/B recommended values are given

  2. The Use of Gas Chromatography and Mass Spectrometry to Introduce General Chemistry Students to Percent Mass and Atomic Mass Calculations

    Science.gov (United States)

    Pfennig, Brian W.; Schaefer, Amy K.

    2011-01-01

    A general chemistry laboratory experiment is described that introduces students to instrumental analysis using gas chromatography-mass spectrometry (GC-MS), while simultaneously reinforcing the concepts of mass percent and the calculation of atomic mass. Working in small groups, students use the GC to separate and quantify the percent composition…

  3. Use of a PhET Interactive Simulation in General Chemistry Laboratory: Models of the Hydrogen Atom

    Science.gov (United States)

    Clark, Ted M.; Chamberlain, Julia M.

    2014-01-01

    An activity supporting the PhET interactive simulation, Models of the Hydrogen Atom, has been designed and used in the laboratory portion of a general chemistry course. This article describes the framework used to successfully accomplish implementation on a large scale. The activity guides students through a comparison and analysis of the six…

  4. 77 FR 65765 - Petition for Exemption From the Federal Motor Vehicle Motor Theft Prevention Standard; General...

    Science.gov (United States)

    2012-10-30

    ... the vehicle. The antenna module translates the radio frequency signal received from the key into a... DEPARTMENT OF TRANSPORTATION National Highway Traffic Safety Administration Petition for Exemption From the Federal Motor Vehicle Motor Theft Prevention Standard; General Motors Corporation AGENCY...

  5. 77 FR 25534 - Petition for Exemption From the Federal Motor Vehicle Motor Theft Prevention Standard; General...

    Science.gov (United States)

    2012-04-30

    ... response back to the vehicle. The antenna module translates the radio frequency signal received from the... DEPARTMENT OF TRANSPORTATION National Highway Traffic Safety Administration Petition for Exemption From the Federal Motor Vehicle Motor Theft Prevention Standard; General Motors Corporation AGENCY...

  6. 78 FR 77790 - Petition for Exemption From the Federal Motor Vehicle Theft Prevention Standard; General Motors...

    Science.gov (United States)

    2013-12-24

    ... vehicle. The antenna module translates the radio frequency signal received from the key into a digital... DEPARTMENT OF TRANSPORTATION National Highway Traffic Safety Administration Petition for Exemption From the Federal Motor Vehicle Theft Prevention Standard; General Motors Corporation AGENCY: National...

  7. NRC [Nuclear Regulatory Commission] staff evaluation of the General Electric Company Nuclear Reactor Study (''Reed Report'')

    International Nuclear Information System (INIS)

    1987-07-01

    In 1975, the General Electric Company (GE) published a Nuclear Reactor Study, also referred to as ''the Reed Report,'' an internal product-improvement study. GE considered the document ''proprietary'' and thus, under the regulations of the Nuclear Regulatory Commission (NRC), exempt from mandatory public disclosure. Nonetheless, members of the NRC staff reviewed the document in 1976 and determined that it did not raise any significant new safety issues. The staff also reached the same conclusion in subsequent reviews. However, in response to recent inquiries about the report, the staff reevaluated the Reed Report from a 1987 perspective. This re-evaluation, documented in this staff report, concluded that: (1) there are no issues raised in the Reed Report that support a need to curtail the operation of any GE boiling water reactor (BWR); (2) there are no new safety issues raised in the Reed Report of which the staff was unaware; and (3) although certain issues addressed by the Reed Report are still being studied by the NRC and the industry, there is no basis for suspending licensing and operation of GE BWR plants while these issues are being resolved

  8. General layout of a 1300 MW PWR buildings: standard lay-out and influence factors

    International Nuclear Information System (INIS)

    Silva Messias, M. da

    1986-01-01

    The Standardization concept of a 1300 MW Nuclear Power Plant equipped with pressurized-water reactors (PWR) will be considered with regard to the factors which have various degrees of influence on the Layout of the Site. Standardization with regard to the Layout of nuclear power plant, stands, however also for a fixed arrangement of the components in the main buildings and a fixed arrangement of the buildings in the site plan. Standardization may result in simplified and shorter licensing procedures and mean less planning effort. Concentration of systems, optimization of the electro-mechanical design, clear separation of the controlled area, need for shielding, maintenance and safety aspects are some of number of factors which have various degress of influence on the layout of the buildings. (Author) [pt

  9. 75 FR 69709 - Office of New Reactors; Notice of Availability of the Final Staff Guidance; Standard Review Plan...

    Science.gov (United States)

    2010-11-15

    ... the Final Staff Guidance; Standard Review Plan, Section 13.6.6, Revision 0 on Cyber Security Plan... Reports for Nuclear Power Plants,'' Section 13.6.6, Revision 0 on ``Cyber Security Plan'' (Agencywide... reviews to amendments to licenses for operating reactors or for activities associated with review of...

  10. Progress in the Use of Isotopes: The Atomic Triad - Reactors, Radioisotopes and Radiation

    Science.gov (United States)

    Libby, W. F.

    1958-08-04

    Recent years have seen a substantial growth in the use of isotopes in medicine, agriculture, and industry: up to the minute information on the production and use of isotopes in the U.S. is presented. The application of radioisotopes to industrial processes and manufacturing operations has expanded more rapidly than any one except its most ardent advocates expected. New uses and new users are numerous. The adoption by industry of low level counting techniques which make possible the use of carbon-14 and tritium in the control of industrial processes and in certain exploratory and research problems is perhaps most promising of current developments. The latest information on savings to industry will be presented. The medical application of isotopes has continued to develop at a rapid pace. The current trend appears to be in the direction of improvements in technique and the substitution of more effective isotopes for those presently in use. Potential and actual benefits accruing from the use of isotopes in agriculture are reviewed. The various methods of production of radioisotopes are discussed. Not only the present methods but also interesting new possibilities are covered. Although isotopes are but one of the many peaceful uses of the atom, it is the first to pay its way. (auth)

  11. 76 FR 12792 - Petition for Exemption From the Federal Motor Vehicle Motor Theft Prevention Standard; General...

    Science.gov (United States)

    2011-03-08

    ... From the Federal Motor Vehicle Motor Theft Prevention Standard; General Motors Corporation AGENCY..., Exemption from the Theft Prevention Standard. This petition is granted because the agency has determined... in reducing and deterring motor vehicle theft as compliance with the parts-marking requirements of...

  12. Use of PSA for design of emergency mitigation systems in a hydrogen production plant using General Atomics SI cycle technology. Section 2: Sulphuric acid decomposition

    International Nuclear Information System (INIS)

    Mendoza, A.; Nelson, P.F.; Francois, J.L.

    2010-01-01

    Throughout the past decades, the need to reduce greenhouse gas emissions has prompted the development of technologies for the production of clean fuels through the use of zero emissions primary energy resources, such as heat from high temperature nuclear reactors. One of the most promising of these technologies is the generation of hydrogen by the sulphur-iodine cycle coupled to a high temperature nuclear reactor, initially proposed by General Atomics. By its nature and because these will have to be large-scale plants, development of these technologies from its current phase to its procurement and construction phase, will have to incorporate emergency mitigation systems in all its sections and nuclear-chemical 'tie-in points' to prevent unwanted events that can compromise the integrity of the plant and the nearby population centres. For the particular case of the SI thermochemical cycle, a large number of safety studies have been developed; however, most of these studies have focused on hydrogen explosions and failures in the primary cooling system. While these are the most catastrophic events, it is also true that there are many other events that without having a direct impact on the nuclear-chemical coupling, could jeopardise plant operations, safety of people in nearby communities and bring economic consequences. This study examined one of these events, which is the formation of a toxic cloud driven by an uncontrolled leakage of concentrated sulphuric acid in the second section of the General Atomics SI cycle. In this section, the concentration of sulphuric acid is close to 90% in conditions of high temperature and positive pressure. Under these conditions, sulphuric acid and sulphur oxides from the reactor would immediately form a toxic cloud, that in contact with operators could cause fatalities, or could produce choking, respiratory problems and eye irritation to people in neighbouring towns. The methodology used for this analysis is the design based on

  13. Generalized Bethe-Negele inequalities for excited states in muonic atoms

    International Nuclear Information System (INIS)

    Klarsfeld, S.

    1976-11-01

    Rigorous upper and lower bounds are derived for the Bethe logarithms in excited states of muonic atoms. Comparison with previous empirical estimates shows that the latter are inadequate in certain cases

  14. Status of the Japanese decay heat standard

    International Nuclear Information System (INIS)

    Katakura, Jun-ichi

    1992-01-01

    Fission product decay heat power plays an important role in the safety evaluation of nuclear power plants, especially for the analysis of hypothetical reactor accident scenarios. The ANS-5.1 decay heat standard for safety evaluation issued in 1979 has been used widely, even in Japan. Since the issuance of the standard, several improvements have been made to measurements and summation calculations. Summation calculations, in particular, have improved because of the adoption of theoretically calculated decay energies for nuclides with incomplete decay data. Taking into consideration those improvements, the Atomic Energy Society of Japan (AESJ) organized a research committee on a standard for decay heat power in nuclear reactors in 1987. The committee issued its recommendation after more than 2 yr discussion. After the AESJ recommendation, the Nuclear Safety Commission of Japan also began to discuss whether the recommendation should be included in its regulatory guide. The commission concluded in 1992 that the recommendation should be approved for licensing analysis of reactors if three times the uncertainties attached to the recommendation are included in the analysis. The AESJ recommendation may now be used for the safety evaluation of reactors in Japan in addition to the standards already used, which include ANS-5.1 (1973), General Electric Corporation (GE) curve, and ANS-5.1 (1979)

  15. 42 CFR 403.206 - General standards for Medicare supplemental policies.

    Science.gov (United States)

    2010-10-01

    ... 42 Public Health 2 2010-10-01 2010-10-01 false General standards for Medicare supplemental policies. 403.206 Section 403.206 Public Health CENTERS FOR MEDICARE & MEDICAID SERVICES, DEPARTMENT OF HEALTH AND HUMAN SERVICES GENERAL PROVISIONS SPECIAL PROGRAMS AND PROJECTS Medicare Supplemental Policies...

  16. Molten salt reactors. Synthesis of studies realized between 1973 and 1983. General synthesis

    International Nuclear Information System (INIS)

    Hery, M.; Lecocq, A.

    1983-03-01

    After a brief recall of the MSBR project, French studies on molten salt reactors are summed up. Theoretical and experimental studies for a graphite moderated 1000 MWe reactor using molten Li, Be, Th and U fluorides cooled by salt-lead direct contact are given. These studies concern the core, molten salt chemistry, graphite, metals (molybdenum, alloy TZM), corrosion, reactor components [fr

  17. The standardisation of trace elements in international biological standard reference materials with neutron activation analysis and atomic absorption spectrophotometry

    International Nuclear Information System (INIS)

    Pieterse, H.

    1981-12-01

    An investigation was undertaken into the analytical procedures and the identification of problem areas, for the certification of a new biological standard reference material supplied by the International Atomic Energy Agency, namely, a human hair sample designated as HH-I. The analyses comprised the determination of the elements As, Cd, Co, Cr, Cu, Fe, Hg, Mn, Ni, Sb, Se, and Zn in the hair sample by using two analytical techniques, namely, Instrumental Neutron Activation Analysis and Atomic Absorption. Three other certified biological reference materials, namely, Orchard Leaves (ORCH-L), Sea Plant Material (SPM-I) and Copepod (MAA-I) were used as control standards. Determinations were made of the moisture content of the samples, using varying conditions of drying, and the necessary corrections were applied to all analytical results so that the final elemental values related to dry weight of samples. Attention was also given to the possible loss of specific elements during ashing of the samples prior to the actual instrumental analysis. The results obtained for the hair sample by the two techniques were in good agreement for the elements Co, Fe, Mn, and Zn, but did not agree for the elements Cr and Sb. As, Hg and Se could only be determined with Instrumental Neutron Activation Analysis, and Cd, Cu and Ni only with Atomic Absorption. Most of the results obtained for the three control standard reference materials were within the ranges specified for the individual elements in each sample. The analytical procedures used for determining Cd, Cr, Cu, Ni and Sb with Instrumental Neutron Activation Analysis and As, Cr, Sb and Se with Atomic Absorption, need further investigation. The measurement of the moisture content and the ashing of samples also require further investigation with a view to improving accuracy

  18. Study of essential safety features of a three-loop 1,000 MWe light water reactor (PWR) and a corresponding heavy water reactor (HWR) on the basis of the IAEA nuclear safety standards

    International Nuclear Information System (INIS)

    1989-02-01

    Based on the IAEA Standards, essential safety aspects of a three-loop pressurized water reactor (1,000 MWe) and a corresponding heavy water reactor were studied by the TUeV Baden e.V. in cooperation with the Gabinete de Proteccao e Seguranca Nuclear, a department of the Ministry which is responsible for Nuclear power plants in Portugal. As the fundamental principles of this study the design data for the light water reactor and the heavy water reactor provided in the safety analysis reports (KWU-SSAR for the 1,000 MWe PWR, KWU-PSAR Nuclear Power Plant ATUCHA II) are used. The assessment of the two different reactor types based on the IAEA Nuclear Safety Standards shows that the reactor plants designed according to the data given in the safety analysis reports of the plant manufacturer meet the design requirements laid down in the pertinent IAEA Standards. (orig.) [de

  19. Boron analyses in the reactor coolant system of French PWR by acid-base titration ([B]) and ICP-MS (10B atomic %): key to NPP safety

    International Nuclear Information System (INIS)

    Jouvet, Fabien; Roux, Sylvie; Carabasse, Stephanie; Felgines, Didier

    2012-09-01

    Boron is widely used by Nuclear Power Plants and especially by EDF Pressurized Water Reactors to ensure the control of the neutron rate in the reactor coolant system and, by this way, the fission reaction. The Boron analysis is thus a major factor of safety which enables operators to guarantee the permanent control of the reactor. Two kinds of analyses carried out by EDF on the Boron species, recently upgraded regarding new method validation standards and developed to enhance the measurement quality by reducing uncertainties, will be discussed in this topic: Acid-Base titration of Boron and Boron isotopic composition by Inductively Coupled Plasma Mass Spectrometer - ICP MS. (authors)

  20. General Atomic HTGR fuel reprocessing pilot plant: results of initial sequential equipment operation

    International Nuclear Information System (INIS)

    1978-09-01

    In September 1977, the processing of 20 large high-temperature gas-cooled reactor (LHTGR) fuel elements was completed sequentially through the head-end cold pilot plant equipment. This report gives a brief description of the equipment and summarizes the results of the sequential operation of the pilot plant. 32 figures, 15 tables

  1. Higher order generalized perturbation theory for boiling water reactor in-core fuel management optimization

    International Nuclear Information System (INIS)

    Moore, B.R.; Turinsky, P.J.

    1998-01-01

    Boiling water reactor (BWR) loading pattern assessment requires solving the two-group, nodal form of the neutron diffusion equation and drift-flux form of the fluid equations simultaneously because these equation sets are strongly coupled via nonlinear feedback. To reduce the computational burden associated with the calculation of the core attributes (that is, core eigenvalue and thermal margins) of a perturbed BWR loading pattern, the analytical and numerical aspects of a higher order generalized perturbation theory (GPT) method, which correctly addresses the strong nonlinear feedbacks of two-phase flow, have been established. Inclusion of Jacobian information in the definition of the generalized flux adjoints provides for a rapidly convergent iterative method for solution of the power distribution and eigenvalue of a loading pattern perturbed from a reference state. Results show that the computational speedup of GPT compared with conventional forward solution methods demanding consistent accuracy is highly dependent on the number of spatial nodes utilized by the core simulator, varying from superior to inferior performance as the number of nodes increases

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

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

  3. Reactors

    DEFF Research Database (Denmark)

    Shah, Vivek; Vaz Salles, Marcos António

    2018-01-01

    The requirements for OLTP database systems are becoming ever more demanding. Domains such as finance and computer games increasingly mandate that developers be able to encode complex application logic and control transaction latencies in in-memory databases. At the same time, infrastructure...... engineers in these domains need to experiment with and deploy OLTP database architectures that ensure application scalability and maximize resource utilization in modern machines. In this paper, we propose a relational actor programming model for in-memory databases as a novel, holistic approach towards......-level function calls. In contrast to classic transactional models, however, reactors allow developers to take advantage of intra-transaction parallelism and state encapsulation in their applications to reduce latency and improve locality. Moreover, reactors enable a new degree of flexibility in database...

  4. 3. IAEA research co-ordination meeting on atomic and plasma-wall interaction data for fusion reactor divertor modeling. Summary report

    International Nuclear Information System (INIS)

    Janev, R.K.

    1999-04-01

    A brief description of the proceedings and the conclusions of the 3rd Research Co-ordination Meeting on 'Atomic and Plasma-Wall Interaction Data for Fusion Reactor Divertor Modeling', held on March 8-9, 1999, at the IAEA Headquarters in Vienna, Austria, is provided. The reports on the activities within the individual projects pertinent to the IAEA Co-ordinated Research program with the same title are given as appendix to the present report. (author)

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  6. Technological methods for the combined recovery of spent fuel elements from thermal neutron atomic reactors

    Energy Technology Data Exchange (ETDEWEB)

    Shevchenko, V B; Smelov, V S; Kozlov, A G; Smetanin, E Ya; Chubukov, V V; Anisimov, V I

    1976-01-01

    Results are presented for using a two-cycle extraction process to recover the uranium from a water-moderated nuclear-plant fuel cell after it had cooled 14 months. The general schematic is as follows: dissolution of the fuel in nitric acid; simultaneous extraction of U, Pu, and Np with tributyl phosphate; reduction with hydrazine and re-extraction of Pu and Np and their selective elution; re-examination of U; the regeneration of the extractant and beginning of the second cycle. The purification coefficients of U from other ..gamma..-radiating products and from Pu were equal to 1.4 x 10/sup 7/ and 5.5 x 10/sup 5/. The losses of U, Pu, and Np with the washing solutions were 0.007, 0.024, and 3.4% respectively of the initial amount. An analogous extraction process is discussed for separating Sr, Am, Cm, and the rare earth elements. A recovery of 92 to 95% of each of the four is effected.

  7. Standardization of digestion procedure for the determination of heavy metals in biological materials by atomic absorption spectrometry

    International Nuclear Information System (INIS)

    Khalid, N.; Chaudhri, S.A.

    1999-01-01

    Proper decomposition of the sample is one of the basic requirements of the atomic absorption spectroscopic analysis. In the present studies, heavy metals (Cu, Fe, Mn and Zn) were determined in biological samples by designating them in a mixture of nitric acid and perchloric acid. The quantification was made with atomic absorption spectrometry using an air-acetylene flame. The reliability of the procedure used was checked by analysing standard reference materials from NBS and IAEA, such as Rice flour (NBS-SRM-1568), Horse Kidney (IAEA H-8), Mixed Human diet(IAEA H-9), Copepod (IAEA MA-A-1) and fish flesh (IAEA MA-A-2) under identical conditions. A good agreement was observed between determined and the certified values reported by NBS and IAEA. (author)

  8. [The revised Dutch College of General Practitioners' standard on COPD and the first international WHO standard: differences and similarities].

    Science.gov (United States)

    van Schayck, C P

    2002-02-23

    Two standards on COPD have recently been published: the revised national standard from the Dutch College of General Practitioners and the first international standard published by the World Health Organization and the US National Heart, Lung and Blood Institute. The reduced emphasis on the role of spirometry in the monitoring and evaluation of treatment is an important change in these new standards compared to previous ones. Cessation of smoking is considered to be central to the prevention and treatment of COPD. Doctors should strongly support this approach and, more than before, are urged to view COPD as a disease caused by addiction. Bronchodilators are the cornerstone of symptomatic treatment of COPD, particularly the long-acting ones due to their ease of administration and effective treatment of morning dyspnoea. Inhalation corticosteroids should only be administered as a trial treatment and only under certain conditions. Continuation of treatment with these agents is only justified if there is a demonstrated improvement in lung function, exacerbations or symptoms, although the precise area of indication is not yet clear.

  9. Legal Time of the Republic of Colombia and its international traceability using the Cesium Atomic Clock - Time and Frequency National Standard

    Science.gov (United States)

    Hernández Forero, Liz Catherine; Bahamón Cortés, Nelson

    2017-06-01

    Around the world, there are different providers of timestamp (mobile, radio or television operators, satellites of the GPS network, astronomical measurements, etc.), however, the source of the legal time for a country is either the national metrology institute or another designated laboratory. This activity requires a time standard based on an atomic time scale. The International Bureau of Weights and Measures (BIPM) calculates a weighted average of the time kept in more than 60 nations and produces a single international time scale, called Coordinated Universal Time (UTC). This article presents the current time scale that generates Legal Time for the Republic of Colombia produced by the Instituto Nacional de Metrología (INM) using the time and frequency national standard, a cesium atomic oscillator. It also illustrates how important it is for the academic, scientific and industrial communities, as well as the general public, to be synchronized with this time scale, which is traceable to the International System (SI) of units, through international comparisons that are made in real time.

  10. Atomic near-degeneracy for photoemission : Generality of 4f excitations

    NARCIS (Netherlands)

    Bagus, Paul S.; Broer, R; Ilton, Eugene S.

    in a previous study of the 3s X-ray photoelectron spectra, XPS, of Mn, we identified a new intra-atomic many-body effect that lead to an similar to 50% increase in the predicted exchange splitting of the main high spin and low spin XPS peaks. The new many-body effect involved the promotion of one

  11. Standard generalized markup language: A guide for transmitting encoded bibliographic records

    International Nuclear Information System (INIS)

    1994-09-01

    This document provides the guidance necessary to transmit to DOE's Office of Scientific and Technical Information (OSTI) an encoded bibliographic record that conforms to International Standard ISO 8879, Information Processing -- Text and office systems -- Standard Generalized Markup Language (SGML). Included in this document are element and attribute tag definitions, sample bibliographic records, the bibliographic document type definition, and instructions on how to transmit a bibliographic record electronically to OSTI

  12. Standard generalized markup language: A guide for transmitting encoded bibliographic records

    Energy Technology Data Exchange (ETDEWEB)

    1994-09-01

    This document provides the guidance necessary to transmit to DOE`s Office of Scientific and Technical Information (OSTI) an encoded bibliographic record that conforms to International Standard ISO 8879, Information Processing -- Text and office systems -- Standard Generalized Markup Language (SGML). Included in this document are element and attribute tag definitions, sample bibliographic records, the bibliographic document type definition, and instructions on how to transmit a bibliographic record electronically to OSTI.

  13. General conditions for the existence of non-standard Lagrangians for dissipative dynamical systems

    International Nuclear Information System (INIS)

    Musielak, Z.E.

    2009-01-01

    Equations of motion describing dissipative dynamical systems with coefficients varying either in time or in space are considered. To identify the equations that admit a Lagrangian description, two classes of non-standard Lagrangians are introduced and general conditions required for the existence of these Lagrangians are determined. The conditions are used to obtain some non-standard Lagrangians and derive equations of motion resulting from these Lagrangians.

  14. TRIO a general computer code for reactor 3-D flows analysis. Application to a LMFBR hot plenum

    International Nuclear Information System (INIS)

    Magnaud, J.P.; Rouzaud, P.

    1985-09-01

    TRIO is a code developed at CEA to investigate general incompressible 2D and 3D viscous flows. Two calculations are presented: the lid driven cubic cavity at Re=400; steady state (velocity and temperature field) of a LMFBR hot plenum, carried out in order to prepare the calculation of a cold shock consecutive to a reactor scram. 8 refs., 26 figs.

  15. Setting Performance Standards for Technical and Nontechnical Competence in General Surgery.

    Science.gov (United States)

    Szasz, Peter; Bonrath, Esther M; Louridas, Marisa; Fecso, Andras B; Howe, Brett; Fehr, Adam; Ott, Michael; Mack, Lloyd A; Harris, Kenneth A; Grantcharov, Teodor P

    2017-07-01

    The objectives of this study were to (1) create a technical and nontechnical performance standard for the laparoscopic cholecystectomy, (2) assess the classification accuracy and (3) credibility of these standards, (4) determine a trainees' ability to meet both standards concurrently, and (5) delineate factors that predict standard acquisition. Scores on performance assessments are difficult to interpret in the absence of established standards. Trained raters observed General Surgery residents performing laparoscopic cholecystectomies using the Objective Structured Assessment of Technical Skill (OSATS) and the Objective Structured Assessment of Non-Technical Skills (OSANTS) instruments, while as also providing a global competent/noncompetent decision for each performance. The global decision was used to divide the trainees into 2 contrasting groups and the OSATS or OSANTS scores were graphed per group to determine the performance standard. Parametric statistics were used to determine classification accuracy and concurrent standard acquisition, receiver operator characteristic (ROC) curves were used to delineate predictive factors. Thirty-six trainees were observed 101 times. The technical standard was an OSATS of 21.04/35.00 and the nontechnical standard an OSANTS of 22.49/35.00. Applying these standards, competent/noncompetent trainees could be discriminated in 94% of technical and 95% of nontechnical performances (P < 0.001). A 21% discordance between technically and nontechnically competent trainees was identified (P < 0.001). ROC analysis demonstrated case experience and trainee level were both able to predict achieving the standards with an area under the curve (AUC) between 0.83 and 0.96 (P < 0.001). The present study presents defensible standards for technical and nontechnical performance. Such standards are imperative to implementing summative assessments into surgical training.

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

    International Nuclear Information System (INIS)

    2014-01-01

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

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

    International Nuclear Information System (INIS)

    2007-01-01

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

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

    International Nuclear Information System (INIS)

    2014-01-01

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

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

    International Nuclear Information System (INIS)

    2014-01-01

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

  20. Internal standardization in atomic-emission spectrometry using inductively coupled plasma

    International Nuclear Information System (INIS)

    Moore, G.L.

    1985-01-01

    The principle of internal standardization has been used in quantitative analytical emission spectroscopy since 1925 to minimize the errors arising from fluctuations in sample preparation, excitation-source conditions, and detection parameters. Although modern spectroscopic excitation sources are far more stable and electronic detection methods are more precise than before, the system for the introduction of the sample in spectrometric analysis using inductively coupled plasma (ICP) introduces significant errors, and internal standardization can still play a useful role in improving the overall precision of the analytical results. The criteria for the selection of the elements to be used as internal standards in arc and spark spectrographic analysis apply to a much lesser extent in ICP-spectrometric analysis. Internal standardization is recommended for use in routine ICP-simultaneous spectrometric analysis to improve its accuracy and precision and to provide a monitor for the reassurance of the analyst. However, the selection of an unsuitable reference element can result in misuse of the principle of internal standardization and, although internal standardization can be applied when a sequential monochromator is used, the main sources of error will not be minimized

  1. Numerical solutions of the aerosol general dynamic equation for nuclear reactor safety studies

    International Nuclear Information System (INIS)

    Park, J.W.

    1988-01-01

    Methods and approximations inherent in modeling of aerosol dynamics and evolution for nuclear reactor source term estimation have been investigated. Several aerosol evolution problems are considered to assess numerical methods of solving the aerosol dynamic equation. A new condensational growth model is constructed by generalizing Mason's formula to arbitrary particle sizes, and arbitrary accommodation of the condensing vapor and background gas at particle surface. Analytical solution is developed for the aerosol growth equation employing the new condensation model. The space-dependent aerosol dynamic equation is solved to assess implications of spatial homogenization of aerosol distributions. The results of our findings are as follows. The sectional method solving the aerosol dynamic equation is quite efficient in modeling of coagulation problems, but should be improved for simulation of strong condensation problems. The J-space transform method is accurate in modeling of condensation problems, but is very slow. For the situation considered, the new condensation model predicts slower aerosol growth than the corresponding isothermal model as well as Mason's model, the effect of partial accommodation is considerable on the particle evolution, and the effect of the energy accommodation coefficient is more pronounced than that of the mass accommodation coefficient. For the initial conditions considered, the space-dependent aerosol dynamics leads to results that are substantially different from those based on the spatially homogeneous aerosol dynamic equation

  2. Determination of H-atom reaction rate constants by the competition kinetic technique using riboflavin as a standard solute [Paper No. RD-7

    International Nuclear Information System (INIS)

    Kishore, Kamal; Moorthy, P.N.; Rao, K.N.

    1982-01-01

    Riboflavin has been used as a standard solute to evaluate H-atom rate constants of other solutes by steady state radiolytic competition kinetic method. The bleaching of absorbance of riboflavin at 445 nm as a result of its reaction with H-atoms is made use of in estimating its decomposition. The merits and demerits of this method are discussed. (author)

  3. Hermetic cable penetrations for containments of nuclear power reactors meet high safety standards

    International Nuclear Information System (INIS)

    Kusserow, J.; Gurr, W.; Pflug, H.

    1985-05-01

    Different types of cable penetrations for containments of nuclear power reactors have been developed and fabricated in the GDR. The technical parameters achieved are in accordance with the radiation protection requirements

  4. NJOY-97, General ENDF/B Processing System for Reactor Design Problems

    International Nuclear Information System (INIS)

    1999-01-01

    1 - Description of program or function: The NJOY nuclear data processing system is a modular computer code used for converting evaluated nuclear data in the ENDF format into libraries useful for applications calculations. Because the Evaluated Nuclear Data File (ENDF) format is used all around the world (e.g., ENDF/B-VI in the US, JEF-2.2 in Europe, JENDL-3.2 in Japan, BROND-2.2 in Russia), NJOY gives its users access to a wide variety of the most up-to-date nuclear data. NJOY provides comprehensive capabilities for processing evaluated data, and it can serve applications ranging from continuous-energy Monte Carlo (MCNP), through deterministic transport codes (DANT, ANISN, DORT), to reactor lattice codes (WIMS, EPRI). NJOY handles a wide variety of nuclear effects, including resonances, Doppler broadening, heating (KERMA), radiation-damage, thermal scattering (even cold moderators), gas production, neutrons and charged particles, photo-atomic interactions, self shielding, probability tables, photon production, and high-energy interactions (to 150 MeV). Output can include printed listings, special library files for applications, and Postscript graphics (plus colour). More information on NJOY is available from the developer's home page at http://t2.lanl.gov. Follow the Tourbus section of the Tour area to find notes from the ICTP lectures held at Trieste in March 1998 on the ENDF format and on the NJOY code. 2 - Methods: NJOY97 consists of a set of modules, each performing a well-defined processing task. Each of these modules is essentially a separate computer program linked together by input and output files and a few common constants. The methods and instructions on how to use them are documented in the LA-12740-M report on NJOY91 and in the 'README' file. No new published document is yet available. NJOY97 is a cleaned up version of NJOY94.105 that features compatibility with a wider variety of compilers and machines, explicit double precision for 32-bit systems, a

  5. Seismic examination for assessment of safety of location of atomic energy objects (by the example of the WWR-K reactor, Ala-Tau village)

    International Nuclear Information System (INIS)

    Belyashova, N.N.

    2001-01-01

    In the Republic of Kazakhstan there are 3 research reactors (the fourth one is temporarily stopped). One of the reactors in 1998 (WWR-K, situated in the Ala Tau village, nearby Almaty city) was conserved because of a number of reasons. Including the reason of the earth crust geological structure insufficient study for the ensuring the seismic safety of the reactor site location. In 1994-1996 a number of geological-geophysical studies was carried out by Kazakhstan specialists confirming the the geological-geophysical conditions in the reactor site location in view of its safety. These condition are meeting to IAEA requirements and up-to-date standards acting in Kazakhstan

  6. A robust and general Schrödinger and Dirac solver for atomic structure calculations

    Czech Academy of Sciences Publication Activity Database

    Čertík, O.; Pask, J.E.; Vackář, Jiří

    2013-01-01

    Roč. 184, č. 7 (2013), s. 1777-1791 ISSN 0010-4655 R&D Projects: GA MŠk(CZ) LC06040; GA ČR GA101/09/1630 Institutional support: RVO:68378271 Keywords : atom * electronic structure * Dirac equation * density-functional theory Subject RIV: BE - Theoretical Physics Impact factor: 2.407, year: 2013 http://www.sciencedirect.com/science/article/pii/S0010465513000714

  7. General theory of the ionization of an atom by an electrostatic field

    International Nuclear Information System (INIS)

    Fonda, L.

    1981-05-01

    The ionization of an atom by an external electrostatic field is reconsidered by taking into account the interactions of the system with the measuring apparatus. The experimental ionization rate is drastically different from the expression obtained when no measurements are present. A dependence on the mean frequency of measurements is found. This fact can be used to determine this quantity once the ionization rate is determined experimentally. (author)

  8. General description of preliminary design of an experimental fusion reactor and the future problems

    International Nuclear Information System (INIS)

    Sako, Kiyoshi

    1976-01-01

    Recently, the studies on plasma physics has progressed rapidly, and promising experimental data emerged successively. Especially expectation mounts high that Tokamak will develop into power reactors. In Japan, the construction of large plasma devices such as JT-60 of JAERI is going to start, and after several years, the studies on plasma physics will come to the end of first stage, then the main research and development will be directed to power reactors. The studies on the design of practical fusion reactors have been in progress since 1973 in JAERI, and the preliminary design is being carried out. The purposes of the preliminary design are the clarification of the concept of the experimental reactor and the requirements for the studies on core plasma, the examination of the problems for developing main components and systems of the reactor, and the development of design technology. The experimental reactor is the quasi-steady reactor of 100 MW fusion reaction output, and the conditions set for the design and the basis of their setting are explained. The outline of the design, namely core plasma, blankets, superconductive magnets and the shielding with them, vacuum wall, neutral particle injection heating device, core fuel supply and exhaust system, and others, is described. In case of scale-up the reactor structural material which can withstand neutron damage must be developed. (Kako, I.)

  9. Correlation study among the International Atomic Energy Agency standards and market standards on management system applicable to a UF6 conversion plant operation

    International Nuclear Information System (INIS)

    Oliveira, Dirceu Paulo de

    2008-01-01

    The Agency - International Atomic Energy Agency (IAEA), following the market trend of the management system integration, has decided to revise the quality assurance standards - IAEA 50-C/SG-Q publishing, in 2006, the standard on Management System (MS). IAEA GS-R-3 and its IAEA GS-G-3.1 guide. Also, the IAEA is about to publish a supplementary guide - IAEA DS349, which consider the integration of several functions involved in management of nuclear facilities, such as: safety, health, environmental and quality, ensuring that nuclear safety is not compromised. Conversion plants of 'Yellowcake' in UF 6 use and process radioactive materials, as well as other substances normally found in the chemical conventional industry, inserting themselves in the organization profile that require a high pattern of definition, implementation and continuous improvement of their MS and, therefore, should consider an approach of management integrated system (MIS). Taking a UF 6 conversion plant as focus, the correlation was performed among the Agency MS standards and those of the market - ISO 9001, ISO 14001 and OHSAS 18001, as well as with the Agency drafts standards on safety (DS316 and DS344), concluding that, in structuring an MIS, in compliance with the Agency MS standards, except for some adjustments, the ISO 9001, ISO 14001, and OHSAS 18001 are going to be met. On the other hand, the structuring of MIS should identify other requirements on safety, health and environmental, which also consider the conventional chemical and industrial characteristics that are out of the scope (ionizing radiation) of the safety standards of the Agency. The research proposes a documental procedure for a MIS applicable to this plant, providing elements for rationalization and contents of the identified documentation, for the promotion of the integration of the considered MS functions. (author)

  10. Pin-wise Reactor Analysis Based on the Generalized Equivalence Theory

    Energy Technology Data Exchange (ETDEWEB)

    Yu, Hwan Yeal; Heo, Woong; Kim, Yong Hee [KAIST, Daejeon (Korea, Republic of)

    2016-05-15

    In this paper, a pin-wise reactor analysis is performed based on the generalized equivalence theory. From the conventional fuel assembly lattice calculations, pin-wise 2-group cross sections and pin DFs are generated. Based on the numerical results on a small PWR benchmark, it is observed that the pin-wise core analysis provide quite accurate prediction on the effective multiplication factor and the peak pin power error is bounded by about 3% in peripheral fuel assemblies facing the baffle-reflector. Also, it was found that relatively large pin power errors occur along the interface between clearly different fuel assemblies. It is expected that the GET-based pin-by-pin core calculation can be further developed as an advanced method for reactor analysis via improving the group constants and discontinuity factors. Recently, high-fidelity multi-dimensional analysis tools are gaining more attention because of their accurate prediction of local parameters for core design and safety assessment. In terms of accuracy, direct whole-core transport is quite promising. However, it is clear that it is still very costly in terms of the computing time and memory requirements. Another possible solution is the pin-by-pin core analysis in which only small fuel pins are homogenized and the 3-D core analysis is still performed using a low-order operator such as the diffusion theory. In this paper, a pin-by-pin core analysis is performed using the hybrid CMFD (HCMFD) method. Hybrid CMFD is a new global-local iteration method that has been developed for efficient parallel calculation of pinby-pin heterogeneous core analysis. For the HCMFD method, the one-node CMFD scheme is combined with a local two-node CMFD method in a non-linear way. Since the SPH method is iterative and SPH factors are not direction dependent, it is clear that SPH method takes more computing cost and cannot take into account the different heterogeneity and transport effects at each pin interface. Unlike the SPH

  11. Standards, intercomparisons and performance evaluations for low-level and environmental radionuclide mass spectrometry and atom counting

    International Nuclear Information System (INIS)

    Inn, K.G.W.; McCurdy, D.; Bell III, T.; Loesch, R.; Barss, N.M.; Morton, J.S.; Povinec, P.; Burns, K.; Henry, R.

    2001-01-01

    Because of the demand for higher sensitivity radionuclide measurements, atom counting technology will become an increasingly used modality in geo-and bio-studies, and process control operations. It is anticipated that requests, intercomparisons and performance evaluations services will surge in near future. In anticipation of such requests, the state-of-the-art needed to be assessed for proactive planning purposes. The results of a workshop focused on these issues indicated that there are several ongoing standards, intercomparisons and performance evaluations thrusts which are expected to expand in the future. Furthermore, new projects were planned and the Council on Ionizing Radiation Measurements and Standards (Public and Environmental Radiation Protection subcommittee) ws proposed as an information coordinator. (author)

  12. Standard Test Method for Hydrophobic Surface Films by the Atomizer Test

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    1965-01-01

    1.1 This test method covers the detection of the presence of hydrophobic (nonwetting) films on surfaces and the presence of hydrophobic organic materials in processing ambients. When properly conducted, the test will enable detection of fractional molecular layers of hydrophobic organic contaminants. On very rough or porous surfaces the sensitivity of the test may be significantly decreased. 1.2 The values stated in inch-pound units are to be regarded as the standard. The values given in parentheses are for information only. 1.3 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

  13. High-speed imaging upgrade for a standard sample scanning atomic force microscope using small cantilevers

    Energy Technology Data Exchange (ETDEWEB)

    Adams, Jonathan D.; Nievergelt, Adrian; Erickson, Blake W.; Yang, Chen; Dukic, Maja; Fantner, Georg E., E-mail: georg.fantner@epfl.ch [Ecole Polytechnique Fédérale de Lausanne, Lausanne (Switzerland)

    2014-09-15

    We present an atomic force microscope (AFM) head for optical beam deflection on small cantilevers. Our AFM head is designed to be small in size, easily integrated into a commercial AFM system, and has a modular architecture facilitating exchange of the optical and electronic assemblies. We present two different designs for both the optical beam deflection and the electronic readout systems, and evaluate their performance. Using small cantilevers with our AFM head on an otherwise unmodified commercial AFM system, we are able to take tapping mode images approximately 5–10 times faster compared to the same AFM system using large cantilevers. By using additional scanner turnaround resonance compensation and a controller designed for high-speed AFM imaging, we show tapping mode imaging of lipid bilayers at line scan rates of 100–500 Hz for scan areas of several micrometers in size.

  14. Measurement standards and the general problem of reference points in chemical analysis

    International Nuclear Information System (INIS)

    Richter, W.; Dube, G.

    2002-01-01

    Besides the measurement standards available in general metrology in the form of the realisations of the units of measurement, measurement standards of chemical composition are needed for the vast field of chemical measurement (measurements of the chemical composition), because it is the main aim of such measurements to quantify non-isolated substances, often in complicated matrices, to which the 'classical' measurement standards and their lower- level derivatives are not directly applicable. At present, material artefacts as well as standard measurement devices serve as chemical measurement standards. These are measurement standards in the full metrological sense only, however, if they are firmly linked to the SI unit in which the composition represented by the standard is expressed. This requirement has the consequence that only a very restricted number of really reliable chemical measurement standards exist at present. Since it is very difficult and time consuming to increase this number substantially and, on the other hand, reliable reference points are increasingly needed for all kinds of chemical measurements, primary methods of measurement and high-level reference measurements will play an increasingly important role for the establishment of worldwide comparability and hence mutual acceptance of chemical measurement results. (author)

  15. Reactor

    International Nuclear Information System (INIS)

    Fujibayashi, Toru.

    1976-01-01

    Object: To provide a boiling water reactor which can enhance a quake resisting strength and flatten power distribution. Structure: At least more than four fuel bundles, in which a plurality of fuel rods are arranged in lattice fashion which upper and lower portions are supported by tie-plates, are bundled and then covered by a square channel box. The control rod is movably arranged within a space formed by adjoining channel boxes. A spacer of trapezoidal section is disposed in the central portion on the side of the channel box over substantially full length in height direction, and a neutron instrumented tube is disposed in the central portion inside the channel box. Thus, where a horizontal load is exerted due to earthquake or the like, the spacers come into contact with each other to support the channel box and prevent it from abnormal vibrations. (Furukawa, Y.)

  16. Terms standardization between the rules of diagnosis radiation equipment safety management and atomic energy law: problems and suggestions

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hwa Gon; Kang, Se Sik; Kim, Chang Soo; Park, Cheol Seo [Catholic University of Pusan, Busan (Korea, Republic of)

    2006-03-15

    The rules and terms are described different meaning in this results the research is accomplished for preventing practical workers from confusion. Atomic law are kept up modification and development in our situation by the ICRP's recommendation, on the other hand, the rules of diagnosis radiation equipment safety managements are modified partial, then resulted in confusion. The study was comparison between the rules of diagnosis radiation equipment safety management and atomic energy law, and the modification items obtained were as follows. With each other different the terms and units are used. With the exception of special terms for affairs usage, it is needless to say that common term uniformity is standardized. The standardization of rules and guidance have not need to confusion radiological practical workers. The following is omitted. The radiation protection against the patient and the hospital visitor. Radiation dose limit of the woman patient who is in the process of becoming pregnant. Radiation dose limit of the person who is not regarded as medical exposure. The control of the exposure of pregnant of women at work.

  17. Terms standardization between the rules of diagnosis radiation equipment safety management and atomic energy law: problems and suggestions

    International Nuclear Information System (INIS)

    Kim, Hwa Gon; Kang, Se Sik; Kim, Chang Soo; Park, Cheol Seo

    2006-01-01

    The rules and terms are described different meaning in this results the research is accomplished for preventing practical workers from confusion. Atomic law are kept up modification and development in our situation by the ICRP's recommendation, on the other hand, the rules of diagnosis radiation equipment safety managements are modified partial, then resulted in confusion. The study was comparison between the rules of diagnosis radiation equipment safety management and atomic energy law, and the modification items obtained were as follows. With each other different the terms and units are used. With the exception of special terms for affairs usage, it is needless to say that common term uniformity is standardized. The standardization of rules and guidance have not need to confusion radiological practical workers. The following is omitted. The radiation protection against the patient and the hospital visitor. Radiation dose limit of the woman patient who is in the process of becoming pregnant. Radiation dose limit of the person who is not regarded as medical exposure. The control of the exposure of pregnant of women at work

  18. Digital control of research reactors

    International Nuclear Information System (INIS)

    Crump, J.C. III.; Richards, W.J.; Heidel, C.C.

    1991-01-01

    Research reactors provide an important service for the nuclear industry. Developments and innovations used for research reactors can be later applied to larger power reactors. Their relatively inexpensive cost allows research reactors to be an excellent testing ground for the reactors of tomorrow. One area of current interest is digital control of research reactor systems. Digital control systems offer the benefits of implementation and superior system response over their analog counterparts. At McClellan Air Force Base in Sacramento, California, the Stationary Neutron Radiography System (SNRS) uses a 1,000-kW TRIGA reactor for neutron radiography and other nuclear research missions. The neutron radiography beams generated by the reactor are used to detect corrosion in aircraft structures. While the use of the reactor to inspect intact F-111 wings is in itself noteworthy, there is another area in which the facility has applied new technology: the instrumentation and control system (ICS). The ICS developed by General Atomics (GA) contains several new and significant items: (a) the ability to servocontrol on three rods, (b) the ability to produce a square wave, and (c) the use of a software configurator to tune parameters affected by the actual reactor core dynamics. These items will probably be present in most, if not all, future research reactors. They were developed with increased control and overall usefulness of the reactor in mind

  19. Leukemia in atomic bomb survivors. 1. General observations. Leukemia in survivors of atomic bombing. Cytologic and biochemical studies on the granulocytes in early leukemia among atomic bomb survivors. Leukemogenic effects of ionizing radiation on atomic bomb survivors in Hiroshima City

    Energy Technology Data Exchange (ETDEWEB)

    Lange, R D; Moloney, W C; Yamawaki, Tokuso; Kastenbaum, M A

    1959-01-01

    This document contains 4 separate reports on leukemia in survivals of the atomic explosions in Hiroshima and Nagasaki. In the first report, observations on seventy-five established cases of leukemia occurring in people exposed to atomic bomb radiation are presented. These data indicate a great increase in the incidence of leukemia among atomic bomb survivors due to a single massive exposure to ionizing radiation. The leukemogenic effects of radiation are manifested equally in both sexes and at all age levels represented in this series. The striking preponderance of chronic myelogenous leukemia compared to chronic lymphatic leukemia has been noted in exposed individuals but it is pointed out that chronic lymphatic leukemia is comparatively rare among the Japanese. Cases of leukemia are still appearing in atomic bomb survivors. However, since 1950 there has been a steady decline in the number of cases. The second report consists of a review of all cases of leukemia referred to the ABCC from 1948 to April 1952, a total of 75 cases. In the third report, hematological and biochemical findings in separated leukocytes of four cases of preclinical myelogenous leukemia developing in atomic bomb survivors are described. The incidence of leukemia among survivors in Hiroshima is the topic of the fourth report. 38 references, 8 figures, 10 tables.

  20. Historical Review of Atomic Frequency Standards Used in Space Systems - 10 Year Update

    Science.gov (United States)

    2007-01-01

    section on 2006 predictions. The authors would like to thank Peter Cash, Bernardo Jaduszliwer, Bob Kern, Robert Lutwak , John Prestage, Bill Riley, and...258- 262. [17] R. Lutwak , D. Emmons, R. M. Garvey, and P. Vlitas, 2003, “Optically pumped cesium-beam frequency standard for GPS-III,” in

  1. The general atomic and molecular electronic structure system HONDO: Version 7.0

    International Nuclear Information System (INIS)

    Dupuis, M.; Watts, J.D.; Villar, H.O.; Hurst, G.J.B.

    1989-01-01

    We describe a computer program for ab initio quantum mechanical calculations of atomic and molecular wavefunctions and energies. Capabilities for the calculation of energy gradients and second derivatives with respect to nuclear coordinates are provided for several types of wavefunctions. Calculations of many molecular properties based on the electron density are possible. The program contains automated algorithms for the determination of equilibrium structures, saddle points, reaction pathways, vibrational spectra including infrared and Raman intensities. We illustrate the capabilities of the program by highlighting research problems recently investigated with the present program. (orig.)

  2. Generalized Kubo formulas for the transport properties of incommensurate 2D atomic heterostructures

    Science.gov (United States)

    Cancès, Eric; Cazeaux, Paul; Luskin, Mitchell

    2017-06-01

    We give an exact formulation for the transport coefficients of incommensurate two-dimensional atomic multilayer systems in the tight-binding approximation. This formulation is based upon the C* algebra framework introduced by Bellissard and collaborators [Coherent and Dissipative Transport in Aperiodic Solids, Lecture Notes in Physics (Springer, 2003), Vol. 597, pp. 413-486 and J. Math. Phys. 35(10), 5373-5451 (1994)] to study aperiodic solids (disordered crystals, quasicrystals, and amorphous materials), notably in the presence of magnetic fields (quantum Hall effect). We also present numerical approximations and test our methods on a one-dimensional incommensurate bilayer system.

  3. The effect of economic growth, oil prices, and the benefits of reactor standardization: Duration of nuclear power plant construction revisited

    International Nuclear Information System (INIS)

    Csereklyei, Zsuzsanna; Thurner, Paul W.; Bauer, Alexander; Küchenhoff, Helmut

    2016-01-01

    The profitability of nuclear power plant investment is largely determined by the construction duration, which directly impacts discounted cash flows, debt and interest payments, as well as variable costs, such as labor. This paper analyzes the key drivers of construction duration using survival models. We focus especially on the strategic expectation formation of private and public utilities engaging in such highly risky megaprojects. Using a balanced dataset of explanatory variables and the IAEA/PRIS dataset of reactor construction starts between 1950 and 2013 we find that the expectation of rising oil prices and higher economic growth, along with the higher per capita GDP of a country tend to reduce the time needed to grid connection. We also identify the reactor models with the fastest construction duration. - Highlights: • We find that higher future economic growth speeds up nuclear reactor construction. • Higher national capacity (measured by income per capita) results in faster projects. • Higher oil prices during construction lead to faster construction times. • Reactor standardization may result in faster building times.

  4. Standard Practice for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This practice covers procedures for designing a surveillance program for monitoring the radiation-induced changes in the mechanical properties of ferritic materials in light-water moderated nuclear power reactor vessels. This practice includes the minimum requirements for the design of a surveillance program, selection of vessel material to be included, and the initial schedule for evaluation of materials. 1.2 This practice was developed for all light-water moderated nuclear power reactor vessels for which the predicted maximum fast neutron fluence (E > 1 MeV) at the end of license (EOL) exceeds 1 × 1021 neutrons/m2 (1 × 1017 n/cm2) at the inside surface of the reactor vessel. 1.3 This practice applies only to the planning and design of surveillance programs for reactor vessels designed and built after the effective date of this practice. Previous versions of Practice E185 apply to earlier reactor vessels. 1.4 This practice does not provide specific procedures for monitoring the radiation induced cha...

  5. Relationship between general safety requirements and safety culture in the improvement of safe operation of I.N.R. TRIGA reactor facilities

    International Nuclear Information System (INIS)

    Ciocanescu, M.; Preda, M.; Chiritescu, M.; Dumitru, M.

    1996-01-01

    Acquiring of the basic principles of ''safety culture'' by a large number of profesionals in the nuclear field drew the attention of the decision factors in the INR managerial structure, who decided to promote certain practical actions at each level in order to improve nuclear safety. Starting from the ''Republican Standards for Nuclear Safety'' issued by CSEN in 1975, where general safety criteria are defined for nuclear reactors and NPPs, the specialists at the TRIGA reactor originated and implemented a coherent and secure system to ensure nuclear safety over all steps of nuclear activities: research, conception, execution, commissioning and operation. This system has been continuosly corrected so that now it is completely integrated in a modern safety system. The paper presents the way in which a modern system for nuclear safety at the TRIGA reactor has been implemented and developed, in accordance to specific criteria and requirements imposed by related National Regulations and with the principles of safety culture. Starting from the definition of specific responsabilities, there are presented the internal stipulations and practical actions at all levels in order to enhance nuclear safety. (orig.)

  6. Atoms at work

    International Nuclear Information System (INIS)

    1982-07-01

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

  7. Nuclear reactor for release of nuclear energy, without a chain reaction using the simultaneous implosion of three, or more, atomic nuclei

    International Nuclear Information System (INIS)

    Pedrick, A.P.

    1976-01-01

    A modified form of what is known as a 'streaking nuclear reactor' is described. In this type of reactor it is proposed to obtain release of nuclear energy from atomic nuclei by stripping such nuclei of their electron clouds or shells, to form a high temperature plasma, and breaking nucleons off the surface of the nuclei. In the apparatus described it is proposed to break up nuclei by causing three or more nuclei to collide with each other at very high velocity. Streams of nuclei, stripped of their electron clouds are directed into a reactor vessel to a focal point or implosion center along three or more ducts, equi-angularly spaced around the implosion center in the same plane, the arrangement being such as to permit mutual simultaneous collision of three or more of the nuclei. The importance of achieving a release of nuclear energy in this manner is that it may be able to use any chemical element that can be converted to a plasma, but it is most likely to be successful with elements of high atomic number, such as Pb or Bi. (U.K.)

  8. TRIGA reactor main systems

    International Nuclear Information System (INIS)

    Boeck, H.; Villa, M.

    2007-01-01

    This module describes the main systems of low power (<2 MW) and higher power (≥2 MW) TRIGA reactors. The most significant difference between the two is that forced reactor cooling and an emergency core cooling system are generally required for the higher power TRIGA reactors. However, those TRIGA reactors that are designed to be operated above 3 MW also use a TRIGA fuel that is specifically designed for those higher power outputs (3 to 14 MW). Typical values are given for the respective systems although each TRIGA facility will have unique characteristics that may only be determined by the experienced facility operators. Due to the inherent wide scope of these research reactor facilities construction and missions, this training module covers those systems found at most operating TRIGA reactor facilities but may also discuss non-standard equipment that was found to be operationally useful although not necessarily required. (author)

  9. UTC(OP) based on LNE-SYRTE atomic fountain primary frequency standards

    International Nuclear Information System (INIS)

    Rovera, G.D.; Bize, S.; Chupin, B.; Guena, J.; Laurent, Ph; Rosenbusch, Ph.; Uhrich, P.; Abgrall, M.

    2016-01-01

    UTC(OP), the French national realization of the international coordinated universal time, was redesigned and rebuilt. The first step was the implementation in October 2012 of a new algorithm based on a H-maser and on atomic fountain data. Thanks to the new implementation, the stability of UTC(OP) was dramatically improved and UTC(OP) competes with the best time scales available today. Then the hardware generation and distribution of the UTC(OP) physical signals were replaced. Part of the new hardware is composed of commercial devices, but the key elements were specifically developed. One of them is a special switch that allows the UTC(OP) signals to be derived from one of two time scales, based on two different H-masers, which are generated simultaneously. This insures the continuity of the UTC(OP) signal even when a change of the reference H-maser is required. With the new hardware implementation, UTC(OP) is made available through three coherent signals: 100 MHz, 10 MHz and 1 PPS. For more than 3 years, UTC(OP) remained well below 10 ns close to UTC, with a difference even less than 5 ns if we except a short period around MJD 56650. (authors)

  10. Standard Guide for Conducting Supplemental Surveillance Tests for Nuclear Power Reactor Vessels, E 706 (IH)

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This guide discusses test procedures that can be used in conjunction with, but not as alternatives to, those required by Practices E185 and E2215 for the surveillance of nuclear reactor vessels. The supplemental mechanical property tests outlined permit the acquisition of additional information on radiation-induced changes in fracture toughness, notch ductility, and yield strength properties of the reactor vessel steels. 1.2 This guide provides recommendations for the preparation of test specimens for irradiation, and identifies special precautions and requirements for reactor surveillance operations and postirradiation test planning. Guidance on data reduction and computational procedures is also given. Reference is made to other ASTM test methods for the physical conduct of specimen tests and for raw data acquisition.

  11. The law for the Japan Atomic Energy Research Institute

    International Nuclear Information System (INIS)

    1985-01-01

    The Act for Japan Atomic Energy Research Institute has been promulgated anew. Contents are the following : general rules, officials, advisors and personnel, duties, financial affairs and accounts, supervision, miscellaneous rules, penal provisions, and additional rules. (In the additional rules, the merger into JAERI of Japan Nuclear Ship Research and Development Agency is treated.) Japan Atomic Energy Research Institute conducts research etc. for the development of atomic energy comprehensively and efficiently, thereby contributing to the promotion of atomic energy research, development and utilization, according to the Atomic Energy Fundamental Act. Duties are atomic energy basic and application research, reactor relation, training of the personnel, RIs relation, etc. (Mori, K.)

  12. ["Epistemic Negotiations" and the Pluralism of the Radiation Protection Regime: The Determination of Radiation Protection Standards for the General Population in the Early Years After World War II].

    Science.gov (United States)

    Higuchi, Toshihiro

    2015-10-01

    Radiation protection standards for the general population have constituted one of the most controversial subjects in the history of atomic energy uses. This paper reexamines the process in which the first such standards evolved in the early postwar period. While the existing literature has emphasized a "collusion" between the standard-setters and users, the paper seeks to examine the horizontal relationship among the standard-setters. It first examines a series of expert consultations between the United States and the United Kingdom. Representing a different configuration of power and interest, the two failed to agree on the assessment of genetic damage and cancer induction whose occurrence might have no threshold and therefore be dependent on the population size. This stalemate prevented the International Commission on Radiological Protection (ICRP), established in 1950, from formulating separate guidelines for the general public. Situations radically changed when the Bikini incident in 1954 led to the creation of more scientific panels. One such panel under the U.S. Academy of Sciences enabled the geneticists to bridge their internal divide, unanimously naming 100 mSv as the genetically permissible dose for the general population. Not to be outdone, ICRP publicized its own guidelines for the same purpose. The case examined in this paper shows that the standard-setting process is best understood as a series of "epistemic negotiations" among and within the standard-setters, whose agendas were determined from the outset but whose outcomes were not.

  13. Statement of the Director General to the forty-third regular session of the General Conference of the International Atomic Energy Agency

    International Nuclear Information System (INIS)

    ElBaradei, M.

    1999-01-01

    In his Statement on the forty-third regular session of the General Conference of the IAEA, the Director General of the Agency presented the past, present and future programme of the Agency in terms of the three 'pillars' that constitute its mandate: Nuclear technology (with emphasis on nuclear power and nuclear applications); Nuclear, radiation and waste safety (with emphasis on international conventions prescribing the basic legal norms for the safe use of nuclear energy, internationally accepted safety standards, and measures to assist Member States in the implementation of these conventions and standards, including technological solutions for improving safety); Nuclear verification and the security of material (with emphasis on IAEA safeguards, and physical protection of nuclear material), including possible new verification activities and specific verification issues (Iraq, DPRK, Middle East region). The Director General also touched in his Statement on two supporting elements that undergird these 'pillars': effective interaction with partners and the public, and achieving excellence in management

  14. Lattice-Assisted Spectroscopy: A Generalized Scanning Tunneling Microscope for Ultracold Atoms.

    Science.gov (United States)

    Kantian, A; Schollwöck, U; Giamarchi, T

    2015-10-16

    We propose a scheme to measure the frequency-resolved local particle and hole spectra of any optical lattice-confined system of correlated ultracold atoms that offers single-site addressing and imaging, which is now an experimental reality. Combining perturbation theory and time-dependent density matrix renormalization group simulations, we quantitatively test and validate this approach of lattice-assisted spectroscopy on several one-dimensional example systems, such as the superfluid and Mott insulator, with and without a parabolic trap, and finally on edge states of the bosonic Su-Schrieffer-Heeger model. We highlight extensions of our basic scheme to obtain an even wider variety of interesting and important frequency resolved spectra.

  15. 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)

  16. The rehabilitation/upgrading of Philippine Research Reactor

    International Nuclear Information System (INIS)

    Renato T, Banaga

    1998-01-01

    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 1 -U-Z 1 -H 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)

  17. Design and safety considerations for the 10 MW(t) multipurpose TRIGA reactor in Thailand

    International Nuclear Information System (INIS)

    Razvi, J.; Bolin, J.M.; Saurwein, J.J.; Whittemore, W.L.; Proongmuang, S.

    1999-01-01

    General Atomics (GA) is constructing the Ongkharak Nuclear Research Center (ONRC) near Bangkok, Thailand for the Office of Atomic Energy for Peace. The ONRC complex includes the following: A multipurpose 10 MW(t) research reactor; An Isotope Production Facility; Centralized Radioactive Waste Processing and Storage Facilities. The Center is being built 60-km northeast of Bangkok, with a 10 MW(t) TRIGA type research reactor as the centerpiece. Facilities are included for neutron transmutation doping of silicon, neutron capture therapy neutron beam research and for production of a variety of radioisotopes. The facility will also be utilized for applied research and technology development as well as training in reactor operations, conduct of experiments and in reactor physics. The multipurpose, pool-type reactor will be fueled with high-density (45 wt%), low-enriched (19.7 wt%) uranium-erbium-zirconium-hydride (UErZrH) fuel rods, cooled and moderated by light water, and reflected by beryllium and heavy water. The general arrangement of the reactor and auxiliary pool structure allows irradiated targets to be transferred entirely under water from their irradiation locations to the hot cell, then pneumatically transferred to the adjacent Isotope Production Facility for processing. The core configuration includes 4 x 4 array standard TRIGA fuel clusters, modified clusters to serve as fast-neutron irradiation facilities, control rods and an in-core Ir-192 production facility. The active core is reflected on two sides by beryllium and on the other two sides by D 2 O. Additional irradiation facilities are also located in the beryllium reflector blocks and the D 2 O reflector blanket. The fuel provides the fundamental safety feature of the ONRC reactor, and as a result of all the well established accident-mitigating characteristics of the UErZrH fuel itself (large prompt negative temperature coefficient of reactivity, fission product retention and chemical stability), a

  18. Nanostructural evolution in surveillance test specimens of a commercial nuclear reactor pressure vessel studied by three-dimensional atom probe and positron annihilation

    International Nuclear Information System (INIS)

    Toyama, T.; Nagai, Y.; Tang, Z.; Hasegawa, M.; Almazouzi, A.; Walle, E. van; Gerard, R.

    2007-01-01

    The nanostructural evolution of irradiation-induced Cu-rich nanoprecipitates (CRNPs) and vacancy clusters in surveillance test specimens of in-service commercial nuclear reactor pressure vessel steel welds of Doel-1 and Doel-2 are revealed by combining the three-dimensional local electrode atom probe and positron annihilation techniques. In both medium (0.13 wt.%) and high (0.30 wt.%) Cu welds, the CRNPs are found to form readily at the very beginning of the reactor lifetime. Thereafter, during the subsequent 30 years of operation, the residual Cu concentration in the matrix shows a slight decrease while the CRNPs coarsen. On the other hand, small vacancy clusters of V 3 -V 4 start appearing after the initial Cu precipitation and accumulate steadily with increasing neutron dose. The observed nanostructural evolution is shown to provide unique and fundamental information about the mechanisms of the irradiation-induced embrittlement of these specific materials

  19. After-operating properties of nuclear reactor vessel materials of Lenin atomic ice breaker and prospective of reactor vessels radiation life prolongation

    International Nuclear Information System (INIS)

    Platonov, P.A.; Shtrombakh, Ya.I.; Amaev, A.D.; Krasikov, E.A.; Korolev, Yu.N.; Zabusov, O.O.; Glushakov, G.M.

    2001-01-01

    A post-operational state of the icebreaker Lenin reactor vessel metal is investigated. It is shown that a base metal of the icebreaker Lenin reactor vessel is of high quality as by an initial value of critical temperature of embrittlement, so by its radiation resistance. The weld metal possesses a sufficient radiation resistance but has an insufficient initial ductile-brittle transition temperature (approximately 63 Deg C). It is necessary to note that the final stage of operation for nuclear steam-generating plant should be carried out at the coolant temperature as high as possible [ru

  20. Proposals for new standardized general diagnostic criteria for the secondary headaches

    DEFF Research Database (Denmark)

    Olesen, J; Steiner, T; Bousser, M-G

    2009-01-01

    headache and chronic migraine. These changes made apparent a further need for broader revisions to the standard formulation of diagnostic criteria for the secondary headaches. Currently, the fourth criterion makes impossible the definitive diagnosis of a secondary headache until the underlying cause has...... propose maintaining a standard approach to the secondary headaches using a set of four criteria A, B, C and D, but we construct these so that the requirement for resolution or successful treatment is removed. The proposal for general diagnostic criteria for the secondary headaches will be entered...... classification are then foreseen for the next 10 years. Until the printing of ICHD-IIR, the printed ICHD-II criteria remain in place for all other purposes. We issue a plea to the headache community to use and study these proposed general criteria for the secondary headaches in order to provide more evidence...

  1. General outline of the operation and utilization of the BR2 reactor

    International Nuclear Information System (INIS)

    Baugnet, J.M.; Leonard, F.; Gandolfo, J.M.; Lenders, H.

    1978-01-01

    The BR2 reactor is a high-flux material testing reactor of the thermal heterogeneous type. The fuel is 93% 235 U enriched uranium in the form of plates clad in aluminium. The moderator consists of beryllium and light water, the water being pressurized (12.5kg/cm 2 )and acting also as coolant. The pressure vessel is of aluminium, and is placed in a pool of demineralized water. One should stress the following main features of the design: the experimental channels are skew, the tube bundle presenting the form of a hyperboloid of revolution (see figure 1)-this gives easy access at the top and bottom reactor covers allowing complex instrumented devices, while maintaining a very high neutron flux at the core; great flexibilty of utilization, due to the fact that it is possible to adapt the core configuration to the experimental loading as the fissile charge can be centred on different experimental channels; although BR2 is a thermal reactor, it is possible to achieve neutron spectra very similar to those obtained in a fast reactor, either by the use of absorbing screens or by the use of fissile material within the experimental device; five 200mm diameter channels are available for loading large experimental irradiation devices, as in-pile sodium, gas or water loops. (author)

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

    International Nuclear Information System (INIS)

    2007-01-01

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

  3. The study of the irradiation-induced embrittlement of reactor pressure vessels. Analysis of surveillance test specimens of a commercial nuclear reactor pressure vessel studied by three-dimensional atom probe and positron annihilation

    International Nuclear Information System (INIS)

    Nagai, Yasuyoshi; Toyama, Takeshi; Hasegawa, Masayuki

    2007-01-01

    The study of embrittlement of nuclear power reactor pressure vessels (RPVs) is of critical importance for the safety assessment in the nuclear industry. Some origins of embrittlement are attributed to fine Cu precipitates, matrix defects, grain boundary segregation of P and late blooming phase. This review article described nanostructural observation by three-dimensional atom probe (3DAP) and positron annihilation spectroscopy (PAS). The density and sizes of Cu-rich nanoprecipitates and grain boundary segregation are sensitively detected by 3DAP, and vacancies are probed by PAS. Element analysis around vacancies and fine microstructural Cu precipitates not containing vacancies are successfully observed by a coincidence doppler broadening method. The nanostructural evolution of irradiation-induced Cu-rich nanoprecipitates (CRNPs) and vacancy clusters in surveillance test specimens of commercial nuclear reactor pressure vessel steel welds of Doel-2 in Belgium were revealed by combining 3DAP and PAS. In both medium (0.13 wt%) and high (0.30 wt%) Cu welds, the CRNPs were found to form readily at the very beginning of the reactor lifetime. On the other hand, small vacancy clusters start appearing after the initial Cu precipitates and accumulate steadily with increasing neutron dose. The CRNPs were also observed at very low dose rate of neutrons in the test specimen of Calder Hall Reactor of Japan Atomic Power Company. The significant enhancement of these Cu precipitates results in the embrittlement in practical RPVs. At very high dose of 2.2x10 18 n/cm 2 by JMTR, the Cu precipitates were scarcely observed, and the irradiation-induced embrittlement was primarily caused from vacancy-impurity complexes and dislocation loops. (author)

  4. Review of footnotes and annotations to the 1949–2013 tables of standard atomic weights and tables of isotopic compositions of the elements (IUPAC Technical Report)

    Science.gov (United States)

    Coplen, Tyler B.; Holden, Norman E.

    2016-01-01

    The Commission on Isotopic Abundances and Atomic Weights uses annotations given in footnotes that are an integral part of the Tables of Standard Atomic Weights to alert users to the possibilities of quite extraordinary occurrences, as well as sources with abnormal atomic-weight values outside an otherwise acceptable range. The basic need for footnotes to the Standard Atomic Weights Table and equivalent annotations to the Table of Isotopic Compositions of the Elements arises from the necessity to provide users with information that is relevant to one or more elements, but that cannot be provided using numerical data in columns. Any desire to increase additional information conveyed by annotations to these Tables is tempered by the need to preserve a compact format and a style that can alert users, who would not be inclined to consult either the last full element-by-element review or the full text of a current Standard Atomic Weights of the Elements report. Since 1989, the footnotes of the Tables of Standard Atomic Weights and the annotations in column 5 of the Table of Isotopic Compositions of the Elements have been harmonized by use of three lowercase footnotes, “g”, “m”, and “r”, that signify geologically exceptionally specimens (“g”), modified isotopic compositions in material subjected to undisclosed or inadvertent isotopic fractionation (“m”), and the range in isotopic composition of normal terrestrial material prevents more precise atomic-weight value being given (“r”). As some elements are assigned intervals for their standard atomic-weight values (applies to 12 elements since 2009), footnotes “g” and “r” are no longer needed for these elements.

  5. Standard Technical Specifications General Electric plants, BWR/4:Bases (Sections 3.4-3.10). Volume 3, Revision 1

    International Nuclear Information System (INIS)

    1995-04-01

    This report documents the results of the combined effort of the NRC and the industry to produce improved Standard Technical Specifications (STS), Revision 1 for General Electric BWR/4 Plants. The changes reflected in Revision 1 resulted from the experience gained from license amendment applications to convert to these improved STS or to adopt partial improvements to existing technical specifications. This NUREG is the result of extensive public technical meetings and discussions between the Nuclear Regulatory Commission (NRC) staff and various nuclear power plant licensees, Nuclear Steam Supply System (NSSS) Owners Groups, NSSS vendors, and the Nuclear Energy Institute (NEI). The improved STS were developed based on the criteria in the Final Commission Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, dated July 22, 1993. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume 1 contains the specifications for all chapters and sections of the improved STS. Volume 2 contains he Bases for Chapters 2.0 and 3.0, and Sections 3.1-3.3 of the improved STS. This document, Volume 3, contains the Bases for Sections 3.4-3.10 of the improved STS

  6. Standard Technical Specifications General Electric plants, BWR/4: Bases (Sections 2.0-3.3). Volume 2, Revision 1

    International Nuclear Information System (INIS)

    1995-04-01

    This report documents the results of the combined effort of the NRC and the industry produce improved Standard Technical Specifications (STS), Revision 1 for General Electric BWR/4 Plants. The changes reflected in Revision 1 resulted from the experience gained from license amendment applications to convert to these improved ST or to adopt partial improvements to existing technical specifications. This NUREG is the result of extensive public technical meetings and discussions between the Nuclear Regulatory Commission (NRC) staff and various nuclear power plant licensees, Nuclear Steam Supply System (NSSS) Owners Groups, NSSS vendors, and the Nuclear Energy Institute (NEI). The improved STS were developed based on the criteria in the Final Commission Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, dated July 22, 1993. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume I contains the Specifications for all chapters and sections of the improved STS. This document, Volume 2, contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1-3.3 of the improved STS. Volume 3 contains the Bases for Sections 3.4-3.10 of the improved STS

  7. Safety-evaluation report related to the license renewal and power increase for the National Bureau of Standards Reactor (Docket No. 50-184)

    International Nuclear Information System (INIS)

    1983-09-01

    This Safety Evaluation Report for the application filed by the National Bureau of Standards (NBS) for an increase in power from 10 MWt to 20 MWt and for a renewal of the Operating License TR-5 to continue to operate the test reactor has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located in Gaithersburg, Maryland, on the site of the National Bureau of Standards, which is a bureau of the Department of Commerce. The staff concludes that the NBS reactor can operate at the 20 MWt power level without endangering the health and safety of the public

  8. Generalized position-momentum uncertainty products: Inclusion of moments with negative order and application to atoms

    International Nuclear Information System (INIS)

    Angulo, J. C.

    2011-01-01

    Rigorous and universal relationships among radial expectation values of any D-dimensional quantum-mechanical system are obtained, using Renyi-like position-momentum inequalities in an information-theoretical framework. Although the results are expressed in terms of four moments (two in position space and two in the momentum one), especially interesting are the cases that provide expressions of uncertainty in terms of products a > 1/a b > 1/b , widely considered in the literature, including the famous Heisenberg relationship 2 > 2 >≥D 2 /4. Improved bounds for these products have recently been provided, but are always restricted to positive orders a,b>0. The interesting part of this work are the inequalities for negative orders. A study of these relationships is carried out for atomic systems in their ground state. Some results are given in terms of relevant physical quantities, including the kinetic and electron-nucleus attraction energies, the diamagnetic susceptibility, and the height of the peak of the Compton profile, among others.

  9. Cable addressed to the Director General by Mr. R. Amrollahi, Deputy Prime Minister of the Islamic Republic of Iran and President of the Atomic Energy Organization of Iran

    International Nuclear Information System (INIS)

    1987-02-01

    The full text of the cable of 9 February 1987 addressed by the President of the Atomic Energy Organization of Iran to the Director General of the IAEA and referring to military attacks on Bushehr Nuclear Power Plant is reproduced

  10. The achievement impact of the inclusion model on the standardized test scores of general education students

    Science.gov (United States)

    Garrett-Rainey, Syrena

    The purpose of this study was to compare the achievement of general education students within regular education classes to the achievement of general education students in inclusion/co-teach classes to determine whether there was a significant difference in the achievement between the two groups. The school district's inclusion/co-teach model included ongoing professional development support for teachers and administrators. General education teachers, special education teachers, and teacher assistants collaborated to develop instructional strategies to provide additional remediation to help students to acquire the skills needed to master course content. This quantitative study reviewed the end-of course test (EoCT) scores of Grade 10 physical science and math students within an urban school district. It is not known whether general education students in an inclusive/co-teach science or math course will demonstrate a higher achievement on the EoCT in math or science than students not in an inclusive/co-teach classroom setting. In addition, this study sought to determine if students classified as low socioeconomic status benefited from participating in co-teaching classrooms as evidenced by standardized tests. Inferential statistics were used to determine whether there was a significant difference between the achievements of the treatment group (inclusion/co-teach) and the control group (non-inclusion/co-teach). The findings can be used to provide school districts with optional instructional strategies to implement in the diverse classroom setting in the modern classroom to increase academic performance on state standardized tests.

  11. A general overview of generation IV molten salt reactor (MSR) and the use of thorium as fuel

    Energy Technology Data Exchange (ETDEWEB)

    Yamaguchi, Carlos H.; Stefani, Giovanni L.; Santos, Thiago A., E-mail: carlos.yamaguchi@usp.br, E-mail: giovanni.stefani@ipen.br, E-mail: thiago.santos@ufabc.edu.br [Universidade de Sao Paulo (USP), SP (Brazil). Instituto de Fisica; Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Universidade Federal do ABC (CECS/UFABC), Santo Andre, SP (Brazil). Centro de Engenharia, Modelagem e Ciencias Sociais Aplicadas

    2017-07-01

    The molten salt reactors (MSRs) make use of fluoride salt as primary cooler, at low pressure. Although considered a generation IV reactor, your concept isn't new, since in the 1960 years the Oak Ridge National Laboratory created a little prototype of 8MWt. Over the 20{sup th} century, other countries, like UK, Japan, Russia, China and France also did research in the area, especially with the use of thorium as fuel. This goes with the fact that Brazil possess the biggest reserve of thorium in the world. In the center of nuclear engineering at IPEN is being created a study group connected to thorium reactors, which purpose is to investigate reactors using thorium to produce {sup 233}U and tailing burn, thus making the MSR using thorium as fuel, an object of study. This present work searches to do a general summary about the researches of MSR's, having as focus the utilization of thorium with the goal being to show it's efficiency and utilization is doable. (author)

  12. Axial stability of VVER-1000 reactor with control with minimum standard deviation

    International Nuclear Information System (INIS)

    Afanas'ev, A.M.; Torlin, B.Z.

    1980-01-01

    Results are given of investigations on the stability of a reactor which has, in addition to an automatic controller, a height distribution regulator (HDR) based on an auxiliary control rod (CR) or a special shortened absorption rod (SAR). The HDR was controlled by using either a special ionization chamber (IC), generating an imbalance signal which sets the CR in motion, or two ionization chambers whose difference signal causes a displacement of the SAR. Since data from numerous pickups can be used to control the height field of the VVER-1000, it is of interest to analyze how this would affect the stability of the reactor. The analysis was carried out with the improved IRINA programs. 11 refs

  13. The French Fast Reactor Program - Innovations in Support to Higher Standards

    International Nuclear Information System (INIS)

    Gauché, François

    2013-01-01

    • From the experience of ASTRID first phase of conceptual design studies (2010-2012), two remarks can be made: → Higher requirements in safety and operability lead to higher costs that cannot be fully recovered by advances in technology. This puts additional pressure on the next phases of the design to optimize the design and to keep the costs to the minimum. → There is a clear link between the level of safety that can be achieved and the maturity of the technology, i.e. the experience accumulated in R&D, design, construction, operation and decommissioning of past reactors. In the field of fast neutron reactors, this gives a strong advantage to the sodium technology, because strengths and weaknesses are well mastered. • Meeting the high requirements set for ASTRID and serving R&D needs of innovative options will require increased industrial and international collaboration

  14. Standardized dose factors for dose calculations - 1982 SRP reactor safety analysis report tritium, iodine, and noble gases

    International Nuclear Information System (INIS)

    Pillinger, W.L.; Marter, W.L.

    1982-01-01

    Standardized dose constants are recommended for calculation of offsite doses in the 1982 SRP Reactor Safety Analysis Report (SAR). Dose constants are proposed for inhalation of tritium and radioiodines and for submersion in a semi-infinite cloud of radioiodines and noble gases. The proposed constants, based on ICRP2 methodology for internal dose and methodology recommended by the US Nuclear Regulatory Commission for external dose, are compatible with dose calculational methods used at the Savannah River Plant and Savannah River Laboratory for normal releases of radioactivity. 8 references

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

    International Nuclear Information System (INIS)

    2001-10-01

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

  16. Safety Evaluation Report related to the renewal of the operating license for the General Electric-Nuclear Test Reactor (GE-NTR) (Docket No. 50-73)

    International Nuclear Information System (INIS)

    1984-09-01

    This Safety Evaluation Report for the application filed by the General Electric Company (GE) for a renewal license number R-33 to continue to operate its research reactor has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is owned and operated by GE and is located in Pleasanton, California. The staff concludes that the reactor can continue to be operated by GE without endangering the health and safety of the public

  17. Analyses and results from standard surveillance programmes of WWER 440/V-213C reactor pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Falcnik, M; Brumovsky, M; Pav, T [Czech Nuclear Society, Prague (Czech Republic)

    1994-12-31

    In Czech and Slovak republics, six units of WWER 440/C type reactors are monitored by surveillance specimens programmes; the specimens are determined for static tensile testing, impact notch toughness testing and fracture toughness evaluation. Results of mechanical properties of these specimens after irradiation in intervals between 1 and 5 years of operation, are summarized and discussed with respect to the effect of individual heats and welded joints, radiation embrittlement, and annealing recovery. (authors). 3 refs., 11 figs., 2 tabs.

  18. Reactor technology

    International Nuclear Information System (INIS)

    Erdoes, P.

    1977-01-01

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

  19. CAC-RA1 1958-1998. The first years of the Constituyentes Atomic Center (CAC). History of the first Argentine nuclear reactor (RA-1); CAC-RA-1 1958-1998. Los primeros anios del CAC. Historia del primer reactor nuclear argentino (RA-1)

    Energy Technology Data Exchange (ETDEWEB)

    Forlerer, Elena; Palacios, Tulio A [comps.

    1998-07-01

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

  20. General empirical model for 60Co generation in pressurized water reactors with continuous refueling

    International Nuclear Information System (INIS)

    Urrutia, G.A.; Blesa, M.A.; Fernandez-Prini, R.; Maroto, A.J.G.

    1984-01-01

    A simplified model is presented that permits one to calculate the average activity on the fuel elements of a reactor that operates under continuous refueling, based on the assumption of crud interchange between fuel element surface and coolant in the form of particulate material only and using the crud specific activity as an empirical parameter determined in plant. The net activity flux from core to out-of-core components is then calculated in the form of parametric curves depending on crud specific activity and rate of particulate release from fuel surface. In pressure vessel reactors, contribution to out-ofcore radionuclide inventory arising in the release of activated materials from core components must be taken into account. The contribution from in situ activation of core components is calculated from the rates of release and the specific activities corresponding to the exposed surface of the component (calculated in a straightforward way on the basis of core geometry and neutron fluxes). The rates of release can be taken from the literature, or in the case of cobalt-rich alloys, can be calculated from experimentally determined cobalt contents of structural components and crud. For pressure vessel reactors operating under continuous refueling, activation of deposited crud and release of activated materials are compared; the latter, in certain cases, may represent a sizable (and even the largest) fraction of the total cobalt activity. It is proposed that the ratio of activities of 59 Fe to 54 Mn may be used as a diagnostic tool for in situ activation of structural materials; available data indicate ratios close to unity for pressure tube heavy water reactors (no in situ activation) and ratios around 4 to 10 for pressure vessel, heavy water reactors

  1. Reactor instrumentation renewal of the TRIGA reactor Vienna, Austria

    International Nuclear Information System (INIS)

    Boeck, H.; Weiss, H.; Hood, W.E.; Hyde, W.K.

    1992-01-01

    The TRIGA Mark-II reactor at the Atominstitut in Vienna, Austria is replacing its twenty-four year old instrumentation system with a microprocessor based control system supplied by General Atomics. Ageing components, new governmental safety requirements and a need for state of the art instrumentation for training students has spurred the demand for new reactor instrumentation. In Austria a government appointed expert is assigned the responsibility of reviewing the proposed installation and verifying all safety aspects. After a positive review, final assembly and checkout of the instrumentation system may commence. The instrumentation system consists of three basic modules: the control system console, the data acquisition console and the NH-1000 wide range channel. Digital communications greatly reduce interwiring requirements. Hardwired safety channels are independent of computer control, thus, the instrumentation system in no way relies on any computer intervention for safety function. In addition, both the CSC and DAC computers are continuously monitored for proper operation via watchdog circuits which are capable of shutting down the reactor in the event of computer malfunction. Safety channels include two interlocked NMP-1000 multi-range linear channels for steady state mode, an NPP-1000 linear safety channel for pulse mode and a set of three independent fuel temperature monitoring channels. The microprocessor controlled wide range NM- 1000 digital neutron monitor (fission chamber based) functions as a startup/operational channel, and provides all power level related Interlocks. The Atominstitut TRIGA reactor is configured for four modes of operation: manual mode, automatic mode (servo control), pulsing mode and square wave mode. Control of the standard control rods is via stepping motor control rod drives, which offers the operator the choice of which control rods are operated by the servo system in automatic and square wave model. (author)

  2. From Classical to Discrete Gravity through Exponential Non-Standard Lagrangians in General Relativity

    Directory of Open Access Journals (Sweden)

    Rami Ahmad El-Nabulsi

    2015-08-01

    Full Text Available Recently, non-standard Lagrangians have gained a growing importance in theoretical physics and in the theory of non-linear differential equations. However, their formulations and implications in general relativity are still in their infancies despite some advances in contemporary cosmology. The main aim of this paper is to fill the gap. Though non-standard Lagrangians may be defined by a multitude form, in this paper, we considered the exponential type. One basic feature of exponential non-standard Lagrangians concerns the modified Euler-Lagrange equation obtained from the standard variational analysis. Accordingly, when applied to spacetime geometries, one unsurprisingly expects modified geodesic equations. However, when taking into account the time-like paths parameterization constraint, remarkably, it was observed that mutually discrete gravity and discrete spacetime emerge in the theory. Two different independent cases were obtained: A geometrical manifold with new spacetime coordinates augmented by a metric signature change and a geometrical manifold characterized by a discretized spacetime metric. Both cases give raise to Einstein’s field equations yet the gravity is discretized and originated from “spacetime discreteness”. A number of mathematical and physical implications of these results were discussed though this paper and perspectives are given accordingly.

  3. Statement to the 34th session of the general conference of the International Atomic Energy Agency, 17 September 1990. Statement to the 45th session of the United Nations general assembly, 23 October 1990

    International Nuclear Information System (INIS)

    Blix, H.

    1991-01-01

    The document contains the following two statements of Hans Blix, Director General of the IAEA: Statement to the 34th Session of the General Conference of the International Atomic Energy Agency, 17 September 1990; Statement to the 45th Session of the United Nations General Assembly, 23 October 1990. A separate abstract was prepared for each of these statements

  4. Statement to the 35th session of the General Conference of the International Atomic Energy Agency 16 September 1991; Statement to the 46th session of the United Nations General Assembly 21 October 1991

    International Nuclear Information System (INIS)

    Blix, H.

    1991-01-01

    The document contains the following two statements of Hans Blix, Director General of the IAEA: Statement to the 35th session of the General Conference of the International Atomic Energy Agency, 16 September 1991; Statement to the 46th session of the United Nations General Assembly, 21 October 1991. A separate abstract was prepared for each of these statements

  5. Survey of legal aspects, regulations, standards and guidelines applicable to radioactive waste management of the Brazilian Multipurpose Reactor - RMB

    International Nuclear Information System (INIS)

    Salvetti, T.C.; Marumo, J.T.

    2017-01-01

    In Brazil, the Brazilian Nuclear Energy Commission (CNEN) and Brazilian Institute of Environment and Renewable Natural Resources (IBAMA) are the agencies responsible for the execution, regulation and control of nuclear and environmental policies, respectively. Such regulatory activities are very comprehensive (IBAMA) or too specific (CNEN), revealing other aspects that would, also, need to be observed so that the management could be carried out efficiently (quality) and effectively (safety), including the three governmental administrative levels: Federal, State and Municipal. In addition to laws, regulations, decrees and resolutions, there are also national and international standards and guides that provide guidelines for structuring the current management and the use of best regulatory practices. The Brazilian Multipurpose Reactor Enterprise (RMB) is a CNEN project, complying with a Multi-Year Plan of the Brazilian Ministry of Planning, Development and Management (MPDG). The Enterprise is being developed under the responsibility of the Directorate of Research and Development - DPD of CNEN and will have a facility for treatment and initial temporary storage of the radioactive waste generated by the operation of the research reactor and the activities carried out in the associated laboratories. The RMB will be built in the city of IPERÓ, located in the state of São Paulo, near ARAMAR Experimental Center of the Brazilian Navy. This work aims to present the research results regarding the various aspects that regulate, legislate and standardize the practices proposed to the Radioactive Waste Management of the RMB project. (author)

  6. Survey of legal aspects, regulations, standards and guidelines applicable to radioactive waste management of the Brazilian Multipurpose Reactor - RMB

    Energy Technology Data Exchange (ETDEWEB)

    Salvetti, T.C.; Marumo, J.T., E-mail: salvetti@ipen.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil)

    2017-07-01

    In Brazil, the Brazilian Nuclear Energy Commission (CNEN) and Brazilian Institute of Environment and Renewable Natural Resources (IBAMA) are the agencies responsible for the execution, regulation and control of nuclear and environmental policies, respectively. Such regulatory activities are very comprehensive (IBAMA) or too specific (CNEN), revealing other aspects that would, also, need to be observed so that the management could be carried out efficiently (quality) and effectively (safety), including the three governmental administrative levels: Federal, State and Municipal. In addition to laws, regulations, decrees and resolutions, there are also national and international standards and guides that provide guidelines for structuring the current management and the use of best regulatory practices. The Brazilian Multipurpose Reactor Enterprise (RMB) is a CNEN project, complying with a Multi-Year Plan of the Brazilian Ministry of Planning, Development and Management (MPDG). The Enterprise is being developed under the responsibility of the Directorate of Research and Development - DPD of CNEN and will have a facility for treatment and initial temporary storage of the radioactive waste generated by the operation of the research reactor and the activities carried out in the associated laboratories. The RMB will be built in the city of IPERÓ, located in the state of São Paulo, near ARAMAR Experimental Center of the Brazilian Navy. This work aims to present the research results regarding the various aspects that regulate, legislate and standardize the practices proposed to the Radioactive Waste Management of the RMB project. (author)

  7. Application of wire electrodes in electric discharge machining of metal samples of reactor blocks of the operative atomic power station

    International Nuclear Information System (INIS)

    Gozhenko, S.V.

    2007-01-01

    Features of application of electroerosive methods are considered during the process of direct definition of properties of metal of the equipment of power units of the atomic power station. Results of development of a complex of the equipment for wire electric discharge machining of metal templet and its use are presented at the control of the basic metal of the main circulating pipelines over blocks of the atomic power station of Ukraine over long terms of operation

  8. Beyond the Floquet theorem: generalized Floquet formalisms and quasienergy methods for atomic and molecular multiphoton processes in intense laser fields

    Science.gov (United States)

    Chu, Shih-I.; Telnov, Dmitry A.

    2004-02-01

    The advancement of high-power and short-pulse laser technology in the past two decades has generated considerable interest in the study of multiphoton and very high-order nonlinear optical processes of atomic and molecular systems in intense and superintense laser fields, leading to the discovery of a host of novel strong-field phenomena which cannot be understood by the conventional perturbation theory. The Floquet theorem and the time-independent Floquet Hamiltonian method are powerful theoretical framework for the study of bound-bound multiphoton transitions driven by periodically time-dependent fields. However, there are a number of significant strong-field processes cannot be directly treated by the conventional Floquet methods. In this review article, we discuss several recent developments of generalized Floquet theorems, formalisms, and quasienergy methods, beyond the conventional Floquet theorem, for accurate nonperturbative treatment of a broad range of strong-field atomic and molecular processes and phenomena of current interests. Topics covered include (a) artificial intelligence (AI)-most-probable-path approach (MPPA) for effective treatment of ultralarge Floquet matrix problem; (b) non-Hermitian Floquet formalisms and complex quasienergy methods for nonperturbative treatment of bound-free and free-free processes such as multiphoton ionization (MPI) and above-threshold ionization (ATI) of atoms and molecules, multiphoton dissociation (MPD) and above-threshold dissociation (ATD) of molecules, chemical bond softening and hardening, charge-resonance enhanced ionization (CREI) of molecular ions, and multiple high-order harmonic generation (HHG), etc.; (c) many-mode Floquet theorem (MMFT) for exact treatment of multiphoton processes in multi-color laser fields with nonperiodic time-dependent Hamiltonian; (d) Floquet-Liouville supermatrix (FLSM) formalism for exact nonperturbative treatment of time-dependent Liouville equation (allowing for relaxations and

  9. Beyond the Floquet theorem: generalized Floquet formalisms and quasienergy methods for atomic and molecular multiphoton processes in intense laser fields

    International Nuclear Information System (INIS)

    Chu, S.-I.; Telnov, D.A.

    2004-01-01

    The advancement of high-power and short-pulse laser technology in the past two decades has generated considerable interest in the study of multiphoton and very high-order nonlinear optical processes of atomic and molecular systems in intense and superintense laser fields, leading to the discovery of a host of novel strong-field phenomena which cannot be understood by the conventional perturbation theory. The Floquet theorem and the time-independent Floquet Hamiltonian method are powerful theoretical framework for the study of bound-bound multiphoton transitions driven by periodically time-dependent fields. However, there are a number of significant strong-field processes cannot be directly treated by the conventional Floquet methods. In this review article, we discuss several recent developments of generalized Floquet theorems, formalisms, and quasienergy methods, beyond the conventional Floquet theorem, for accurate nonperturbative treatment of a broad range of strong-field atomic and molecular processes and phenomena of current interests. Topics covered include (a) artificial intelligence (AI)-most-probable-path approach (MPPA) for effective treatment of ultralarge Floquet matrix problem; (b) non-Hermitian Floquet formalisms and complex quasienergy methods for nonperturbative treatment of bound-free and free-free processes such as multiphoton ionization (MPI) and above-threshold ionization (ATI) of atoms and molecules, multiphoton dissociation (MPD) and above-threshold dissociation (ATD) of molecules, chemical bond softening and hardening, charge-resonance enhanced ionization (CREI) of molecular ions, and multiple high-order harmonic generation (HHG), etc.; (c) many-mode Floquet theorem (MMFT) for exact treatment of multiphoton processes in multi-color laser fields with nonperiodic time-dependent Hamiltonian; (d) Floquet-Liouville supermatrix (FLSM) formalism for exact nonperturbative treatment of time-dependent Liouville equation (allowing for relaxations and

  10. Mirror hybrid (fusion--fission) reactor

    International Nuclear Information System (INIS)

    Bender, D.J.; Lee, J.D.; Neef, W.S.; Devoto, R.S.; Galloway, T.R.; Fink, J.H.; Schultz, K.R.; Culver, D.; Rao, S.

    1977-10-01

    The reference mirror hybrid reactor design performed by LLL and General Atomic is summarized. The reactor parameters have been chosen to minimize the cost of producing fissile fuel for consumption in fission power reactors. As in the past, we have emphasized the use of existing technology where possible and a minimum extrapolation of technology otherwise. The resulting reactor may thus be viewed as a comparatively near-term goal of the fusion program, and we project improved performance for the hybrid in the future as more advanced technology becomes available

  11. Reactor safety: a discussion by officials of the Nuclear Regulatory Commission

    International Nuclear Information System (INIS)

    Anders, W.A.; Rusche, B.C.; Stello, V. Jr.; Minogue, R.B.

    1976-01-01

    William A. Anders, Chairman of the Nuclear Regulatory Commission (NRC), and several senior officials spoke to the Joint Committee on Atomic Energy on the subject of nuclear safety, improvements in reactor plant safety, and quality assurance. The NRC, during its first year of organization, has developed new initiatives to improve safety and safeguards regulations. Anders stressed that NRC is not stifling internal discussion of opposing views, that it has been honest with the public, and that operating reactors are meeting rigorous safety standards. Other speakers discussed comparative safety of old and new reactors. Backfitting of older plants with new features is done when substantial safety protection can be added, but detuning an integrated system is not done indiscriminately. Officials of NRC do not agree with former General Electric employees, who testified that the regulatory procedure is inadequate. Safety improvements since August 28, 1962 and outlines of the review process are included in the Appendixes

  12. Flow and pressure profiles for the primary heat transport system of Rajasthan Atomic Power Station for the operation with few isolated reactor channels near the end shield cracks

    Energy Technology Data Exchange (ETDEWEB)

    Gaikwad, A J; Chaki, S K; Sehgal, R L; Venkat Raj, V [Reactor Safety Division, Bhabha Atomic Research Centre, Mumbai (India)

    1994-06-01

    The RAPS (Rajasthan Atomic Power Station) unit-1 is now operating at reduced power due to the removal of fifteen fuel channels for repair of south end shield cracks. The power level is restricted to 50% of the full power capacity as a precautionary measure. The relative difference that operation at 50% power and higher power would make to the end shield structure is being currently analysed with a view to operate this reactor at higher power levels. As a prerequisite, a detailed thermal hydraulic analysis is essential to assess the effect of reactor operation with isolated channels on the primary heat transport (PHT) system pressure, flow, temperature. The adequacy of the existing trip set points for the plant operation under this mode is also required to be assessed. In the present study, analysis of the PHT system has been carried out to determine the flow and pressure profiles for the RAPS heat transport system for operation of the reactor with isolated channels. (author). 5 refs., 1 fig., 1 tab.

  13. State of opening the cover and carrying out the checkup of the reactor vessel of the nuclear-powered ship 'Mutsu' by Japan Atomic Energy Research Institute

    International Nuclear Information System (INIS)

    1989-01-01

    In the checkup by opening the cover of the reactor vessel of the nuclear-powered ship 'Mutsu', Japan Atomic Energy Research Institute carried out the checkup and maintenance for the reactor proper, control system and primary coolant facilities including the secondary side of steam generators and the pressure balancing valve of the containment vessel. The works were classified into the opening of the reactor, checkup, maintenance and restoration. The opening was begun on August 4, 1988, and finished on December 5. The checkup and maintenance were begun on September 22, and are still continued now. The maximum radiation dose rate on the surfaces of fuel assemblies and control rods and at the positions 1 m distant from them was measured. The results of the checkup of various components are reported. In 290 absorbent rods of control rods, spot corrosion and discoloration were observed, of which the spot corrosion penetrated the walls of 4 rods. Also in 12 fuel rods, spot corrosion was observed near the welded end plugs, but leak was not observed. (K.I.)

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

    International Nuclear Information System (INIS)

    Gougar, Hans

    2010-01-01

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

  15. Double standards with radioactive waste

    International Nuclear Information System (INIS)

    Parkinson, A.

    2002-01-01

    The author claims that whether it is short-lived waste from Commonwealth facilities, long-lived plutonium waste from an atomic bomb test site on Aboriginal land, or reactor waste from Lucas Heights, the Australian government applies double standards to suit its own agenda

  16. Effective description of general extensions of the Standard Model: the complete tree-level dictionary

    Science.gov (United States)

    de Blas, J.; Criado, J. C.; Pérez-Victoria, M.; Santiago, J.

    2018-03-01

    We compute all the tree-level contributions to the Wilson coefficients of the dimension-six Standard-Model effective theory in ultraviolet completions with general scalar, spinor and vector field content and arbitrary interactions. No assumption about the renormalizability of the high-energy theory is made. This provides a complete ultraviolet/infrared dictionary at the classical level, which can be used to study the low-energy implications of any model of interest, and also to look for explicit completions consistent with low-energy data.

  17. Study of inflationary generalized cosmic Chaplygin gas for standard and tachyon scalar fields

    Energy Technology Data Exchange (ETDEWEB)

    Sharif, M.; Saleem, Rabia [University of the Punjab, Department of Mathematics, Lahore (Pakistan)

    2014-07-15

    We consider an inflationary universe model in the context of the generalized cosmic Chaplygin gas by taking the matter field as standard and tachyon scalar fields. We evaluate the corresponding scalar fields and scalar potentials during the intermediate and logamediate inflationary regimes by modifying the first Friedmann equation. In each case, we evaluate the number of e-folds, scalar as well as tensor power spectra, scalar spectral index, and the important observational parameter, the tensor-scalar ratio in terms of inflation. The graphical behavior of this parameter shows that the model remains incompatible with WMAP7 and Planck observational data in each case. (orig.)

  18. Study of inflationary generalized cosmic Chaplygin gas for standard and tachyon scalar fields

    International Nuclear Information System (INIS)

    Sharif, M.; Saleem, Rabia

    2014-01-01

    We consider an inflationary universe model in the context of the generalized cosmic Chaplygin gas by taking the matter field as standard and tachyon scalar fields. We evaluate the corresponding scalar fields and scalar potentials during the intermediate and logamediate inflationary regimes by modifying the first Friedmann equation. In each case, we evaluate the number of e-folds, scalar as well as tensor power spectra, scalar spectral index, and the important observational parameter, the tensor-scalar ratio in terms of inflation. The graphical behavior of this parameter shows that the model remains incompatible with WMAP7 and Planck observational data in each case. (orig.)

  19. Training and qualification of licensed reactor operators at General Public Utilities Nuclear Corporation

    International Nuclear Information System (INIS)

    Long, R.L.; Coe, R.P.

    1992-01-01

    Following the Three Mile Island-2 (TMI-2) accident in 1979, the utility responsible for managing the facility has looked closely at the training and qualification of its reactor operators. Performance-based operator training programmes are now in place, as required by the United States National Academy for Nuclear Training. Operators also participate directly in the development of a professional code of behaviour. (UK)

  20. Application of JAERI research reactors to education

    International Nuclear Information System (INIS)

    Ogawa, Shigeru; Morozumi, Minoru

    1987-01-01

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

  1. Guidelines for nuclear reactor equipments safety-analysis

    International Nuclear Information System (INIS)

    1978-01-01

    The safety analysis in approving the applications for nuclear reactor constructions (or alterations) is performed by the Committee on Examination of Reactor Safety in accordance with various guidelines prescribed by the Atomic Energy Commission. In addition, the above Committee set forth its own regulations for the safety analysis on common problems among various types of nuclear reactors. This book has collected and edited those guidelines and regulations. It has two parts: Part I includes the guidelines issued to date by the Atomic Energy Commission: and Part II - regulations of the Committee. Part I has collected 8 categories of guidelines which relate to following matters: nuclear reactor sites analysis guidelines and standards for their applications; standard exposure dose of plutonium; nuclear ship operation guidelines; safety design analysis guidelines for light-water type, electricity generating nuclear reactor equipments; safety evaluation guidelines for emergency reactor core cooling system of light-water type power reactors; guidelines for exposure dose target values around light-water type electricity generating nuclear reactor equipments, and guidelines for evaluation of above target values; and meteorological guidelines for the safety analysis of electricity generating nuclear reactor equipments. Part II includes regulations of the Committee concerning - the fuel assembly used in boiling-water type and in pressurized-water type reactors; techniques of reactor core heat designs, etc. in boiling-water reactors; and others

  2. Interlayer Exchange Coupling: A General Scheme Turning Chiral Magnets into Magnetic Multilayers Carrying Atomic-Scale Skyrmions.

    Science.gov (United States)

    Nandy, Ashis Kumar; Kiselev, Nikolai S; Blügel, Stefan

    2016-04-29

    We report on a general principle using interlayer exchange coupling to extend the regime of chiral magnetic films in which stable or metastable magnetic Skyrmions can appear at a zero magnetic field. We verify this concept on the basis of a first-principles model for a Mn monolayer on a W(001) substrate, a prototype chiral magnet for which the atomic-scale magnetic texture is determined by the frustration of exchange interactions, impossible to unwind by laboratory magnetic fields. By means of ab initio calculations for the Mn/W_{m}/Co_{n}/Pt/W(001) multilayer system we show that for certain thicknesses m of the W spacer and n of the Co reference layer, the effective field of the reference layer fully substitutes the required magnetic field for Skyrmion formation.

  3. Final Environmental Statement related to license renewal and power increase for the National Bureau of Standards Reactor: Docket No. 50-184

    International Nuclear Information System (INIS)

    1982-08-01

    This Final Environmental Statement contains an assessment of the environmental impact associated with renewal of Operating License No. TR-5 for the National Bureau of Standards (NBS) reactor for a period of 20 years at a power level of 20 MW. This reactor is located on the 576-acre NBS site near Gaithersburg in Montgomery County, Maryland, about 20 mi northwest of the center of Washington, DC. The reactor is a high-flux heavy-water-moderated, cooled and reflected test reactor, which first went critical on December 7, 1967. Though the reactor was originally designed for 20-MW operation, it has been operating for 14 years at a maximum authorized power level to 10 MW. Program demand is now great enough to warrant operation at a power level of 20 MW. No additional major changes to the physical plant are required to operate at 20 MW

  4. Shutdown channels and fitted interlocks in atomic reactors; Chaines de securite et verrouillages installes sur les piles atomiques

    Energy Technology Data Exchange (ETDEWEB)

    Furet, J; Landauer, C [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1968-07-01

    This catalogue consists of tables (one per reactor) giving the following information: number and type of detectors, range of the shutdown channels, nature of the associated electronics, thresholds setting off the alarms, fitted interlocks. These cards have been drawn up with a view to an examination of the reactors safety by the 'Reactor Safety Sub-Commission', they take into account the latest decisions. The reactors involved in this review are: Azur, Cabri, Castor-Pollux, Cesar-Marius-2, Edf-2, EL3, EL4, Eole, G1, G2-G3, Harmonie, Isis, Masurca, Melusine, Minerve, Osiris, Pegase, Peggy, PAT, Rapsodie, SENA, Siloe, Siloette, Triton-Nereide, and Ulysse. (authors) [French] Ce catalogue est compose d'un ensemble de tableaux (a raison de un tableau par pile) donnant les renseignements suivants: nombre et nature des detecteurs, dynamique des chaines, nature de l'electronique associee, seuils provoquant des actions de securite, verrouillages installes. Ces fiches ont ete etablies en vue de l'examen de la securite des piles par la 'Sous-Commission de Surete des Piles', et tiennent compte des decisions de celle-ci. Les reacteurs concernes sont: Azur, Cabri, Cator-Pollux, Cesar-Marius-2, Edf-2, EL3, EL4, Eole, G1, G2-G3, Harmonie, Isis, Masurca, Melusine, Minerve, Osiris, Pegase, Peggy, PAT, Rapsodie, SENA, Siloe, Siloette, Triton-Nereide, et Ulysse. (auteurs)

  5. On generalized Yang-Mills theories and extensions of the standard model in Clifford (tensorial) spaces

    International Nuclear Information System (INIS)

    Castro, Carlos

    2006-01-01

    We construct the Clifford-space tensorial-gauge fields generalizations of Yang-Mills theories and the Standard Model that allows to predict the existence of new particles (bosons, fermions) and tensor-gauge fields of higher-spins in the 10 Tev regime. We proceed with a detailed discussion of the unique D 4 - D 5 - E 6 - E 7 - E 8 model of Smith based on the underlying Clifford algebraic structures in D = 8, and which furnishes all the properties of the Standard Model and Gravity in four-dimensions, at low energies. A generalization and extension of Smith's model to the full Clifford-space is presented when we write explicitly all the terms of the extended Clifford-space Lagrangian. We conclude by explaining the relevance of multiple-foldings of D = 8 dimensions related to the modulo 8 periodicity of the real Cliford algebras and display the interplay among Clifford, Division, Jordan, and Exceptional algebras, within the context of D = 26, 27, 28 dimensions, corresponding to bosonic string, M and F theory, respectively, advanced earlier by Smith. To finalize we describe explicitly how the E 8 x E 8 Yang-Mills theory can be obtained from a Gauge Theory based on the Clifford (16) group

  6. Atomic clocks for geodesy

    Science.gov (United States)

    Mehlstäubler, Tanja E.; Grosche, Gesine; Lisdat, Christian; Schmidt, Piet O.; Denker, Heiner

    2018-06-01

    We review experimental progress on optical atomic clocks and frequency transfer, and consider the prospects of using these technologies for geodetic measurements. Today, optical atomic frequency standards have reached relative frequency inaccuracies below 10‑17, opening new fields of fundamental and applied research. The dependence of atomic frequencies on the gravitational potential makes atomic clocks ideal candidates for the search for deviations in the predictions of Einstein’s general relativity, tests of modern unifying theories and the development of new gravity field sensors. In this review, we introduce the concepts of optical atomic clocks and present the status of international clock development and comparison. Besides further improvement in stability and accuracy of today’s best clocks, a large effort is put into increasing the reliability and technological readiness for applications outside of specialized laboratories with compact, portable devices. With relative frequency uncertainties of 10‑18, comparisons of optical frequency standards are foreseen to contribute together with satellite and terrestrial data to the precise determination of fundamental height reference systems in geodesy with a resolution at the cm-level. The long-term stability of atomic standards will deliver excellent long-term height references for geodetic measurements and for the modelling and understanding of our Earth.

  7. Standardization of accelerator irradiation procedures for simulation of neutron induced damage in reactor structural materials

    Science.gov (United States)

    Shao, Lin; Gigax, Jonathan; Chen, Di; Kim, Hyosim; Garner, Frank A.; Wang, Jing; Toloczko, Mychailo B.

    2017-10-01

    Self-ion irradiation is widely used as a method to simulate neutron damage in reactor structural materials. Accelerator-based simulation of void swelling, however, introduces a number of neutron-atypical features which require careful data extraction and, in some cases, introduction of innovative irradiation techniques to alleviate these issues. We briefly summarize three such atypical features: defect imbalance effects, pulsed beam effects, and carbon contamination. The latter issue has just been recently recognized as being relevant to simulation of void swelling and is discussed here in greater detail. It is shown that carbon ions are entrained in the ion beam by Coulomb force drag and accelerated toward the target surface. Beam-contaminant interactions are modeled using molecular dynamics simulation. By applying a multiple beam deflection technique, carbon and other contaminants can be effectively filtered out, as demonstrated in an irradiation of HT-9 alloy by 3.5 MeV Fe ions.

  8. The processing and management of wastes from atomic reactors; Nouvelles installations industrielles du C.E.A. pour le traitement des dechets radioactifs liquides et solides

    Energy Technology Data Exchange (ETDEWEB)

    Cerre, P; Mestre, E; Bourdrez, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    The policy concerning radioactive wastes studied by all Atomic Centres has led to various procedures which, while apparently numerous, come under a few standard headings. Whether the wastes are in the liquid or solid state their management depends on their physical and chemical nature. The procedure adopted is governed by three general principles: - determination of the most economical means possible of storage and processing by volume reduction; - conversion to a solid compact form; - complete acceptance of the accepted standards at all places and all times. In this communication all the standard solutions adopted and used by the various Centres of the Commissariat a l'Energie Atomique will be examined bearing in mind the preceding remarks. Particular mention will be made of the following: - For liquids, physical, chemical and physico-chemical processing - For solids, decontamination, volume reduction and long-term conditioning techniques. The different procedures for collecting and storing solid wastes before and after processing are also discussed. The paper ends with a brief review of the studies, both technical and economic, being pursued on this subject. (authors) [French] La gestion des dechets etudies par tous les Centres Atomiques a donne lieu a des solutions qui - bien que nombreuses en apparence - se ramenent a quelques solutions types, peu nombreuses. Qu'il s'agisse de dechets solides ou liquides, la nature physique et chimique des dechets conditionne leur mode de gestion. Celle-ci procede de trois principes generaux: - recherche du mode de stockage et de traitement aussi economique que possible par reduction de volume; - mise sous forme compacte solide; - garantie du respect des normes en tous lieux et en tous temps. Dans cette communication, nous examinons toutes les solutions types, compte tenu des remarques precedentes, qui ont ete adoptees et sont utilisees par les differents Centres du Commissariat a l'Energie Atomique. Nous rappelons en

  9. Insights into the Electronic Structure of Ozone and Sulfur Dioxide from Generalized Valence Bond Theory: Addition of Hydrogen Atoms.

    Science.gov (United States)

    Lindquist, Beth A; Takeshita, Tyler Y; Dunning, Thom H

    2016-05-05

    Ozone (O3) and sulfur dioxide (SO2) are valence isoelectronic species, yet their properties and reactivities differ dramatically. In particular, O3 is highly reactive, whereas SO2 is chemically relatively stable. In this paper, we investigate serial addition of hydrogen atoms to both the terminal atoms of O3 and SO2 and to the central atom of these species. It is well-known that the terminal atoms of O3 are much more amenable to bond formation than those of SO2. We show that the differences in the electronic structure of the π systems in the parent triatomic species account for the differences in the addition of hydrogen atoms to the terminal atoms of O3 and SO2. Further, we find that the π system in SO2, which is a recoupled pair bond dyad, facilitates the addition of hydrogen atoms to the sulfur atom, resulting in stable HSO2 and H2SO2 species.

  10. Safety evaluation report related to the license renewal and power increase for the National Bureau of Standards reactor (Docket No. 50-184)

    International Nuclear Information System (INIS)

    Bernard, H.

    1984-03-01

    Supplement 1 to the Safety Evaluation Report (SER) related to the renewal of the operating license and for a power increase (10 MWt to 20 MWt) for the research reactor at the National Bureau of Standards (NBS) facility has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement reports on the review of the licensee's emergency plan, which had not been reviewed at the time the Safety Evaluation Report (NUREG-1007) was published, and the review of the NBS application by the Advisory Committee on Reactor Safeguards, which was completed subsequent to the publication of the SER

  11. NJOY-94, General ENDF/B Processing System for Reactor Design Problems

    International Nuclear Information System (INIS)

    1997-01-01

    1 - Description of program or function: The NJOY nuclear data processing system is a comprehensive computer code system for producing pointwise and multigroup cross sections and related quantities from ENDF/B evaluated nuclear data in the ENDF format, including the latest US library, ENDF/B-VI. The NJOY code works with neutrons, photons, and charged particles and produces libraries for a wide variety of particle transport and reactor analysis codes. It is capable of processing data in ENDF/B-4, ENDF/B-5, and ENDF/B-6 formats for evaluated data. Short descriptions of the different modules follow: RECONR Reconstructs pointwise cross sections from ENDF/B resonance parameters and interpolation schemes. BROADR Doppler broadens and thins pointwise cross sections. UNRESR Computes effective self-shielded pointwise cross sections in the unresolved-resonance region. HEATR Generates pointwise heat production cross sections and radiation-damage-energy production cross sections. THERMR Produces incoherent inelastic energy-to-energy matrices for free or bound scatterers, coherent elastic cross sections for hexagonal materials, and incoherent elastic cross sections. GROUPR Generates self-shielded multigroup cross sections, group- to-group neutron scattering matrices, and photon production matrices from pointwise input. GAMINR Calculates multigroup photon interaction cross sections and KERMA factors and group-to-group photon scattering matrices. ERRORR Produces multigroup covariance matrices from ENDF/B uncertainties. COVR Reads the output of ERRORR and performs covariance plotting and output-formatting operations. DTFR Formats multigroup data for transport codes such as DTF-IV and ANISN. CCCCR Formats multigroup data for the CCCC standard interface files ISOTXS, BRKOXS, and DLAYXS. MATXSR Formats multigroup data for the MATXS cross section interface file. ACER Prepares libraries for the Los Alamos continuous-energy Monte Carlo code MCNP. POWR Prepares libraries for the EPRI

  12. A diode-laser optical frequency standard based on laser-cooled Ca atoms: sub-kilohertz spectroscopy by optical shelving detection

    International Nuclear Information System (INIS)

    Oates, C.W.; Bondu, F.; Fox, R.W.; Hollberg, L.

    1999-01-01

    We report an optical frequency standard at 657 nm based on laser-cooled/trapped Ca atoms. The system consists of a novel, compact magneto-optic trap which uses 50 mW of frequency-doubled diode laser light at 423 nm and can trap >10 7 Ca atoms in 20 ms. High resolution spectroscopy on this atomic sample using the narrow 657 nm intercombination line resolves linewidths (FWHM) as narrow as 400 Hz, the natural linewidth of the transition. The spectroscopic signal-to-noise ratio is enhanced by an order of magnitude with the implementation of a ''shelving'' detection scheme on the 423 nm transition. Our present apparatus achieves a fractional frequency instability of 5 x 10 -14 in 1 s with a potential atom shot-noise-limited performance of 10 -16 τ -1/2 and excellent prospects for high accuracy. (orig.)

  13. Offsite dose calculation manual guidance: Standard radiological effluent controls for boiling water reactors

    International Nuclear Information System (INIS)

    Meinke, W.W.; Essig, T.H.

    1991-04-01

    This report contains guidance which may be voluntarily used by licensees who choose to implement the provision of Generic Letter 89-- 01, which allows Radiological Effluent Technical Specifications (RETS) to be removed from the main body of the Technical Specifications and placed in the Offsite Dose Calculation Manual (ODCM). Guidance is provided for Standard Effluent Controls definitions, Controls for effluent monitoring instrumentation, Controls for effluent releases, Controls for radiological environmental monitoring, and the basis for Controls. Guidance on the formulation of RETS has been available in draft form for a number of years; the current effort simply recasts those RETS into Standard Radiological Effluent Controls for application to the ODCM. 11 tabs

  14. Development of a standard for calculation and measurement of the moderator temperature coefficient of reactivity in water-moderated power reactors

    International Nuclear Information System (INIS)

    Mosteller, R.D.; Hall, R.A.; Lancaster, D.B.; Young, E.H.; Gavin, P.H.; Robertson, S.T.

    1998-01-01

    The contents of ANS 19.11, the standard for ''Calculation and Measurement of the Moderator Temperature Coefficient of Reactivity in Water-Moderated Power Reactors,'' are described. The standard addresses the calculation of the moderator temperature coefficient (MTC) both at standby conditions and at power. In addition, it describes several methods for the measurement of the at-power MTC and assesses their relative advantages and disadvantages. Finally, it specifies a minimum set of documentation requirements for compliance with the standard

  15. Clinch River Breeder Reactor Plant. License application, statement of general information

    International Nuclear Information System (INIS)

    1975-01-01

    Application is made for a reactor facility consisting of a liquid metal cooled reactor and steam generator system, a steam turbine driven electric generating system, electrical switchyard, and related auxiliaries and supporting structures. The primary system is located in an inert atmosphere in shielded vaults within a containment structure. Sodium coolant is used to remove heat from the core and radial blanket. Heat from the primary sodium is transferred in heat exchangers to non radioactive sodium which is used to convert feed-water into steam which is superheated to drive a tandem-compound generator. A single shaft multi-stage turbine generator produces 380 MW(e) with steam conditions of 1450 psig at 900 0 F. Fuel is sintered ceramic pellets of mixed uranium-plutonium oxides encapsulated in stainless steel. There are 198 fuel assemblies with each assembly consisting of 217 fuel rods placed in a hexagonal channel. Plutonium enrichment ranges from 1817 to 32.0 percent by weight. Axial blanket sections contain depleted UO 2 with 99.8 percent 238 U and 0.2 percent 235 U by weight. The proposed location of the plant is within the corporate limits of the city of Oak Ridge in Roane County, Tennessee. (U.S.)

  16. Fuel performance of rod-type research reactor fuel using a centrifugally atomized U-Mo powder

    International Nuclear Information System (INIS)

    Ryu, Ho Jin; Park, Jong Man; Lee, Yoon Sang; Kim, Chang Kyu

    2009-01-01

    A low enriched uranium nuclear fuel for research reactors has been developed in order to replace a highly enriched uranium fuel according to the non-proliferation policy under the reduced enrichment for research and test reactors (RERTR) program. In KAERI, a rod-type U 3 Si dispersion fuel has been developed for a localization of the HANARO fuel and a U 3 Si/Al dispersion fuel of 3.15 gU/cc has been used at HANARO as a driver fuel since 2005. Although uranium silicide dispersion fuels such as U 3 Si 2 /Al and U 3 Si/Al are being used widely, high uranium density dispersion fuels (8-9 g/cm 3 ) are required for some high performance research reactors. U-Mo alloys have been considered as one of the most promising uranium alloys for a dispersion fuel due to their good irradiation performance. An international qualification program on U-Mo fuel to replace a uranium silicide dispersion fuel with a U-Mo dispersion fuel has been carried out

  17. Evaluation and standardization of neutron activation analysis according to the K0 method in the RP-10 reactor

    International Nuclear Information System (INIS)

    Montoya R, E.

    1995-01-01

    It has been characterized and standardized an irradiation of the RP-10 Research Nuclear Reactor for use of the K 0 method of neutron activation analysis using the Hoegdahl convention; also it has been evaluate the behaviour of such method in regard to the accuracy and precision of the results obtained in the quantitative multi elemental analysis of several certified materials of reference. In order to prove that the analytical method is totally under statistical control, it has been used the Heydorn method. It has been verified that the method is exact, precise and reliable to determine the aluminium, antimuonium, arsenic, bromine, calcium, chloride, copper, magnesium, manganese, sodium, titanium, vanadium, zinc and other elements. Also, they are discussed, in regard to the use of K 0 constants, the different formalisms employed to calculate the integral of the reaction rate by nucleus in the activation. (author). 58 refs., 18 tabs., 6 figs

  18. Reactor operation

    CERN Document Server

    Shaw, J

    2013-01-01

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

  19. 1: the atom. 2: radioactivity. 3: man and radiations. 4: the energy. 5: nuclear energy: fusion and fission. 6: the operation of a nuclear reactor. 7: the nuclear fuel cycle

    International Nuclear Information System (INIS)

    2002-01-01

    This series of 7 digest booklets present the bases of the nuclear physics and of the nuclear energy: 1 - the atom (structure of matter, chemical elements and isotopes, the four fundamental interactions, nuclear physics); 2 - radioactivity (definition, origins of radioelements, applications of radioactivity); 3 - man and radiations (radiations diversity, biological effects, radioprotection, examples of radiation applications); 4 - energy (energy states, different forms of energy, characteristics); 5 - nuclear energy: fusion and fission (nuclear energy release, thermonuclear fusion, nuclear fission and chain reaction); 6 - operation of a nuclear reactor (nuclear fission, reactor components, reactor types); 7 - nuclear fuel cycle (nuclear fuel preparation, fuel consumption, reprocessing, wastes management). (J.S.)

  20. General-purpose stepping motor-encoder positioning subsystem with standard asynchronous serial-line interface

    International Nuclear Information System (INIS)

    Stubblefield, F.W.; Alberi, J.L.

    1982-01-01

    A general-purpose mechanical positioning subsystem for open-loop control of experiment devices which have their positions established and read out by stepping motor-encoder combinations has been developed. The subsystem is to be used mainly for experiments to be conducted at the National Synchrotron Light Source at Brookhaven National Laboratory. The subsystem unit has been designed to be compatible with a wide variety of stepping motor and encoder types. The unit may be operated by any device capable of driving a standard RS-232-C asynchronous serial communication line. An informal survey has shown that several experiments at the Light Source will use one particular type of computer, operating system, and programming language. Accordingly, a library of subroutines compatible with this combination of computer system elements has been written to facilitate driving the positioning subsystem unit

  1. Standard Practice for Characterizing Neutron Exposures in Iron and Low Alloy Steels in Terms of Displacements Per Atom (DPA), E 706(ID)

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2001-01-01

    1.1 This practice describes a standard procedure for characterizing neutron irradiations of iron (and low alloy steels) in terms of the exposure index displacements per atom (dpa) for iron. 1.2 Although the general procedures of this practice apply to any material for which a displacement cross section d(E) is known (see Practice E 521), this practice is written specifically for iron. 1.3 It is assumed that the displacement cross section for iron is an adequate approximation for calculating displacements in steels that are mostly iron (95 to 100 %) in radiation fields for which secondary damage processes are not important. 1.4 Procedures analogous to this one can be formulated for calculating dpa in charged particle irradiations. (See Practice E 521.) 1.5 The application of this practice requires knowledge of the total neutron fluence and flux spectrum. Refer to Practice E 521 for determining these quantities. 1.6 The correlation of radiation effects data is beyond the scope of this practice. This stand...

  2. Statement to the 41st session of the General Conference of the International Atomic Energy Agency 29 September 1997; Statement to the 52nd session of the United Nations General Assembly 12 November 1997

    International Nuclear Information System (INIS)

    Blix, H.

    1997-12-01

    The document presents Statement to the 41st Session of the General Conference of the International Atomic Energy Agency and Statement to the 52nd Session of the United Nations General Assembly made at the 40th anniversary of the IAEA

  3. Statement to the 40th session of the General Conference of the International Atomic Energy Agency 16 September 1996; Statement to the 51st session of the United Nations General Assembly 28 October 1996

    International Nuclear Information System (INIS)

    Blix, H.

    1996-11-01

    In the Statement to the 40th Session of the General Conference of the International Atomic Energy Agency main directions of the IAEA activities and plans for the future are outlined. In the Statement to the 51th Session of the United Nations General Assembly main achievements and challenge facing the international community in the field of IAEA competence are discussed

  4. Management of research reactor ageing

    International Nuclear Information System (INIS)

    1995-03-01

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

  5. Management of research reactor ageing

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-03-01

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

  6. Offsite dose calculation manual guidance: Standard radiological effluent controls for pressurized water reactors

    International Nuclear Information System (INIS)

    Meinke, W.W.; Essig, T.H.

    1991-04-01

    This report contains guidance which may be voluntarily used by licensees who choose to implement the provision of Generic Letter 89-01, which allows Radiological Effect Technical Specifications (RETS) to be removed from the main body of the Technical Specifications and placed in the Offsite Dose Calculation Manual (ODCM). Guidance is provided for Standard Effluent Controls definitions, Controls for effluent monitoring instrumentation, Controls for effluent releases, Controls for radiological environmental monitoring, and the basis for Controls. Guidance on the formulation of RETS has been available in draft from (NUREG-0471 and -0473) for a number of years; the current effort simply recasts those RETS into Standard Radiological Effluent Controls for application to the ODCM. Also included for completeness are: (1) radiological environmental monitoring program guidance previously which had been available as a Branch Technical Position (Rev. 1, November 1979); (2) existing ODCM guidance; and (3) a reproduction of generic Letter 89-01

  7. Estimation of CANDU reactor zone controller level by generalized perturbation method

    International Nuclear Information System (INIS)

    Kim, Do Heon; Kim, Jong Kyung; Choi, Hang Bok; Roh, Gyu Hong; Yang, Won Sik

    1998-01-01

    The zone controller level change due to refueling operation has been studied using a generalized perturbation method. The generalized perturbation method provides sensitivity of zone power to individual refueling operation and incremental change of zone controller level. By constructing a system equation for each zone power, the zone controller level change was obtained. The details and a proposed model for future work are described

  8. Utilization of research reactors

    International Nuclear Information System (INIS)

    1962-01-01

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

  9. Communication to the Director General of 24 April 1996 received from the resident representative of Saudi Arabia to the International Atomic Energy Agency

    International Nuclear Information System (INIS)

    1996-01-01

    The document reproduces the text of a letter dated 24 April 1996 received by the Director General of the IAEA on 25 April 1996 from the Resident Representative of Saudi Arabia on behalf of Members of the Agency and Members of the Board of Governors from the Arab States, concerning the Israeli Dimona reactor

  10. Narrow Line Cooling of 88Sr Atoms in the Magneto-optical Trap for Precision Frequency Standard

    Science.gov (United States)

    Strelkin, S. A.; Galyshev, A. A.; Berdasov, O. I.; Gribov, A. Yu.; Sutyrin, D. V.; Khabarova, K. Yu.; Kolachevsky, N. N.; Slyusarev, S. N.

    We report on our progress toward the realization of a Strontium optical lattice clock, which is under development at VNIIFTRI as a part of GLONASS program. We've prepared the narrow line width laser system for secondary cooling of 88Sr atoms which allows us to reach atom cloud temperature below 3 μK after second cooling stage.

  11. Proceedings of the twenty-first symposium of atomic energy research on WWER physics and reactor safety

    International Nuclear Information System (INIS)

    Vidovszky, I.

    2011-10-01

    The present volume contains 61 papers, presented on the twenty-first symposium of atomic energy research, held in Dresden, Germany, 19-23 September 2011. The papers are presented in their original form, i. e. no corrections or modifications were carried out. The content of this volume is divided into thematic groups: Improvement, extension and validation of parameterized few-group libraries for WWER-440 and WWER-1000.

  12. A generalized adjoint framework for sensitivity and global error estimation in time-dependent nuclear reactor simulations

    International Nuclear Information System (INIS)

    Stripling, H.F.; Anitescu, M.; Adams, M.L.

    2013-01-01

    Highlights: ► We develop an abstract framework for computing the adjoint to the neutron/nuclide burnup equations posed as a system of differential algebraic equations. ► We validate use of the adjoint for computing both sensitivity to uncertain inputs and for estimating global time discretization error. ► Flexibility of the framework is leveraged to add heat transfer physics and compute its adjoint without a reformulation of the adjoint system. ► Such flexibility is crucial for high performance computing applications. -- Abstract: We develop a general framework for computing the adjoint variable to nuclear engineering problems governed by a set of differential–algebraic equations (DAEs). The nuclear engineering community has a rich history of developing and applying adjoints for sensitivity calculations; many such formulations, however, are specific to a certain set of equations, variables, or solution techniques. Any change or addition to the physics model would require a reformulation of the adjoint problem and substantial difficulties in its software implementation. In this work we propose an abstract framework that allows for the modification and expansion of the governing equations, leverages the existing theory of adjoint formulation for DAEs, and results in adjoint equations that can be used to efficiently compute sensitivities for parametric uncertainty quantification. Moreover, as we justify theoretically and demonstrate numerically, the same framework can be used to estimate global time discretization error. We first motivate the framework and show that the coupled Bateman and transport equations, which govern the time-dependent neutronic behavior of a nuclear reactor, may be formulated as a DAE system with a power constraint. We then use a variational approach to develop the parameter-dependent adjoint framework and apply existing theory to give formulations for sensitivity and global time discretization error estimates using the adjoint

  13. Consequences arising for research reactor operation from the planned amendment of the German atomic energy act (AtG) and atomic energy policy. Germany soon to be a third-world country in nuclear research?

    International Nuclear Information System (INIS)

    Krull, W.

    1999-01-01

    In the opinion of Dr. Wilfried Krull from Geesthacht, the Chairman of the German Research Reactor Trades Union, there is a great danger of the Federal Republic of Germany falling to the level of a third-world country as far as nuclear research is concerned. He says that it is a matter of urgency for binding exceptions from the general restrictions applicable to nuclear power stations to be incorporated into the opt-out amendment to nuclear law tabled by Juergen Trittin, the Minister for Environmental Affairs. He says that prohibition of reprocessing of spent fuel elements constitutes a violation of the surrender treaties concluded with the USA as part of non-proliferation. Furthermore, he states that a ban on reprocessing of spent nuclear fuel from research and training reactors is counterproductive. He says that there would be a particular danger if the limitation of research reactors to a rating of one megawatt, which Trittin has in the first instance planned in an initial draft law, were to be implemented in the form of a regulation. At that point, he claims, Germany would, as far as nuclear medicine (including cancer therapy) and modern technical inspection procedures (non-destructive, of materials) are concerned, become an importing country. (orig.) [de

  14. Russian seismic standards and demands for equipment and their conformity with international standards

    International Nuclear Information System (INIS)

    Kaznovsky, S.; Ostretsov, I.

    1993-01-01

    The principle regulations of standard documents concerning seismic safety of NPPs and demands for reactor equipment conformity with international standards are presented in this report. General state of NPP safety standards is reviewed, with a special emphasis on the state of seismic design standards for NPP equipment and piping. Russian standards documents on seismic resistance of NPPs and requirements are compared to international ones

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

    International Nuclear Information System (INIS)

    Lovasic, Z.; Einziger, R.

    2009-01-01

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

  16. Atomic data for heavy element impurities in fusion reactors. Summary report of the final research coordination meeting

    International Nuclear Information System (INIS)

    Clark, R.E.H.

    2009-04-01

    Eleven experts on the properties of heavy elements of relevance to fusion energy research attended the final Research Coordination Meeting (RCM) on Data for Heavy Element Impurities in Fusion Reactors, held at IAEA Headquarters on 4-6 March 2009. Participants summarized their accomplishments with respect to the revised work plan formulated at the second RCM. The overall work plan for the CRP was assessed and reviewed in detail, and achievements were noted. Discussions, conclusions and recommendations of the RCM are briefly described in this report. (author)

  17. General design and main problems of a gas-heavy-water power reactor contained in a pressure vessel

    International Nuclear Information System (INIS)

    Roche, R.; Gaudez, J.C.

    1964-01-01

    In the framework of research carried out on a CO 2 -cooled power reactor moderated by heavy water, the so-called 'pressure vessel' solution involves the total integration of the core, of the primary circuit (exchanges and blowers) and of the fuel handling machine inside a single, strong, sealed vessel made of pre-stressed concrete. A vertical design has been chosen: the handling 'attic' is placed above the core, the exchanges being underneath. This solution makes it possible to standardize the type of reactor which is moderated by heavy-water or graphite and cooled by a downward stream of carbon dioxide gas; it has certain advantages and disadvantages with respect to the pressure tube solution and these are considered in detail in this report. Extrapolation presents in particular.problems due specifically to the heavy water (for example its cooling,its purification, the balancing of the pressures of the heavy water and of the gas, the assembling of the internal structures, the height of the attic, etc. (authors) [fr

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

    International Nuclear Information System (INIS)

    Vivas, Ary de Souza; Carneiro, Alvaro Luiz Guimaraes

    2013-01-01

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

  19. Elementary particles. From the atoms via the Standard Model until the Higgs boson. 2. ed.; Elementare Teilchen. Von den Atomen ueber das Standard-Modell bis zum Higgs-Boson

    Energy Technology Data Exchange (ETDEWEB)

    Bleck-Neuhaus, Joern [Bremen Univ. (Germany)

    2013-07-01

    The current state of knowledge of nuclear and elementary-particle physics has a checkered history, often characterized by shocking new concept formations, which also opens up to the present day students of physics only with difficulty. This book uses those controversial yet educational development in order to enable learners to improve access to the new concepts. It helps to understand how the physical picture of the smallest particles is today, and why it is so and not otherwise originated: Beginning in the detection of the atoms up to the current standard model of elementary-particle physics and the Higgs boson. So readers gain an impression of that great field, which is originated in the constant interplay between established theoretical models, confirmatory or contradictory findings, sometimes controversial new concept formations, and improved experiments - a process, that surely continues in the future. Guideline of the presentation is a comprehensible also in detail as possible reasoning argumentation. Students of physics before their B.Sc. degree will thus be able to acquire knowledge of the subatomic physics relating to general knowledge in their field. Also for physics teachers at schools or colleges, this new representation should be interesting. The second edition has been updated to the newest state of knowledge, in particular first results of the LHC have been incorporated.

  20. Syntheses of deuterated leu-enkephalins and their use as internal standards for the quantification of leu-enkephalin by fast atom bombardment mass spectrometry

    Energy Technology Data Exchange (ETDEWEB)

    Benfenati, E. (Istituto di Ricerche Farmacologiche Mario Negri, Bergamo (Italy) Istituto di Ricerche Farmacologiche Mario Negri, Milan (Italy)); Icardi, G.; Chen, S. (Istituto di Ricerche Farmacologiche Mario Negri, Bergamo (Italy)); Fanelli, R. (Istituto di Ricerche Farmacologiche Mario Negri, Milan (Italy))

    1990-04-01

    We have developed a synthetic method for the preparation of di- and pentadeuterated leu-enkephalin (LE), Tyr-Gly-Gly-Phe-Leu, by proton-deuterium exchange using CF[sub 3]COOO[sup 2]H. Four to six deuterium atoms are introduced using a reaction temperature of 120[sup o]C and if 5% of [sup 2]H[sub 2]O is added the di-deuterated LE is obtained. These deuterated compounds are used as internal standards to plot calibration curves of LE using fast atom bombardment mass spectrometry. (author).