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Sample records for godiva iv reactor

  1. Source-Term and building-Wake Consequence Modeling for the Godiva IV Reactor at Los Alamos National Laboratory

    International Nuclear Information System (INIS)

    Letellier, B.C.; McClure, P.; Restrepo, L.

    1999-01-01

    The objectives of this work were to evaluate the consequences of a postulated accident to onsite security personnel stationed near the facility during operations of the Godiva IV critical assembly and to identify controls needed to protect these personnel in case of an extreme criticality excursion equivalent to the design-basis accident (DBA). This paper presents the methodology and results of the source-term calculations, building ventilation rates, air concentrations, and consequence calculations that were performed using a multidisciplinary approach with several phenomenology models. Identification of controls needed to mitigate the consequences to near-field receptors is discussed

  2. Dosimetry characterization of the Godiva Reactor under burst conditions

    Energy Technology Data Exchange (ETDEWEB)

    Hickman, D. P. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Heinrichs, D. P. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Hudson, R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Wong, C. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Ward, D. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Wilson, C. [Atomic Weapons Establishment (AWE), Berkshire (United Kingdom); Clark, L. [Atomic Weapons Establishment (AWE), Berkshire (United Kingdom); Trompier, F. [Inst. for Radiation Protection and Nuclear Safety, Fontenay-aux-Roses (France)

    2017-06-22

    A series of sixteen (16) burst irradiations were performed in May 2014, fifteen of which were part of an international collaboration to characterize the Godiva IV fast burst reactor at the National Criticality Experiments Research Center (NCERC). Godiva IV is a bare cylindrical assembly of approximately 65 kg of highly enriched uranium fuel (93.2% 235U metal alloyed with 1.5% molybdenum for strength) and is designed to perform controlled prompt critical excursions (Myers 2010, Goda 2013). Twelve of the irradiations were dedicated to neutron spectral measurements using a Bonner multiple sphere spectrometer. Three irradiations, with core temperature increases of 71.1°C, 136.9°C, and 229.9°C, were performed for generating comparative fluence data, establishing corrections for varying heights, testing linearity with burst temperature, and establishing gamma dose characteristics.

  3. Plan for IER-443 Testing of the Y-12 and AWE Criticality Accident Alarm System Detectors at the Godiva IV Burst Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Scorby, J. C. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Hickman, D. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Hudson, B. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Garbett, S. [Atomic Weapons Establishment (AWE), Berkshire (United Kingdom); Auld, G. [Atomic Weapons Establishment (AWE), Berkshire (United Kingdom); Horrne, A. [Atomic Weapons Establishment (AWE), Berkshire (United Kingdom); Beller, T. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Goda, J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Haught, C. [Y-12 National Security Complex, Oak Ridge, TN (United States); Woodrow, C. [Y-12 National Security Complex, Oak Ridge, TN (United States); Ward, D. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2017-07-24

    This document provides the scope and details of the “Plan for Testing the Y-12 and AWE Criticality Accident Alarm System Detectors at the Godiva IV Burst Reactor”. Due to the relative simplicity of the testing goals, scope, and methodology, the NCSP Manager approved execution of the test when ready. No preliminary CED-1 or final design CED-2 reports were required or issued. The test will subject Criticality Accident Alarm System (CAAS) detectors supplied by Y- 12 and AWE to very intense and short duration mixed neutron and gamma radiation fields. The goals of the test will be to (1) substantiate functionality, for both existing and newly acquired Y- 12 CAAS detectors, and (2) the ability of the AWE detectors to provide quality temporal dose information after a hypothetical criticality accident. ANSI/ANS-8.3.1997 states that the “system shall be sufficiently robust as to actuate an alarm signal when exposed to the maximum radiation expected”, which has been defined at Y-12, in Documented Safety Analyses (DSAs), to be a dose rate of 10 Rad/s. ANSI/ANS-8.3.1997 further states that “alarm actuation shall occur as a result of a minimum duration transient” which may be assumed to be 1 msec. The pulse widths and dose rates which will be achieved in this test will exceed these requirements. Pulsed radiation fields will be produced by the Godiva IV fast metal burst reactor at the National Criticality Experimental Research Center (NCERC) at the Nevada National Security Site (NNSS). The magnitude of the pulses and the relative distances to the detectors will be varied to afford a wide range of radiation fluence and pulse widths. The magnitude of the neutron and gamma fields will be determined by reactor temperature rise to fluence and dose conversions which have been previously established through extensive measurements performed under IER-147. The requirements for CAAS systems to detect and alarm under a “minimum accident of concern” as well as other

  4. Final Report for the Testing of the Y-12 Criticality Accident Alarm System Detectors at the Godiva IV Burst Reactor (IER-443)

    Energy Technology Data Exchange (ETDEWEB)

    Scorby, John C. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Hickman, David [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Hudson, Becka [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Beller, Tim [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Goda, Joetta [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Haught, Chris [Y-12 National Security Complex, Oak Ridge, TN (United States); Woodrow, Christopher [Y-12 National Security Complex, Oak Ridge, TN (United States); Ward, Dann [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Wilson, Chris [Atomic Weapons Establishment (AWE), Berkshire (United Kingdom); Clark, Leo [Atomic Weapons Establishment (AWE), Berkshire (United Kingdom)

    2018-01-05

    This report documents the experimental conditions and final results for the performance testing of the Y-12 Criticality Accident Alarm System (CAAS) detectors at the Godiva IV Burst Reactor at the National Criticality Experimental Research Center (NCERC) at the Nevada National Security Site (NNSS). The testing followed a previously issued test plan and was conducted during the week of July 17, 2017, with completion on Thursday July 20. The test subjected CAAS detectors supplied by Y-12 to very intense and short duration mixed neutron and gamma radiation fields to establish compliance to maximum radiation and minimum pulse width requirements. ANSI/ANS- 8.3.1997 states that the “system shall be sufficiently robust as to actuate an alarm signal when exposed to the maximum radiation expected”, which has been defined at Y-12, in Documented Safety Analyses (DSAs), to be a dose rate of 10 Rad/s. ANSI/ANS-8.3.1997 further states that “alarm actuation shall occur as a result of a minimum duration transient” which may be assumed to be 1 msec. The pulse widths and dose rates provided by each burst during the test exceeded those requirements. The CAAS detectors all provided an immediate alarm signal and remained operable after the bursts establishing compliance to the requirements and fitness for re-deployment at Y-12.

  5. Godiva IV and Juliet Diagnostics CED-1, Rev. 1 (IER-176)

    Energy Technology Data Exchange (ETDEWEB)

    Scorby, J C; Myers, W L

    2012-04-11

    The Juliet experiment is currently in preliminary design (IER-128). This experiment will utilize a suite of diagnostics to measure the physical state of the device (temperature, surface motion, stress, etc.) and the total and time rate of change of neutron and gamma fluxes. A variety of potential diagnostics has been proposed in this CED-1 report. Based on schedule and funding, a subset of diagnostics will be selected for testing using the Godiva IV pulsed reactor as a source of neutrons and gammas. The diagnostics development and testing will occur over a two year period (FY12-13) culminating in a final set of diagnostics to be fielded for he Juliet experiment currently proposed for execution in FY15.

  6. International Intercomparison Exercise for Nuclear Accident Dosimetry at the DAF Using GODIVA-IV

    Energy Technology Data Exchange (ETDEWEB)

    Hickman, David [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Hudson, Becka [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2016-12-15

    The Nuclear Criticality Safety Program operated under the direction of Dr. Jerry McKamy completed the first NNSA Nuclear Accident Dosimetry exercise on May 27, 2016. Participants in the exercise were from Lawrence Livermore National Laboratory (LLNL), Los Alamos National Laboratory (LANL), Sandia National Laboratory (SNL), Savanah River Site (SRS), Pacific Northwest National Laboratory (PNNL), US Navy, the Atomic Weapons Establishment (United Kingdom) under the auspices of JOWOG 30, and the Institute for Radiological Protection and Nuclear Safety (France) by special invitation and NCSP memorandum of understanding. This exercise was the culmination of a series of Integral Experiment Requests (IER) that included the establishment of the Nuclear Criticality Experimental Research Center, (NCERC) the startup of the Godiva Reactor (IER-194), the establishment of a the Nuclear Accident Dosimetry Laboratory (NAD LAB) in Mercury, NV, and the determination of reference dosimetry values for the mixed neutron and photon radiation field of Godiva within NCERC.

  7. Final Design for an International Intercomparison Exercise for Nuclear Accident Dosimetry at the DAF Using Godiva-IV: IER-148 CED-2 Report

    Energy Technology Data Exchange (ETDEWEB)

    Heinrichs, Dave [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Beller, Tim [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Burch, Jennifer [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Cummings, Rick [National Security Technologies, LLC. (NSTec), Mercury, NV (United States) Nevada National Security Site; Duluc, Matthieu [Inst. de Radioprotection et de Sûrete Nucleaire (ISRN), Fontenay-aux-Roses (France); Gadd, Milan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Goda, Joetta [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Hickman, David [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); McAvoy, Doug [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Rathbone, Bruce [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Sullivan, Randy [Savannah River Site (SRS), Aiken, SC (United States); Trompier, Francois [Inst. de Radioprotection et de Sûrete Nucleaire (ISRN), Fontenay-aux-Roses (France); Veinot, Ken [Y-12 National Security Complex, Oak Ridge, TN (United States); Ward, Dann [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Will, Rashelle [National Security Technologies, LLC. (NSTec), Mercury, NV (United States) Nevada National Security Site; Wilson, Chris [Atomic Weapons Establishment (AWE), Berkshire (United Kingdom); Zieziulewicz, Thomas [Knolls Atomic Power Lab. (KAPL), Niskayuna, NY (United States)

    2014-09-30

    This document is the Final Design (CED-2) Report for IER-148, “International Inter-comparison Exercise for Nuclear Accident Dosimetry at the DAF Using Godiva-IV.” The report describes the structure of the exercise consisting of three irradiations; identifies the participating laboratories and their points of contact; provides the details of all dosimetry elements and their placement in proximity to Godiva-IV on support stands or phantoms ; and lists the counting and spectroscopy equipment each laboratory will utilize in the Mercury NAD Lab. The exercise is tentatively scheduled for one week in August 2015.

  8. Godiva and Juliet Diagnostics CED-1 (IER-176)

    Energy Technology Data Exchange (ETDEWEB)

    Scorby, J C

    2011-12-21

    A suite of diagnostics are being proposed for use in the Juliet experiment (IER-128). In order to calibrate and test the diagnostics prior to use, the LLNL calibration facility and Godiva pulsed reactor will be used to provide intense sources of neutrons and gammas. Due to the similarities of the Godiva and Juliet radiation fields, the diagnostics being developed and tested for Juliet can also play an on-going role in diagnostics for Godiva as well as, perhaps, other critical assembly experiments. Similar work is also being conducted for IER-147 for the purpose of characterizing the Godiva radiation field in support of an upcoming international nuclear accident dosimetry exercise. Diagnostics developed and fielded under IER-147 can provide valuable data with respect to the neutron and gamma energy spectrums in the vicinity of Godiva which is relevant to the calibration of Juliet diagnostics.

  9. Generation IV reactors: reactor concepts

    International Nuclear Information System (INIS)

    Cardonnier, J.L.; Dumaz, P.; Antoni, O.; Arnoux, P.; Bergeron, A.; Renault, C.; Rimpault, G.; Delpech, M.; Garnier, J.C.; Anzieu, P.; Francois, G.; Lecomte, M.

    2003-01-01

    Liquid metal reactor concept looks promising because of its hard neutron spectrum. Sodium reactors benefit a large feedback experience in Japan and in France. Lead reactors have serious assets concerning safety but they require a great effort in technological research to overcome the corrosion issue and they lack a leader country to develop this innovative technology. In molten salt reactor concept, salt is both the nuclear fuel and the coolant fluid. The high exit temperature of the primary salt (700 Celsius degrees) allows a high energy efficiency (44%). Furthermore molten salts have interesting specificities concerning the transmutation of actinides: they are almost insensitive to irradiation damage, some salts can dissolve large quantities of actinides and they are compatible with most reprocessing processes based on pyro-chemistry. Supercritical water reactor concept is based on operating temperature and pressure conditions that infers water to be beyond its critical point. In this range water gets some useful characteristics: - boiling crisis is no more possible because liquid and vapour phase can not coexist, - a high heat transfer coefficient due to the low thermal conductivity of supercritical water, and - a high global energy efficiency due to the high temperature of water. Gas-cooled fast reactors combining hard neutron spectrum and closed fuel cycle open the way to a high valorization of natural uranium while minimizing ultimate radioactive wastes and proliferation risks. Very high temperature gas-cooled reactor concept is developed in the prospect of producing hydrogen from no-fossil fuels in large scale. This use implies a reactor producing helium over 1000 Celsius degrees. (A.C.)

  10. Generation IV reactors: economics

    International Nuclear Information System (INIS)

    Dupraz, B.; Bertel, E.

    2003-01-01

    The operating nuclear reactors were built over a short period: no more than 10 years and today their average age rounds 18 years. EDF (French electricity company) plans to renew its reactor park over a far longer period : 30 years from 2020 to 2050. According to EDF this objective implies 3 constraints: 1) a service life of 50 to 60 years for a significant part of the present operating reactors, 2) to be ready to built a generation 3+ unit in 2020 which infers the third constraint: 3) to launch the construction of an EPR (European pressurized reactor) prototype as soon as possible in order to have it operating in 2010. In this scheme, generation 4 reactor will benefit the feedback experience of generation 3 and will take over in 2030. Economic analysis is an important tool that has been used by the generation 4 international forum to select the likely future reactor systems. This analysis is based on 4 independent criteria: the basic construction cost, the construction time, the operation and maintenance costs and the fuel cycle cost. This analysis leads to the evaluation of the global cost of electricity generation and of the total investment required for each of the reactor system. The former defines the economic competitiveness in a de-regulated energy market while the latter is linked to the financial risk taken by the investor. It appears, within the limits of the assumptions and models used, that generation 4 reactors will be characterized by a better competitiveness and an equivalent financial risk when compared with the previous generation. (A.C.)

  11. Generation IV reactors: international projects

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  12. Pursuing public and professional perception of nuclear energy with the Smithsonian Lady Godiva replica and demonstration/display

    International Nuclear Information System (INIS)

    Grundl, J.A.

    1987-01-01

    The following topics were addressed by the speaker at the Fast Burst Reactor Workshop banquet: (1) Lady Godiva and the perception of nuclear energy, (2) the question of public and professional perception of nuclear energy, (3) notions of public perception among Lady Godiva replicators, and (4) the nuclear fission chain: complete embodiment of nuclear energy

  13. Materials for generation-IV nuclear reactors

    International Nuclear Information System (INIS)

    Alvarez, M. G.

    2009-01-01

    Materials science and materials development are key issues for the implementation of innovative reactor systems such as those defined in the framework of the Generation IV. Six systems have been selected for Generation IV consideration: gas-cooled fast reactor, lead-cooled fast reactor, molten salt-cooled reactor, sodium-cooled fast reactor, supercritical water-cooled reactor, and very high temperature reactor. The structural materials need to resist much higher temperatures, higher neutron doses and extremely corrosive environment, which are beyond the experience of the current nuclear power plants. For this reason, the first consideration in the development of Generation-IV concepts is selection and deployment of materials that operate successfully in the aggressive operating environments expected in the Gen-IV concepts. This paper summarizes the Gen-IV operating environments and describes the various candidate materials under consideration for use in different structural applications. (author)

  14. Sandia Pulse Reactor-IV Project

    International Nuclear Information System (INIS)

    Reuscher, J.A.

    1983-01-01

    Sandia National Laboratories has developed, designed and operated fast burst reactors for over 20 years. These reactors have been used for a variety of radiation effects programs. During this period, programs have required larger irradiation volumes primarily to expose complex electronic systems to postulated threat environments. As experiment volumes increased, a new reactor was built so that these components could be tested. The Sandia Pulse Reactor-IV is a logical evolution of the two decades of fast burst reactor development at Sandia

  15. Generation-IV nuclear reactors, SFR concept

    International Nuclear Information System (INIS)

    Dufour, P.

    2010-01-01

    In this presentation author deals with development of sodium-cooled fast reactors and lead-cooled fast reactors. He concluded that: - SFR is a proved concept that has never achieved industrial deployment; - The GEN IV objectives need to reconsider the design of both the core and the reactor design : innovations are being analysed; Future design will benefit from considerable feedback of design, licensing, construction and operation of PX, SPX, etc.

  16. Reactor physics challenges in GEN-IV reactor design

    International Nuclear Information System (INIS)

    Driscoll, Michael K.; Hejzlar, Pavel

    2005-01-01

    An overview of the reactor physics aspects of GENeration Four (GEN-IV) advanced reactors is presented, emphasizing how their special requirements for enhanced sustainability, safety and economics motivates consideration of features not thoroughly analyzed in the past. The resulting concept-specific requirements for better data and methods are surveyed, and some approaches and initiatives are suggested to meet the challenges faced by the international reactor physics community. No unresolvable impediments to successful development of any of the six major types of proposed reactors are identified, given appropriate and timely devotion of resources

  17. Reactor physics challenges in GEN-IV reactor design

    Energy Technology Data Exchange (ETDEWEB)

    Driscoll, Michael K.; Hejzlar, Pavel [Massachusetts Institute of Technology, MA (United States)

    2005-02-15

    An overview of the reactor physics aspects of GENeration Four (GEN-IV) advanced reactors is presented, emphasizing how their special requirements for enhanced sustainability, safety and economics motivates consideration of features not thoroughly analyzed in the past. The resulting concept-specific requirements for better data and methods are surveyed, and some approaches and initiatives are suggested to meet the challenges faced by the international reactor physics community. No unresolvable impediments to successful development of any of the six major types of proposed reactors are identified, given appropriate and timely devotion of resources.

  18. ASN’s actions in GEN IV reactors and Sodium Fast Reactors (SFR)

    International Nuclear Information System (INIS)

    Belot, Clotilde

    2013-01-01

    The ASN is involved in 3 actions concerning GEN IV: • Overview of nuclear reactor GEN IV systems; • Specific analysis about transmutation; • Prototype reactor ASTRID (SFR). Furthermore theses actions are in the beginning (no conclusions or results available)

  19. Transient Analysis of Generation IV quick reactors; Analisis de Transitorios en Reactores Rapidos de Generacion IV

    Energy Technology Data Exchange (ETDEWEB)

    Vazquez, M.; Martin-Fuertes, F.

    2013-07-01

    As a complement to the attached code 3D neutron-CIEMAT thermohydraulic added a module to simulate transient. Temporary kinetics is resolved by factoring flow in a spatial part and another storm. MCNP provides the reactivity and updated spatial function and COBRA-IV calculates the temperature distribution. Temporary dependence of amplitude is calculated using time delayed neutron Kinetic equations. As an example of application, examines a transient loss of flow in MYRRHA, a lead-cooled experimental reactor.

  20. Experimental facilities for Generation IV reactors research

    International Nuclear Information System (INIS)

    Krecanova, E.; Di Gabriele, F.; Berka, J.; Zychova, M.; Macak, J.; Vojacek, A.

    2013-06-01

    Centrum Vyzkumu Rez (CVR) is research and development Company situated in Czech Republic and member of the UJV group. One of its major fields is material research for Generation IV reactor concepts, especially supercritical water-cooled reactor (SCWR), very high temperature/gas-cooled fast reactor (VHTR/GFR) and lead-cooled fast reactor (LFR). The CVR is equipped by and is building unique experimental facilities which simulate the environment in the active zones of these reactor concepts and enable to pre-qualify and to select proper constructional materials for the most stressed components of the facility (cladding, vessel, piping). New infrastructure is founded within the Sustainable Energy project focused on implementation the Generation IV and fusion experimental facilities. The research of SCWR concept is divided to research and development of the constructional materials ensured by SuperCritical Water Loop (SCWL) and fuel components research on Fuel Qualification Test loop (SCWL-FQT). SCWL provides environment of the primary circuits of European SCWR, pressure 25 MPa, temperature 600 deg. C and its major purpose is to simulate behavior of the primary medium and candidate constructional materials. On-line monitoring system is included to collect the operational data relevant to experiment and its evaluation (pH, conductivity, chemical species concentration). SCWL-FQT is facility focused on the behavior of cladding material and fuel at the conditions of so-called preheater, the first pass of the medium through the fuel (in case of European SCWR concept). The conditions are 450 deg. C and 25 MPa. SCWL-FQT is unique facility enabling research of the shortened fuel rods. VHTR/GFR research covers material testing and also cleaning methods of the medium in primary circuit. The High Temperature Helium Loop (HTHL) enables exposure of materials and simulates the VHTR/GFR core environment to analyze the behavior of medium, especially in presence of organic compounds and

  1. Fluidized bed nuclear reactor as a IV generation reactor

    International Nuclear Information System (INIS)

    Sefidvash, Farhang

    2002-01-01

    The object of this paper is to analyze the characteristics of the Fluidized Bed Nuclear Reactor (FBNR) concept under the light of the requirements set for the IV generation nuclear reactors. It is seen that FBNR generally meets the goals of providing sustainable energy generation that meets clean air objectives and promotes long-term availability of systems and effective fuel utilization for worldwide energy production; minimize and manage their nuclear waste and notably reduce the long term stewardship burden in the future, thereby improving protection for the public health and the environment; increase the assurance that it is a very unattractive and least desirable route for diversion or theft of weapons-usable materials; excel in safety and reliability; have a very low likelihood and degree of reactor core damage; eliminate the need for offsite emergency response; have a clear life-cycle cost advantage over other energy sources; have a level of financial risk comparable to other energy projects. The other advantages of the proposed design are being modular, low environmental impact, exclusion of severe accidents, short construction period, flexible adaptation to demand, excellent load following characteristics, and competitive economics. (author)

  2. Transient Analysis Needs for Generation IV Reactor Concepts

    International Nuclear Information System (INIS)

    Siefken, L.J.; Harvego, E.A.; Coryell, E.W.; Davis, C.B.

    2002-01-01

    The importance of nuclear energy as a vital and strategic resource in the U. S. and world's energy supply mix has led to an initiative, termed Generation IV by the U.S. Department of Energy (DOE), to develop and demonstrate new and improved reactor technologies. These new Generation IV reactor concepts are expected to be substantially improved over the current generation of reactors with respect to economics, safety, proliferation resistance and waste characteristics. Although a number of light water reactor concepts have been proposed as Generation IV candidates, the majority of proposed designs have fundamentally different characteristics than the current generation of commercial LWRs operating in the U.S. and other countries. This paper presents the results of a review of these new reactor technologies and defines the transient analyses required to support the evaluation and future development of the Generation IV concepts. The ultimate objective of this work is to identify and develop new capabilities needed by INEEL to support DOE's Generation IV initiative. In particular, the focus of this study is on needed extensions or enhancements to SCDAP/RELAP5/3D code. This code and the RELAP5-3D code from which it evolved are the primary analysis tools used by the INEEL and others for the analysis of design-basis and beyond-design-basis accidents in current generation light water reactors. (authors)

  3. Policy-induced market introduction of Generation IV reactor systems

    International Nuclear Information System (INIS)

    Heek, Aliki Irina van; Roelofs, Ferry

    2011-01-01

    Almost 10 years ago the U.S. Department of Energy (DOE) started the Generation IV Initiative (GenIV) with 9 other national governments with a positive ground attitude towards nuclear energy. Some of these Generation IV systems, like the fast reactors, are nearing the demonstration stage. The question on how their market introduction will be implemented becomes increasingly urgent. One main topic for future reactor technologies is the treatment of radioactive waste products. Technological solutions to this issue are being developed. One possible process is the transformation of long-living radioactive nuclides into short living ones; a process known as transmutation, which can be done in a nuclear reactor only. Various Generation IV reactor concepts are suitable for this process, and of these systems most experience has been gained with the sodium-cooled fast reactor (SFR). However, both these first generation SFR plants and their Generation IV successors are designed as electricity generating plants, and therefore supposed to be commercially viable in the electricity markets. Various studies indicate that the generation costs of a combined LWR-(S)FR nuclear generating park (LWR: light water reactor) will be higher than that of an LWR-only park. To investigate the effects of the deployment of the different reactors and fuel cycles on the waste produced, resources used and costs incurred as a function of time, a dynamic fuel cycle assessment is performed. This study will focus on the waste impact of the introduction of a fraction of fast reactors in the European nuclear reactor park with a cost increase as described in the previous paragraph. The nuclear fuel cycle scenario code DANESS is used for this, as well as the nuclear park model of the EU-27 used for the previous study. (orig.)

  4. Building competencies for New Gen IV Reactors

    International Nuclear Information System (INIS)

    Pavel, G.L.; Ghitescu, P.

    2015-01-01

    The Advanced Lead Fast Reactor European Demonstrator - ALFRED is designed and sustained by several European countries. It is a 300 MWt (125 MWe) reactor, intended to be built in Romania, near the Pitesti site. Pure lead is used as primary coolant and it is foreseen to have a 40% thermal efficiency. Secondary cycle contains superheated water steam at around 450 Celsius degrees. Through ARCADIA cooperation, 26 partners from all over Europe joined their forces to provide the necessary research support for ALFRED. In Romania, several entities are providing nuclear courses but only the University Politechnica of Bucharest is offering a complete training program for nuclear industry but targeted courses for LFR technology need to be developed and implemented. Issues like physics of breeding, coolant analysis and behavior, targeted computer codes, core design and dynamics, safety still needs to be tackled

  5. CANDU technology for generation III + AND IV reactors

    International Nuclear Information System (INIS)

    Torgerson, D.F.

    2005-01-01

    Atomic Energy of Canada Limited (AECL) is the original developer of the CANDU?reactor, one of the three major commercial power reactor designs now used throughout the world. For over 60 years, AECL has continued to evolve the CANDU design from the CANDU prototypes in the 1950s and 1960s through to the second generation reactors now in operation, including the Generation II+ CANDU 6. The next phase of this evolution, the Generation III+ Advanced CANDU ReactorTM (ACRTM), continues the strategy of basing next generation technology on existing CANDU reactors. Beyond the ACR, AECL is developing the Generation IV CANDU Super Critical Water Reactor. Owing to the evolutionary nature of these advanced reactors, advanced technology from the development programs is also being applied to operating CANDU plants, for both refurbishments and upgrading of existing systems and components. In addition, AECL is developing advanced technology that covers the entire life cycle of the CANDU plant, including waste management and decommissioning. Thus, AECL maintains state-of-the-art expertise and technology to support both operating and future CANDU plants. This paper outlines the scale of the current core knowledge base that is the foundation for advancement and support of CANDU technology. The knowledge base includes advancements in materials, fuel, safety, plant operations, components and systems, environmental technology, waste management, and construction. Our approach in each of these areas is to develop the underlying science, carry out integrated engineering scale tests, and perform large-scale demonstration testing. AECL has comprehensive R and D and engineering development programs to cover all of these elements. The paper will show how the ongoing expansion of the CANDU knowledge base has led to the development of the Advanced CANDU Reactor. The ACR is a Generation III+ reactor with substantially reduced costs, faster construction, and enhanced passive safety and operating

  6. Generation IV nuclear reactors: Current status and future prospects

    International Nuclear Information System (INIS)

    Locatelli, Giorgio; Mancini, Mauro; Todeschini, Nicola

    2013-01-01

    Generation IV nuclear power plants (GEN IV NPPs) are supposed to become, in many countries, an important source of base load power in the middle–long term (2030–2050). Nowadays there are many designs of these NPPs but for political, strategic and economic reasons only few of them will be deployed. International literature proposes many papers and reports dealing with GEN IV NPPs, but there is an evident difference in the types and structures of the information and a general unbiased overview is missing. This paper fills the gap, presenting the state-of-the-art for GEN IV NPPs technologies (VHTR, SFR, SCWR, GFR, LFR and MSR) providing a comprehensive literature review of the different designs, discussing the major R and D challenges and comparing them with other advanced technologies available for the middle- and long-term energy market. The result of this research shows that the possible applications for GEN IV technologies are wider than current NPPs. The economics of some GEN IV NPPs is similar to actual NPPs but the “carbon cost” for fossil-fired power plants would increase the relative valuation. However, GEN IV NPPs still require substantial R and D effort, preventing short-term commercial adoption. - Highlights: • Generation IV reactors are the middle–long term technology for nuclear energy. • This paper provides an overview and a taxonomy for the designs under consideration. • R and D efforts are in the material, heat exchangers, power conversion unit and fuel. • The life cycle costs are competitive with other innovative technologies. • The hydrogen economy will foster the development of Generation IV reactors

  7. Overview of Generation IV (Gen IV) Reactor Designs - Safety and Radiological Protection Considerations

    International Nuclear Information System (INIS)

    Baudrand, Olivier; Blanc, Daniel; Ivanov, Evgeny; Bonneville, Herve; Clement, Bernard; Kissane, Martin; Meignen, Renaud; Monhardt, Daniel; Nicaise, Gregory; Bourgois, Thierry; Bruna, Giovanni; Hache, Georges; Repussard, Jacques

    2012-01-01

    The purpose of this document is to provide an updated overview of specific safety and radiological protection issues for all the reactor concepts adopted by the GIF (Generation IV International Forum), independent of their advantages or disadvantages in terms of resource optimization or long-lived-waste reduction. In particular, this new document attempts to bring out the advantages and disadvantages of each concept in terms of safety, taking into account the Western European Nuclear Regulators' Association (WENRA) statement concerning safety objectives for new nuclear power plants. Using an identical framework for each reactor concept (sodium-cooled fast reactors or SFR, high / very-high temperature helium-cooled reactors of V/HTR, gas-cooled fast reactors or GFR, lead-or lead / bismuth-cooled fast reactors or LFR, molten salt reactors or MSR, and supercritical-water-cooled reactors or SCWR), this summary report provides some general conclusions regarding their safety and radiological protection issues, inspired by WENRA's safety objectives and on the basis of available information. Initial lessons drawn from the events at the Fukushima-Daiichi nuclear power plant in March 2011 have also been taken into account in IRSN's analysis of each reactor concept

  8. Nordic Nuclear Materials Forum for Generation IV Reactors

    International Nuclear Information System (INIS)

    Anghel, C.; Penttilae, S.

    2010-03-01

    A network for material issues for Generation IV nuclear power has been initiated within the Nordic countries. The objectives of the Generation IV Nordic Nuclear Materials Forum (NOMAGE4) are to put the basis of a sustainable forum for Gen IV issues, especially focussing on fuels, cladding, structural materials and coolant interaction. Other issues include reactor physics, dynamics and diagnostics, core and fuel design. The present report summarizes the work performed during the year 2009. The efforts made include identification of organisations involved in Gen IV issues in the Nordic countries, update of the forum website, http://www.studsvik.se/GenerationIV, and investigation of capabilities for research within the area of Gen IV. Within the NOMAGE4 project a seminar on Generation IV Nuclear Energy Systems has been organized during 15-16th of October 2009. The aim of the seminar was to provide a forum for exchange of information, discussion on future research needs and networking of experts on Generation IV reactor concepts. As an outcome of the NOMAGE4, a few collaboration project proposals have been prepared/planned in 2009. The network was welcomed by the European Commission and was mentioned as an exemplary network with representatives from industries, universities, power companies and research institutes. NOMAGE4 has been invited to participate to the 'European Energy Research Alliance, EERA, workshop for nuclear structural materials' http://www.eera-set.eu/index.php?index=41 as external observers. Future plans include a new Nordic application for continuation of NOMAGE4 network. (author)

  9. Nordic Nuclear Materials Forum for Generation IV Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Anghel, C. (Studsvik Nuclear AB, Nykoeping (Sweden)); Penttilae, S. (Technical Research Centre of Finland, VTT (Finland))

    2010-03-15

    A network for material issues for Generation IV nuclear power has been initiated within the Nordic countries. The objectives of the Generation IV Nordic Nuclear Materials Forum (NOMAGE4) are to put the basis of a sustainable forum for Gen IV issues, especially focussing on fuels, cladding, structural materials and coolant interaction. Other issues include reactor physics, dynamics and diagnostics, core and fuel design. The present report summarizes the work performed during the year 2009. The efforts made include identification of organisations involved in Gen IV issues in the Nordic countries, update of the forum website, http://www.studsvik.se/GenerationIV, and investigation of capabilities for research within the area of Gen IV. Within the NOMAGE4 project a seminar on Generation IV Nuclear Energy Systems has been organized during 15-16th of October 2009. The aim of the seminar was to provide a forum for exchange of information, discussion on future research needs and networking of experts on Generation IV reactor concepts. As an outcome of the NOMAGE4, a few collaboration project proposals have been prepared/planned in 2009. The network was welcomed by the European Commission and was mentioned as an exemplary network with representatives from industries, universities, power companies and research institutes. NOMAGE4 has been invited to participate to the 'European Energy Research Alliance, EERA, workshop for nuclear structural materials' http://www.eera-set.eu/index.php?index=41 as external observers. Future plans include a new Nordic application for continuation of NOMAGE4 network. (author)

  10. The generation IV nuclear reactor systems - Energy of future

    International Nuclear Information System (INIS)

    Ohai, Dumitru; Jianu, Adrian

    2006-01-01

    Ten nations joined within the Generation IV International Forum (GIF), agreeing on a framework for international cooperation in research. Their goal is to develop future-generation nuclear energy systems that can be licensed, constructed, and operated in an economically competitive way while addressing the issues of safety, proliferation, and other public perception concerns. The objective is for the Gen IV systems to be available for deployment by 2030. Using more than 100 nuclear experts from its 10 member nations, the GIF has developed a Gen IV Technology Roadmap to guide the research and development of the world's most advanced, efficient and safe nuclear power systems. The Gen IV Technology Roadmap calls for extensive research and development of six different potential future reactor systems. These include water-cooled, gas-cooled, liquid metal-cooled and nonclassical systems. One or more of these reactor systems will provide the best combination of safety, reliability, efficiency and proliferation resistance at a competitive cost. The main goals for the Gen IV Nuclear Energy Systems are: - Provide sustainable energy generation that meets clean air objectives and promotes long-term availability of systems and effective fuel use for worldwide energy production; - Minimize and manage their nuclear waste and noticeably reduce the long-term stewardship burden in the future, improving the protection of public health and the environment; - Increase the assurance that these reactors are very unattractive and the least desirable route for diversion or theft of weapons-usable materials, and provide increased protection against acts of terrorism; - Have a clear life-cycle cost advantage over other energy sources; - Have a level of financial risk comparable to other energy projects; - Excel in safety and reliability; - Have a low likelihood and degree of reactor core damage. (authors)

  11. GEN IV reactors: Where we are, where we should go

    International Nuclear Information System (INIS)

    Locatelli, G.; Mancini, M.; Todeschini, N.

    2012-01-01

    GEN IV power plants represent the mid-long term option of the nuclear sector. International literature proposes many papers and reports dealing with these reactors, but there is an evident difference of type and shape of information making impossible each kind of detailed comparison. Moreover, authors are often strongly involved in some particular design; this creates many difficulties in their super-partes position. Therefore it is necessary to put order in the most relevant information to understand strengths and weaknesses of each design and derive an overview useful for technicians and policy makers. This paper presents the state-of the art for GEN IV nuclear reactors providing a comprehensive literature review of the different designs with a relate taxonomy. It presents the more relevant references, data, advantages, disadvantages and barriers to the adoptions. In order to promote an efficient and wide adoption of GEN IV reactors the paper provides the pre-conditions that must be accomplished, enabling factors promoting the implementation and barriers limiting the extent and intensity of its implementation. It concludes outlying the state of the art of the most important R and D areas and the future achievements that must be accomplished for a wide adoption of these technologies. (authors)

  12. ASTRID, Generation IV advanced sodium technological reactor for industrial demonstration

    International Nuclear Information System (INIS)

    Gauche, F.

    2013-01-01

    ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) is an integrated technology demonstrator designed to demonstrate the operability of the innovative choices enabling fast neutron reactor technology to meet the Generation IV criteria. ASTRID is a sodium-cooled fast reactor with an electricity generating power of 600 MWe. In order to meet the generation IV goals, ASTRID will incorporate the following decisive innovations: -) an improved core with a very low, even negative void coefficient; -) the possible installation of additional safety devices in the core. For example, passive anti-reactivity insertion devices are explored; -) more core instrumentation; -) an energy conversion system with modular steam generators, to limit the effects of a possible sodium-water reaction, or sodium-nitrogen exchangers; -) considerable thermal inertia combined with natural convection to deal with decay heat; -)elimination of major sodium fires by bunkerization and/or inert atmosphere in the premises; -) to take into account off-site hazards (earthquake, airplane crash,...) right from the design stage; -) a complete rethink of the reactor architecture in order to limit the risk of proliferation. ASTRID will also include systems for reducing the length of refueling outages and increasing the burn-up and the duration of the cycle. In-service inspection, maintenance and repair are also taken into account right from the start of the project. The ASTRID prototype should be operational by about 2023. (A.C.)

  13. Comparative analysis of power conversion cycles optimized for fast reactors of generation IV; Analisis comparativo de ciclos de conversion de potencia optimizados para reactores rapidos de generacion IV

    Energy Technology Data Exchange (ETDEWEB)

    Perez Pichel, G. D.

    2011-07-01

    For the study, which is presented here, has been chosen as the specific parameters of each reactor, which are today the three largest projects within generation IV technology development: ESFR for the reactor's sodium, LEADER for the lead reactor's and finally, GoFastR in the case of reactor gas-cooled.

  14. Comparative analysis of power conversion cycles optimized for fast reactors of generation IV

    International Nuclear Information System (INIS)

    Perez Pichel, G. D.

    2011-01-01

    For the study, which is presented here, has been chosen as the specific parameters of each reactor, which are today the three largest projects within generation IV technology development: ESFR for the reactor's sodium, LEADER for the lead reactor's and finally, GoFastR in the case of reactor gas-cooled.

  15. Pebble bed modular reactor - The first Generation IV reactor to be constructed

    International Nuclear Information System (INIS)

    Ion, S.; Nicholls, D.; Matzie, R.; Matzner, D.

    2004-01-01

    Substantial interest has been generated in advanced reactors over the past few years. This interest is motivated by the view that new nuclear power reactors will be needed to provide low carbon generation of electricity and possibly hydrogen to support the future growth in demand for both of these commodities. Some governments feel that substantially different designs will be needed to satisfy the desires for public perception, improved safety, proliferation resistance, reduced waste and competitive economics. This has motivated the creation of the Generation IV Nuclear Energy Systems programme in which ten countries have agreed on a framework for international cooperation in research for advanced reactors. Six designs have been selected for continued evaluation, with the objective of deployment by 2030. One of these designs is the very high temperature reactor (VHTR), which is a thermal neutron spectrum system with a helium-cooled core utilising carbon-based fuel. The pebble bed modular reactor (PBMR), being developed in South Africa through a worldwide international collaborative effort led by Eskom, the national utility, will represent a key milestone on the way to achievement of the VHTR design objectives, but in the much nearer term. This paper outlines the design objectives, safety approach and design details of the PBMR, which is already at a very advanced stage of development. (author)

  16. Sargent-IV Project. Development of new methodologies for safety analysis of Generation IV reactors; Proyecto SARGEB-IV. Desarrollo de nuevas metodologias de analisis de seguridad para reactores de Generacion IV

    Energy Technology Data Exchange (ETDEWEB)

    Queral, C.; Gallego, E.; Jimenez, G.

    2013-07-01

    The main result of this paper is the proposal for the addition of new ingredients in the safety analysis methodologies for Generation-IV reactors that integrates the features of probabilistic safety analysis within deterministic. This ensures a higher degree of integration between the classical deterministic and probabilistic methodologies.

  17. European project SARGEN IV: safety approach and assessment of GEN IV reactors

    International Nuclear Information System (INIS)

    Ammirabile, L.

    2013-01-01

    • SARGEN I V has elaborated a proposal for the harmonization of safety assessment practices for GEN IV NPP. • An overall reinforcement of DiD is expected for GEN I V NPP, including improved independence between all levels of DiD. • An inherent approach should reinforce the fulfillment of fundamental safety functions e.g. the consequences for some situations should be reduced and the grace periods should be extended. For the same reason, the use of passive systems can be envisaged. • The need of complementary and integrated deterministic and probabilistic approaches is reiterated. • Methodologies: Some of them are not yet applied. • Assessment of hazards would be a challenging aspect of next generation of NPP safety assessment and should be improved, which is confirmed by the first insights of Fukushima Daiichi TEPCO reactors accidents. • Provisions to cope with extreme events notably to improve the grace period before cliff-edge effects and thus allowing back-up measures to be implemented have to be defined and should be considered as hardened equipments

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-08-01

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

  19. Overview of Generation IV (Gen IV) Reactor Designs - Safety and Radiological Protection Considerations. Published on September 24, 2012

    International Nuclear Information System (INIS)

    Couturier, Jean; Bruna, Giovanni; Baudrand, Olivier; Blanc, Daniel; Ivanov, Evgeny; Bonneville, Herve; Clement, Bernard; Kissane, Martin; Meignen, Renaud; Monhardt, Daniel; Nicaise, Gregory; Bourgois, Thierry; Hache, Georges

    2012-01-01

    The purpose of this document is to provide an updated overview of specific safety and radiological protection issues for all the reactor concepts adopted by the GIF (Generation IV International Forum), independent of their advantages or disadvantages in terms of resource optimization or long-lived-waste reduction. In particular, this new document attempts to bring out the advantages and disadvantages of each concept in terms of safety, taking into account the Western European Nuclear Regulators' Association (WENRA) statement concerning safety objectives for new nuclear power plants. Using an identical framework for each reactor concept (sodium-cooled fast reactors or SFR, high / very-high temperature helium-cooled reactors of V/HTR, gas-cooled fast reactors or GFR, lead-or lead / bismuth-cooled fast reactors or LFR, molten salt reactors or MSR, and supercritical-water-cooled reactors or SCWR), this summary report provides some general conclusions regarding their safety and radiological protection issues, inspired by WENRA's safety objectives and on the basis of available information. Initial lessons drawn from the events at the Fukushima-Daiichi nuclear power plant in March 2011 have also been taken into account in IRSN's analysis of each reactor concept

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  1. Technological studies for obtaining lead oxide compacts used in generation IV nuclear reactors

    International Nuclear Information System (INIS)

    Paraschiv, I.; Benga, D.

    2016-01-01

    One of the main concerns of the nuclear research at this moment is the development of the necessary technologies for Generation IV reactors. The main candidate as coolant agent in these reactors is molten lead but this material involves ensuring the oxygen control, due to potential contamination of coolant through the formation of solid oxides and the influence on the corrosion rate of structural parts and for this reason, the oxygen concentration must be kept in a well specified domain. One of the proposed methods for oxygen monitoring and control in the technology of Generation IV reactors, is the use of PbO compacts. For this paper technological tests were performed for developing and setting the optimal parameters in order to attain lead oxide compacts necessary for the oxygen control technology in Generation IV nuclear reactors. (authors)

  2. Light water reactor fuel analysis code FEMAXI-IV(Ver.2). Detailed structure and user's manual

    International Nuclear Information System (INIS)

    Suzuki, Motoe; Saitou, Hiroaki.

    1997-11-01

    A light water reactor fuel behavior analysis code FEMAXI-IV(Ver.2) was developed as an improved version of FEMAXI-IV. Development of FEMAXI-IV has been already finished in 1992, though a detailed structure and input manual of the code have not been open to users yet. Here, the basic theories and structure, the models and numerical solutions applied to FEMAXI-IV(Ver.2), and the material properties adopted in the code are described in detail. In FEMAXI-IV(Ver.2), programming bugs in previous FEMAXI-IV were eliminated, renewal of the pellet thermal conductivity was performed, and a model of thermal-stress restraint on FP gas release was incorporated. For facilitation of effective and wide-ranging application of the code, methods of input/output of the code are also described in detail, and sample output is included. (author)

  3. A Virtual Reality Framework to Optimize Design, Operation and Refueling of GEN-IV Reactors

    International Nuclear Information System (INIS)

    Rizwan-uddin; Nick Karancevic; Stefano Markidis; Joel Dixon; Cheng Luo; Jared Reynolds

    2008-01-01

    Many GEN-IV candidate designs are currently under investigation. Technical issues related to material, safety and economics are being addressed at research laboratories, industry and in academia. After safety, economic feasibility is likely to be the most important criterion in the success of GEN-IV design(s). Lessons learned from the designers and operators of GEN-II (and GEN-III) reactors must play a vital role in achieving both safety and economic feasibility goals

  4. A Virtual Reality Framework to Optimize Design, Operation and Refueling of GEN-IV Reactors.

    Energy Technology Data Exchange (ETDEWEB)

    Rizwan-uddin; Nick Karancevic; Stefano Markidis; Joel Dixon; Cheng Luo; Jared Reynolds

    2008-04-23

    many GEN-IV candidate designs are currently under investigation. Technical issues related to material, safety and economics are being addressed at research laboratories, industry and in academia. After safety, economic feasibility is likely to be the most important crterion in the success of GEN-IV design(s). Lessons learned from the designers and operators of GEN-II (and GEN-III) reactors must play a vital role in achieving both safety and economic feasibility goals.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-12-01

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

  6. Theoretical analysis of nuclear reactors (Phase I), I-V, Part IV, Nuclear fuel depletion

    International Nuclear Information System (INIS)

    Pop-Jordanov, J.

    1962-07-01

    Nuclear fuel depletion is analyzed in order to estimate the qualitative and quantitative fuel property changes during irradiation and the influence of changes on the reactivity during long-term reactor operation. The changes of fuel properties are described by changes of neutron absorption and fission cross sections. Part one of this report covers the economic significance of fuel burnup and the review of fuel isotopic changes during depletion. Pat two contains the analysis of the U 235 chain, analytical expressions for the concentrations of U 235 , U 236 and Np 237 as a function of burnup. Part three contains the analysis of neutron spectrum influence on the Westcott method for calculating the cross sections. Part four contains the calculation method applied on Calder Hall type reactor. The results were obtained by applying ZUSE-22 R digital computer

  7. IGORR-IV: Proceedings of the fourth meeting of the International Group On Research Reactors

    International Nuclear Information System (INIS)

    Rosenbalm, K.F.

    1995-01-01

    The fourth meeting of the International Group on Research Reactors (IGORR-IV) was attended by was good 55 registered participants from 28 organizations in 13 countries, which compares well with the previous meetings. Twenty-nine papers were presented in five sessions over the two-day meeting. Session subjects were: Operating Research Reactors and Facility Upgrades; Research Reactors in Desin and Construction; Research, Development, and Analysis Results of Thermal Hydraulic Calculations, U 3 Si 2 Fuel Performance and Faibrication; Structural Materials Performance; Neutronics; Severe Accident analysis. Written versions of the papers or hard copies of the viewgraphs used are published in these Proceedings

  8. IGORR-IV: Proceedings of the fourth meeting of the International Group On Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rosenbalm, K F [comp.

    1995-07-01

    The fourth meeting of the International Group on Research Reactors (IGORR-IV) was attended by was good 55 registered participants from 28 organizations in 13 countries, which compares well with the previous meetings. Twenty-nine papers were presented in five sessions over the two-day meeting. Session subjects were: Operating Research Reactors and Facility Upgrades; Research Reactors in Desin and Construction; Research, Development, and Analysis Results of Thermal Hydraulic Calculations, U{sub 3}Si{sub 2} Fuel Performance and Faibrication; Structural Materials Performance; Neutronics; Severe Accident analysis. Written versions of the papers or hard copies of the viewgraphs used are published in these Proceedings.

  9. Perspective steels for generation IV and fusion reactors

    International Nuclear Information System (INIS)

    Bartosva, I.; Cizek, J.

    2013-01-01

    In this study we focus on the F/M steel Eurofer, the European candidate material for the future fusion reactor and for the strengthening we consider oxides of yttrium. The oxides of yttrium and complex yttrium titanium oxides reinforce the material by forming more or less stable obstacles to dislocations, and by promoting grain refinement by pinning grain boundaries. It appears that part of the yttrium titanium oxides particles dissolves from about 600 grad C while pure Yttria particles are stable at least to 1000 grad C in the steel. The aims of this study are the following: 1) Prove the positive effect of strengthening by yttrium oxides. 2) Measure the hardness of base Eurofer and ODS version by Vickers hardness test (HV). 3) Investigate the behaviour of steels at different annealing temperatures and the changes in strength. 4) Assess defects in microstructure by Coincidence Doppler Broadening (CDB) and Positron Annihilation Lifetime Spectroscopy (PALS) at chosen annealing temperatures. (authors)

  10. 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

  11. The ARIES-II and ARIES-IV second-stability tokamak reactors

    International Nuclear Information System (INIS)

    Najmabadi, F.; Conn, R.W.; Hasan, M.Z.; Mau, T.-K.; Sharafat, S.; Baxi, C.B.; Leuer, J.A.; McQuillan, B.W.; Puhn, F.A.; Schultz, K.R.; Wong, C.P.C.; Brooks, J.; Ehst, D.A.; Hassanein, A.; Hua, T.; Hull, A.; Mattis, R.; Picologlou, B.; Sze, D.-K.; Dolan, T.J.; Herring, J.S.; Bathke, C.G.; Krakowski, R.A.; Werley, K.A.; Bromberg, L.; Schultz, J.; Davis, F.; Holmes, J.A.; Lousteau, D.C.; Strickler, D.J.; Jardin, S.C.; Kessel, C.; Snead, L.; Steiner, D.; Valenti, M.; El-Guebaly, L.A.; Emmert, G.A.; Khater, H.Y.; Santarius, J.F.; Sawan, M.; Sviatoslavsky, I.N.; Cheng, E.T.

    1992-01-01

    The ARIES research program is a multi-institutional effort to develop several visions of tokamak reactors with enhanced economic, safety, and environmental features. Four ARIES visions are currently planned for the ARIES program. The ARIES-I design is a DT-burning reactor based on modest extrapolations from the present tokamak physics database and relies on either existing technology or technology for which trends are already in place, often in programs outside fusion. The ARIES-III study focuses on the potential of tokamaks to operate with D- 3 He fuel system as an alternative to deuterium and tritium. The ARIES-II and ARIES-IV designs have the same fusion plasma but different fusion-power-core designs. The ARIES-II reactor uses liquid lithium as the coolant and tritium breeder and vanadium alloy as the structural material in order to study the potential of low-activation metallic blankets. The ARIES-IV reactor uses helium as the coolant, a solid tritium-breeding material, and silicon carbide composite as the structural material in order to achieve the safety and environmental characteristic of fusion. In this paper the authors describe the trade-off leading to the optimum regime of operation for the ARIES-II and ARIES-IV second-stability reactors and review the engineering design of the fusion power cores

  12. Overview of nuclear safety activities performed by JRC-IE on Gen IV fast reactor concepts

    Energy Technology Data Exchange (ETDEWEB)

    Tsige-Tamirat, H.; Ammirabile, L.; D' Agata, E.; Fuetterer, M.; Ranguelova, V. [European Commission, Joint Research Centre, Institute for Energy, Westerduinweg 3, 1755LE Petten (Netherlands)

    2010-07-01

    The European Strategic Energy Technology (SET) Plan recognizes the need to develop new energy technologies, in order to reduce greenhouse gas emissions and secure energy supply in Europe. Besides renewable energy and improved energy efficiency, a new generation of nuclear power plants and innovative nuclear power applications can play a significant role to achieve this goal. The JRC Institute for Energy 'Safety of Future Nuclear Reactors' (SFNR) Unit is engaged in experimental research, numerical simulation and modelling, scientific, feasibility and engineering studies on innovative nuclear reactor systems. This also represents a significant EURATOM contribution to the Generation IV International Forum. Its activities deal with, among others, the performance assessment of innovative fuels and materials, development of new reactor core concepts and safety solutions, and knowledge management and preservation. Special attention is given to fast reactor concepts, namely the sodium (SFR) and lead (LFR) cooled reactors. Recognizing the maturity of the SFR technology, the European Sustainable Nuclear Energy Technology Platform (SNETP) considers a prototype SFR to be built as a next-step towards the deployment of a first-of-a-kind Gen IV SFR. This paper gives an overview of current research preformed at JRC-IE with emphasis on the work performed in the Collaborative Project on European Sodium Fast Reactor (CP-ESFR) within the European Commission's Seventh Framework Program. (authors)

  13. Overview of nuclear safety activities performed by JRC-IE on Gen IV fast reactor concepts

    International Nuclear Information System (INIS)

    Tsige-Tamirat, H.; Ammirabile, L.; D'Agata, E.; Fuetterer, M.; Ranguelova, V.

    2010-01-01

    The European Strategic Energy Technology (SET) Plan recognizes the need to develop new energy technologies, in order to reduce greenhouse gas emissions and secure energy supply in Europe. Besides renewable energy and improved energy efficiency, a new generation of nuclear power plants and innovative nuclear power applications can play a significant role to achieve this goal. The JRC Institute for Energy 'Safety of Future Nuclear Reactors' (SFNR) Unit is engaged in experimental research, numerical simulation and modelling, scientific, feasibility and engineering studies on innovative nuclear reactor systems. This also represents a significant EURATOM contribution to the Generation IV International Forum. Its activities deal with, among others, the performance assessment of innovative fuels and materials, development of new reactor core concepts and safety solutions, and knowledge management and preservation. Special attention is given to fast reactor concepts, namely the sodium (SFR) and lead (LFR) cooled reactors. Recognizing the maturity of the SFR technology, the European Sustainable Nuclear Energy Technology Platform (SNETP) considers a prototype SFR to be built as a next-step towards the deployment of a first-of-a-kind Gen IV SFR. This paper gives an overview of current research preformed at JRC-IE with emphasis on the work performed in the Collaborative Project on European Sodium Fast Reactor (CP-ESFR) within the European Commission's Seventh Framework Program. (authors)

  14. IRIS - Generation IV Advanced Light Water Reactor for Countries with Small and Medium Electricity Grids

    International Nuclear Information System (INIS)

    Carelli, M. D.

    2002-01-01

    An international consortium of industry, laboratory, university and utility establishments, led by Westinghouse, is developing a Generation IV Reactor, International Reactor Innovative and Secure (IRIS). IRIS is a modular, integral, light water cooled, low-to-medium power (100-350 MWe) reactor which addresses the requirements defined by the US DOE for Generation IV reactors, i.e., fuel cycle sustainability, enhanced reliability and safety, and improved economics. It features innovative, advanced engineering, but it does not require new technology development since it relies on the proven technology of light water reactors. This paper presents the current reference IRIS design, which features a 1000 MWt thermal core with proven 5%-enriched uranium oxide fuel and four-year long straight burn fuel cycle, integral reactor vessel housing helical tube steam generators and immersed spool pumps. Other major contributors to the high level of safety and economic attractiveness are the safety by design and optimized maintenance approaches, which allow elimination of some classes of accidents, lower capital cost, long operating cycle, and high capacity factors. The path forward for possible future extension to a eight-year cycle will be also discussed. IRIS has a large potential worldwide market because of its proven technology, modularity, low financing, compatibility with existing grids and very limited infrastructure requirements. It is especially appealing to developing countries because of ease of operation and because its medium power is more adaptable to smaller grids. (author)

  15. Generation IV SFR Nuclear Reactors: Under Sodium Robotics for ASTRID

    International Nuclear Information System (INIS)

    Jouan-de-Kervenoael, T.; Rey, F.; Baque, F.

    2013-06-01

    For non-removable components of the future ASTRID prototype, repair operations will be performed in a gas environment. If the faulty area is located under the sodium free level, the gas-tight system will have to contain the inspection and repair tools and to protect them from the surrounding liquid sodium. Concerning repair tools, the unique laser tool has been selected for future SFRs: the repair scenario for in-sodium structures will first involve removing the sodium (after bulk draining), then machining and finally welding. Concerning conventional tools (brush or gas blower for sodium removal, milling machine for machining and TIG for welding for which its feasibility was demonstrated in the 1990's) are still considered as a back-up solution. The maintenance of future ASTRID nuclear reactor prototype (inspection, repair) will be performed during shut down periods with some robotic carriers which have to be introduced within the main vessel, in primary 200 deg. C sodium coolant with argon gas cover. Inspection campaigns will be 20 days long. These robots (or carriers) will allow bringing inspection and repairing tools up to concerned components and structures. The needed degrees of freedom associated to these operations will be assumed either directly by the carrier itself or by specifics lower end carrier device for accurate local positioning. Several carriers will be designed, well adapted to specific needs: type of maintenance operation and location of inspection and repair sites. Feedback experience was gained during Superphenix SFR operation with the MIR robot which allowed to successfully make the Non Destructive Examination of main vessel welding joints, the carrier being outside bulk sodium. Operating conditions for the ASTRID robots will be harder from those of the MIR robot: temperature ranging from 180 deg. C to 200 deg. C, radiation dose ranging from 105 to 106 Gy, but mainly direct immersion within liquid sodium coolant. At the design phase of

  16. Structural materials for Gen-IV nuclear reactors: Challenges and opportunities

    Science.gov (United States)

    Murty, K. L.; Charit, I.

    2008-12-01

    Generation-IV reactor design concepts envisioned thus far cater toward a common goal of providing safer, longer lasting, proliferation-resistant and economically viable nuclear power plants. The foremost consideration in the successful development and deployment of Gen-IV reactor systems is the performance and reliability issues involving structural materials for both in-core and out-of-core applications. The structural materials need to endure much higher temperatures, higher neutron doses and extremely corrosive environment, which are beyond the experience of the current nuclear power plants. Materials under active consideration for use in different reactor components include various ferritic/martensitic steels, austenitic stainless steels, nickel-base superalloys, ceramics, composites, etc. This paper presents a summary of various Gen-IV reactor concepts, with emphasis on the structural materials issues depending on the specific application areas. This paper also discusses the challenges involved in using the existing materials under both service and off-normal conditions. Tasks become increasingly complex due to the operation of various fundamental phenomena like radiation-induced segregation, radiation-enhanced diffusion, precipitation, interactions between impurity elements and radiation-produced defects, swelling, helium generation and so forth. Further, high temperature capability (e.g. creep properties) of these materials is a critical, performance-limiting factor. It is demonstrated that novel alloy and microstructural design approaches coupled with new materials processing and fabrication techniques may mitigate the challenges, and the optimum system performance may be achieved under much demanding conditions.

  17. Environmental sensitivity studies for Gen-IV roadmap fast reactor scenario

    International Nuclear Information System (INIS)

    Jeong, Chang Joon

    2004-03-01

    The environmental effect of the self-sufficient fast reactor scenario, which is considered as one of the full fissile plutonium and transuranic recycle scenario in Gen-IV roadmap, has been analyzed by using the dynamic analysis method. Through the parametric calculations for the fast reactor deployment time and capacity, the environmental effects of the fuel cycle for important parameters such as the amount of spent fuel and the combined amounts of plutonium and minor actinides were estimated and compared to those of the once-through LWR fuel cycle. The results of the sensitivity calculations showed that an early deployment of the fast reactor with a high capacity can reduce the accumulation of spent fuel by up to 37%. Furthermore, the recycling of plutonium and minor actinides can reduce the key repository parameter (long term decay heat). Therefore the favorable environmental effects can be expected with the implementation of the symbiotic fast reactor scenario

  18. Market share scenarios for Gen-DIII and gen-IV reactors in Europe

    International Nuclear Information System (INIS)

    Roelofs, F.; Heek, A. V.; Durpel, L. V. D.

    2008-01-01

    Nuclear energy is back on the agenda worldwide in order to meet growing energy demand and especially the growth in electricity demand. Many objectives direct to an increased use of nuclear energy, i.e. minimising energy costs, reducing climate change effects and others. In the light of the potential renewed growth of nuclear energy, the public demands a clear view on what nuclear energy may contribute towards meeting these objectives and especially how nuclear energy may address some socio-political obstructions with respect to economics, radioactive waste, safety and proliferation of fissile materials. To address these questions, the future nuclear reactor park mix in Europe has been analysed applying an integrated dynamic process modelling technique. Various market share scenarios for nuclear energy are derived including sub-variants with regard to the intra-nuclear options. In the analyses, it is assumed that different types of new reactors may be built, taking into account the introduction date of considered Gen-Ill (i.e. EPR) and Gen-IV (i.e. SCWR, HTR, FR) reactors, and the economic evaluation of the complete fuel cycle. The assessment was undertaken using the DANESS code (Dynamic Analysis of Nuclear Energy System Strategies). The analyses show that given the considered realistic nuclear energy demand and given a limited number of available Gen-III and Gen-IV reactor types, the future European nuclear park will exist of combinations of Gen-III and Gen-IV reactors. This mix will always consist of a set of reactor types each having its specific strengths. The analyses also highlight the triggers influencing the choice between different nuclear energy deployment scenarios. (authors)

  19. Fast reactor development and worldwide cooperation in Generation-IV International Forum

    International Nuclear Information System (INIS)

    Sagayama, Yutaka

    2013-01-01

    Objectives of Gen-IV systems development: Goals: Four challenging technology goals have been defined to be applied to innovative nuclear reactor concepts in the 21st century: 1) Safety and Reliability (safe and reliable operation, no offsite emergency response); 2) Sustainability (effective fuel utilization, minimization of nuclear waste); 3) Proliferation Resistance & Physical Protection (to assure unattractive and the least desirable route for diversion or theft of weapons-usable materials, and provide increased physical protection against acts of terrorism); 4) Economic Competitiveness (life-cycle cost advantage over other energy resources). Phase: Each Generation-IV reactor system is one of three stages. 1) Viability Phase; 2) Performance Phase; 3) Demonstration Phase. Target: Commercial Deployment is expected around 2030s or beyond

  20. Description and user's manual of light water reactor fuel analysis code FEMAXI-IV (Ver.2)

    International Nuclear Information System (INIS)

    Suzuki, Motoe; Saitou, Hiroaki.

    1997-03-01

    FEMAXI-IV is an advanced version of FEMAXI-III, the analysis code of light water reactor fuel behavior in which various functions and improvements have been incorporated. The present report describes in detail the basic theories and structure, the models and numerical solutions applied, and the material properties adopted in the version 2 which is an improved version of the first version of FEMAXI-IV. In FEMAXI-IV (Ver.2), bugs have been fixed, pellet thermal conductivity properties have been updated, and thermal-stress-induced FP gas release model have been incorporated. In order to facilitate effective and wide-ranging application of the code, types and methods of input/output of the code are also described, and a sample output in an actual form is included. (author)

  1. Generation IV reactors and the ASTRID prototype: lessons from the Fukushima accident

    International Nuclear Information System (INIS)

    Gauche, F.

    2012-01-01

    In France, the ASTRID prototype is an industrial demonstrator of a sodium-cooled fast neutron reactor (SFR), fulfilling the criteria for Generation IV reactors. ASTRID will meet safety requirements as stringent as for third generation reactors, and it takes into account lessons from the Fukushima accident. The objectives are to reinforce the robustness of the safety demonstration for all safety functions. ASTRID will feature an innovative core with a negative sodium void coefficient, it will take advantage of the large thermal inertia of SFR for decay heat removal, and will provide for a design either eliminating the sodium-water reaction, or guaranteeing no consequences for safety in case such reaction would take place. (author)

  2. DESIGN AND LAYOUT CONCEPTS FOR COMPACT, FACTORY-PRODUCED, TRANSPORTABLE, GENERATION IV REACTOR SYSTEMS

    Energy Technology Data Exchange (ETDEWEB)

    Mynatt Fred R.; Townsend, L.W.; Williamson, Martin; Williams, Wesley; Miller, Laurence W.; Khan, M. Khurram; McConn, Joe; Kadak, Andrew C.; Berte, Marc V.; Sawhney, Rapinder; Fife, Jacob; Sedler, Todd L.; Conway, Larry E.; Felde, Dave K.

    2003-11-12

    The purpose of this research project is to develop compact (100 to 400 MWe) Generation IV nuclear power plant design and layout concepts that maximize the benefits of factory-based fabrication and optimal packaging, transportation and siting. The reactor concepts selected were compact designs under development in the 2000 to 2001 period. This interdisciplinary project was comprised of three university-led nuclear engineering teams identified by reactor coolant type (water, gas, and liquid metal) and a fourth Industrial Engineering team. The reactors included a Modular Pebble Bed helium-cooled concept being developed at MIT, the IRIS water-cooled concept being developed by a team led by Westinghouse Electric Company, and a Lead-Bismuth-cooled concept developed by UT. In addition to the design and layout concepts this report includes a section on heat exchanger manufacturing simulations and a section on construction and cost impacts of proposed modular designs.

  3. DESIGN AND LAYOUT CONCEPTS FOR COMPACT, FACTORY-PRODUCED, TRANSPORTABLE, GENERATION IV REACTOR SYSTEMS

    International Nuclear Information System (INIS)

    Mynatt, Fred R.; Townsend, L.W.; Williamson, Martin; Williams, Wesley; Miller, Laurence W.; Khan, M. Khurram; McConn, Joe; Kadak, Andrew C.; Berte, Marc V.; Sawhney, Rapinder; Fife, Jacob; Sedler, Todd L.; Conway, Larry E.; Felde, Dave K.

    2003-01-01

    The purpose of this research project is to develop compact (100 to 400 MWe) Generation IV nuclear power plant design and layout concepts that maximize the benefits of factory-based fabrication and optimal packaging, transportation and siting. The reactor concepts selected were compact designs under development in the 2000 to 2001 period. This interdisciplinary project was comprised of three university-led nuclear engineering teams identified by reactor coolant type (water, gas, and liquid metal) and a fourth Industrial Engineering team. The reactors included a Modular Pebble Bed helium-cooled concept being developed at MIT, the IRIS water-cooled concept being developed by a team led by Westinghouse Electric Company, and a Lead-Bismuth-cooled concept developed by UT. In addition to the design and layout concepts this report includes a section on heat exchanger manufacturing simulations and a section on construction and cost impacts of proposed modular designs

  4. A preliminary safety analysis for the prototype Gen IV Sodium-Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kwi Lim; Ha, Kwi Seok; Jeong, Jae Ho; Choi, Chi Woong; Jeong, Tae Kyeong; Ahn, Sang June; Lee, Seung Won; Chang, Won Pyo; Kang, Seok Hun; Yoo, Jae Woon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Korea Atomic Energy Research Institute has been developing a pool-type sodium-cooled fast reactor of the Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR). To assess the effectiveness of the inherent safety features of the PGSFR, the system transients during design basis accidents and design extended conditions are analyzed with MARS-LMR and the subchannel blockage events are analyzed with MATRA-LMR-FB. In addition, the in-vessel source term is calculated based on the super-safe, small, and simple reactor methodology. The results show that the PGSFR meets safety acceptance criteria with a sufficient margin during the events and keeps accidents from deteriorating into more severe accidents.

  5. Synergies in the design and development of fusion and generation IV fission reactors

    International Nuclear Information System (INIS)

    Bogusch, E.; Carre, F.; Knebel, J.; Aoto, K.

    2007-01-01

    Future fusion reactors or systems and Generation IV fission reactors are designed and developed in worldwide programmes mostly involving the same partners to investigate and assess their potential for realisation and contribution to meet the future energy needs beyond 2030. Huge scientific and financial effort is necessary to meet these objectives. First programmes have been launched in Generation IV International Forum (GIF) for fission and in the Broader Approach for fusion reactor system development. Except the physics basis for the energy source, future fusion and fission reactors, in particular those with fast neutron core face similar design issues and development needs. Therefore the call for the identification of synergies became evident. Beyond ITER cooled by water, future fusion reactors or systems will be designed for helium and liquid metal cooling and higher temperatures similar to those proposed for some of the six fission reactor concepts in GIF with their diverse coolants. Beside materials developments which are not discussed in this paper, design and performance of components and systems related to the diverse coolants including lifetime and maintenance aspects might offer significant potentials for synergies. Furthermore, the use of process heat for applications in addition to electricity production as well as their safety approaches might create synergistic design and development programmes. Therefore an early identification of possible synergies in the relevant programmes should be endorsed to minimise the effort for future power plants in terms of investments and resources. In addition to a general overview of a possible synergistic work programme which promotes the interaction between fusion and fission programmes towards an integrated organisation of their design and R and D programmes, some specific remarks will be given for joint design tools, numerical code systems and joint experiments in support of common technologies. (orig.)

  6. Synergies in the design and development of fusion and generation IV fission reactors

    International Nuclear Information System (INIS)

    Bogusch, E.; Carre, F.; Knebel, J.U.; Aoto, K.

    2008-01-01

    Future fusion reactor and Generation IV fission reactor systems are designed and developed in worldwide programmes to investigate and assess their potential for realisation and contribution to the future energy needs beyond 2030 mostly involving the same partners. Huge scientific and financial effort is necessary to meet these objectives. First programmes have been launched in Generation IV International Forum (GIF) for fission and in the Broader Approach for fusion reactor system development. Except for the physics basis for the energy source, future fusion and fission reactors, in particular those with fast neutron core, face similar design issues and development needs. Therefore, the call for the identification of synergies became evident. Beyond ITER cooled by water, future fusion reactor systems will be designed for high-temperature helium and liquid metal cooling but also water including supercritical water and molten salt similar to those proposed for some of the six fission reactor concepts in GIF with their diverse coolants. Beside materials developments which are not discussed in this paper, design and performance of components and systems related to the diverse coolants including lifetime and maintenance aspects might offer significant potentials for synergies. Furthermore, the use of process heat for applications in addition to electricity production as well as their safety approaches can create synergistic design and development programmes. Therefore, an early identification of possible synergies in the relevant programmes should be endorsed to minimise the effort for future power plants in terms of investments and resources. In addition to a general overview of a possible synergistic work programme which promotes the interaction between fusion and fission programmes towards an integrated organisation of their design and R and D programmes, some specific remarks will be given for joint design tools, numerical code systems and joint experiments in

  7. Analysis of a Spanish energy scenario with Generation IV nuclear reactors

    International Nuclear Information System (INIS)

    Ochoa, Raquel; Jimenez, Gonzalo; Perez-Martin, Sara

    2013-01-01

    Highlights: • Spanish energy scenario for the hypothetical deployment of Gen-IV SFR reactors. • Availability of national resources is assessed, considering SFR’s breeding. • An assessment of the impact of transmuting MA on the final repository. • SERPENT code with own pre- and post-processing tools were employed. • The employed SFR core design is based on the specifications of the CP-ESFR. - Abstract: The advantages of fast-spectrum reactors consist not only of an efficient use of fuel through the breeding of fissile material and the use of natural or depleted uranium, but also of the potential reduction of the amount of actinides such as americium and neptunium contained in the irradiated fuel. The first aspect means a guaranteed future nuclear fuel supply. The second fact is key for high-level radioactive waste management, because these elements are the main responsible for the radioactivity of the irradiated fuel in the long term. The present study aims to analyze the hypothetical deployment of a Gen-IV Sodium Fast Reactor (SFR) fleet in Spain. A nuclear fleet of fast reactors would enable a fuel cycle strategy different than the open cycle, currently adopted by most of the countries with nuclear power. A transition from the current Gen-II to Gen-IV fleet is envisaged through an intermediate deployment of Gen-III reactors. Fuel reprocessing from the Gen-II and Gen-III Light Water Reactors (LWR) has been considered. In the so-called advanced fuel cycle, the reprocessed fuel used to produce energy will breed new fissile fuel and transmute minor actinides at the same time. A reference case scenario has been postulated and further sensitivity studies have been performed to analyze the impact of the different parameters on the required reactor fleet. The potential capability of Spain to supply the required fleet for the reference scenario using national resources has been verified. Finally, some consequences on irradiated final fuel inventory are assessed

  8. Advanced Computational Materials Science: Application to Fusion and Generation IV Fission Reactors (Workshop Report)

    Energy Technology Data Exchange (ETDEWEB)

    Stoller, RE

    2004-07-15

    The ''Workshop on Advanced Computational Materials Science: Application to Fusion and Generation IV Fission Reactors'' was convened to determine the degree to which an increased effort in modeling and simulation could help bridge the gap between the data that is needed to support the implementation of these advanced nuclear technologies and the data that can be obtained in available experimental facilities. The need to develop materials capable of performing in the severe operating environments expected in fusion and fission (Generation IV) reactors represents a significant challenge in materials science. There is a range of potential Gen-IV fission reactor design concepts and each concept has its own unique demands. Improved economic performance is a major goal of the Gen-IV designs. As a result, most designs call for significantly higher operating temperatures than the current generation of LWRs to obtain higher thermal efficiency. In many cases, the desired operating temperatures rule out the use of the structural alloys employed today. The very high operating temperature (up to 1000 C) associated with the NGNP is a prime example of an attractive new system that will require the development of new structural materials. Fusion power plants represent an even greater challenge to structural materials development and application. The operating temperatures, neutron exposure levels and thermo-mechanical stresses are comparable to or greater than those for proposed Gen-IV fission reactors. In addition, the transmutation products created in the structural materials by the high energy neutrons produced in the DT plasma can profoundly influence the microstructural evolution and mechanical behavior of these materials. Although the workshop addressed issues relevant to both Gen-IV and fusion reactor materials, much of the discussion focused on fusion; the same focus is reflected in this report. Most of the physical models and computational methods

  9. Benchmark analysis of SPERT-IV reactor with Monte Carlo code MVP

    International Nuclear Information System (INIS)

    Motalab, M.A.; Mahmood, M.S.; Khan, M.J.H.; Badrun, N.H.; Lyric, Z.I.; Altaf, M.H.

    2014-01-01

    Highlights: • MVP was used for SPERT-IV core modeling. • Neutronics analysis of SPERT-IV reactor was performed. • Calculation performed to estimate critical rod height, excess reactivity. • Neutron flux, time integrated neutron flux and Cd-ratio also calculated. • Calculated values agree with experimental data. - Abstract: The benchmark experiment of the SPERT-IV D-12/25 reactor core has been analyzed with the Monte Carlo code MVP using the cross-section libraries based on JENDL-3.3. The MVP simulation was performed for the clean and cold core. The estimated values of K eff at the experimental critical rod height and the core excess reactivity were within 5% with the experimental data. Thermal neutron flux profiles at different vertical and horizontal positions of the core were also estimated. Cadmium Ratio at different point of the core was also estimated. All estimated results have been compared with the experimental results. Generally good agreement has been found between experimentally determined and the calculated results

  10. The SGR Multipurpose - Generation IV - Transportable Cogeneration Nuclear Reactor with Innovative Shielding

    International Nuclear Information System (INIS)

    Pahladsingh, R.R.

    2002-01-01

    Deregulation and liberalization are changing the global energy-markets. At the same time innovative technologies are introduced in the electricity industry; often as a requirement from the upcoming Digital Society. Energy solutions for the future are more seen as a mix of energy-sources for generation-, transmission- and distribution energy-services. The Internet Energy-web based 'Virtual' enterprises are coming up and will gradually change our society. It the fast changing world we have to realize that there will be less time to look for the adequate solutions to anticipate on global developments and the way they will influence our own societies. Global population may reach 9 billion people by 2030; this will put tremendous pressure on energy-, water- and food supply in the global economy. It is time to think about some major issues as described below and come up with the right answers. These are needed on very short term to secure a humane global economic growth and the sustainable global environment. The DOE (Department of Energy - USA) has started the Generation IV initiative for the new generation of nuclear reactors that must lead to much better safety, economics and public acceptance the new reactors. The SGR (Simplified Gas-cooled Reactor) is being proposed as a Generation IV modular nuclear reactor, using graphite pebbles as fuel, whereby an attempt has been made to meet all the DOE requirements, to be used for future nuclear reactors. The focus in this paper is on the changing and emerging global energy-markets and shows some relevant criteria to the nuclear industry and how we can anticipate with improved and new designs towards the coming Digital Society. (author)

  11. Theoretical analysis of nuclear reactors (Phase I), I-V, Part III, Reactor poisoning

    International Nuclear Information System (INIS)

    Pop-Jordanov, J.

    1962-07-01

    Method was developed for calculation of Xe 135 static effect and kinetic effects of Xe 135 and Sm 149 with separate treatment of iodine effect and influence of reactor poisoning during power increase. Mentioned effects are treated first for uranium fuel and then the basic formulae were generalized for a mixture of fissile material. The annex contains a table with data needed for calculations and the Xe 13 5 microscopic capture cross section dependent on temperature [sr

  12. Use of thorium in the generation IV Molten Salt reactors and perspectives for Brazil

    International Nuclear Information System (INIS)

    Seneda, Jose A.; Lainetti, Paulo E.O.

    2013-01-01

    Interest in thorium stems mainly from the fact that it is expected a substantial increase in uranium prices over the next fifty years. The reactors currently in operation consume 65,500 tons of uranium per year. Each electrical gigawatt (GWe) additional need about 200 tU mined per year. So advanced fuel cycles, which increase the reserves of nuclear materials are interesting, particularly the use of thorium to produce the fissile isotope 233 U. It is important to mention some thorium advantages. Thorium is three to five times more abundant than uranium in the earth's crust. Thorium has only one oxidation state. Additionally, thoria produces less radiotoxicity than the UO 2 because it produces fewer amounts of actinides, reducing the radiotoxicity of long life nuclear waste. ThO 2 has higher corrosion resistance than UO 2 , besides being chemically stable due to their low water solubility. The burning of Pu in a reactor based in thorium also decreases the inventories of Pu from the current fuel cycles, resulting in lower risks of material diversion for use in nuclear weapons. There are some ongoing projects in the world, taking into consideration the proposed goals for Generation IV reactors, namely: sustainability, economics, safety and reliability, proliferation resistance and physical protection. Some developments on the use of thorium in reactors are underway, with the support of the IAEA and some governs. Can be highlighted some reactor concepts using thorium as fuel: CANDU; ADTR -Accelerator Driven Thorium Reactor; AHWR -Advanced Heavy Water Reactor proposed by India (light water cooled and moderated by heavy water) and the MSR -Molten Salt Reactor. The latter is based on a reactor concept that has already been successfully tested in the U.S. in the 50s, for use in aircrafts. In this paper, we discuss the future importance of thorium, particularly for Brazil, which has large mineral reserves of this strategic element, the characteristics of the molten salt

  13. Use of thorium in the generation IV Molten Salt reactors and perspectives for Brazil

    Energy Technology Data Exchange (ETDEWEB)

    Seneda, Jose A.; Lainetti, Paulo E.O., E-mail: jaseneda@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2013-07-01

    Interest in thorium stems mainly from the fact that it is expected a substantial increase in uranium prices over the next fifty years. The reactors currently in operation consume 65,500 tons of uranium per year. Each electrical gigawatt (GWe) additional need about 200 tU mined per year. So advanced fuel cycles, which increase the reserves of nuclear materials are interesting, particularly the use of thorium to produce the fissile isotope {sup 233}U. It is important to mention some thorium advantages. Thorium is three to five times more abundant than uranium in the earth's crust. Thorium has only one oxidation state. Additionally, thoria produces less radiotoxicity than the UO{sub 2} because it produces fewer amounts of actinides, reducing the radiotoxicity of long life nuclear waste. ThO{sub 2} has higher corrosion resistance than UO{sub 2}, besides being chemically stable due to their low water solubility. The burning of Pu in a reactor based in thorium also decreases the inventories of Pu from the current fuel cycles, resulting in lower risks of material diversion for use in nuclear weapons. There are some ongoing projects in the world, taking into consideration the proposed goals for Generation IV reactors, namely: sustainability, economics, safety and reliability, proliferation resistance and physical protection. Some developments on the use of thorium in reactors are underway, with the support of the IAEA and some governs. Can be highlighted some reactor concepts using thorium as fuel: CANDU; ADTR -Accelerator Driven Thorium Reactor; AHWR -Advanced Heavy Water Reactor proposed by India (light water cooled and moderated by heavy water) and the MSR -Molten Salt Reactor. The latter is based on a reactor concept that has already been successfully tested in the U.S. in the 50s, for use in aircrafts. In this paper, we discuss the future importance of thorium, particularly for Brazil, which has large mineral reserves of this strategic element, the

  14. Thermal stability study for candidate stainless steels of GEN IV reactors

    International Nuclear Information System (INIS)

    Simeg Veternikova, J.; Degmova, J.; Pekarcikova, M.; Simko, F.; Petriska, M.; Skarba, M.; Mikula, P.; Pupala, M.

    2016-01-01

    Highlights: • Thermal resistance of advanced stainless steels were observed at 1000 °C. • GEN IV candidate steels were confronted to classic AISI steels. • ODS AISI 316 has weaker thermal resistance than classic AISI steel. • Ferritic ODS steels and NF 709 has better thermal resistance than AISI steels. - Abstract: Candidate stainless steels for GEN IV reactors were investigated in term of thermal and corrosion stability at high temperatures. New austenitic steel (NF 709), austenitic ODS steel (ODS 316) and two ferritic ODS steels (MA 956 and MA 957) were exposed to around 1000 °C in inert argon atmosphere at pressure of ∼8 MPa. The steels were further studied in a light of vacancy defects presence by positron annihilation spectroscopy and their thermal resistance was confronted to classic AISI steels. The thermal strain supported a creation of oxide layers observed by scanning electron microscopy (SEM).

  15. Thermal stability study for candidate stainless steels of GEN IV reactors

    Energy Technology Data Exchange (ETDEWEB)

    Simeg Veternikova, J., E-mail: jana.veternikova@stuba.sk [Institute of Nuclear and Physical Engineering, Faculty of Electrical and Information Technology, Slovak University of Technology, Ilkovicova 3, 812 19 Bratislava (Slovakia); Degmova, J. [Institute of Nuclear and Physical Engineering, Faculty of Electrical and Information Technology, Slovak University of Technology, Ilkovicova 3, 812 19 Bratislava (Slovakia); Pekarcikova, M. [Institute of Materials Science, Faculty of Materials Science and Technology, Slovak University of Technology, Paulinska 16, 917 24 Trnava (Slovakia); Simko, F. [Department of Molten Salts, Institute of Inorganic Chemistry, Slovak Academy of Sciences, Dubravska cesta 9, 845 36 Bratislava (Slovakia); Petriska, M. [Institute of Nuclear and Physical Engineering, Faculty of Electrical and Information Technology, Slovak University of Technology, Ilkovicova 3, 812 19 Bratislava (Slovakia); Skarba, M. [Slovak University of Technology, Vazovova 5, 812 43 Bratislava (Slovakia); Mikula, P. [Institute of Nuclear and Physical Engineering, Faculty of Electrical and Information Technology, Slovak University of Technology, Ilkovicova 3, 812 19 Bratislava (Slovakia); Pupala, M. [Department of Molten Salts, Institute of Inorganic Chemistry, Slovak Academy of Sciences, Dubravska cesta 9, 845 36 Bratislava (Slovakia)

    2016-11-30

    Highlights: • Thermal resistance of advanced stainless steels were observed at 1000 °C. • GEN IV candidate steels were confronted to classic AISI steels. • ODS AISI 316 has weaker thermal resistance than classic AISI steel. • Ferritic ODS steels and NF 709 has better thermal resistance than AISI steels. - Abstract: Candidate stainless steels for GEN IV reactors were investigated in term of thermal and corrosion stability at high temperatures. New austenitic steel (NF 709), austenitic ODS steel (ODS 316) and two ferritic ODS steels (MA 956 and MA 957) were exposed to around 1000 °C in inert argon atmosphere at pressure of ∼8 MPa. The steels were further studied in a light of vacancy defects presence by positron annihilation spectroscopy and their thermal resistance was confronted to classic AISI steels. The thermal strain supported a creation of oxide layers observed by scanning electron microscopy (SEM).

  16. Nordic forum for generation IV reactors, status and activities in 2012

    International Nuclear Information System (INIS)

    Van Nieuwenhove, R.; Lauritzen, B.; Nonboel, E.

    2012-12-01

    The Nordic-Gen4 (continuation from NOMAGE4) seminar was this year hosted by DTU Nutech at Risoe, Denmark. The seminar was well attended (49 participants from 12 countries). The presentations covered many aspects in Gen-IV reactor research and gave a good overview of the activities within this field at the various institutes and universities. The present report contains book of abstracts. The individual Power Point presentations are indexed in INIS and may be found at http://nordic-gen4.org/seminars/nordic-gen4-riso-2012-2/ (LN)

  17. Nordic forum for generation IV reactors, status and activities in 2012

    Energy Technology Data Exchange (ETDEWEB)

    Van Nieuwenhove, R. [Institutt for Energiteknikk, OECD Halden Reactor Project, Kjeller (Norway); Lauritzen, B.; Nonboel, E. [Technical Univ. of Denmark. DTU Nutech, Roskilde (Denmark)

    2012-12-15

    The Nordic-Gen4 (continuation from NOMAGE4) seminar was this year hosted by DTU Nutech at Risoe, Denmark. The seminar was well attended (49 participants from 12 countries). The presentations covered many aspects in Gen-IV reactor research and gave a good overview of the activities within this field at the various institutes and universities. The present report contains book of abstracts. The individual Power Point presentations are indexed in INIS and may be found at http://nordic-gen4.org/seminars/nordic-gen4-riso-2012-2/ (LN)

  18. Decay Heat Removal in GEN IV Gas-Cooled Fast Reactors

    International Nuclear Information System (INIS)

    Lap-Yan, C.; Wie, T. Y. C.

    2009-01-01

    The safety goal of the current designs of advanced high-temperature thermal gas-cooled reactors (HTRs) is that no core meltdown would occur in a depressurization event with a combination of concurrent safety system failures. This study focused on the analysis of passive decay heat removal (DHR) in a GEN IV direct-cycle gas-cooled fast reactor (GFR) which is based on the technology developments of the HTRs. Given the different criteria and design characteristics of the GFR, an approach different from that taken for the HTRs for passive DHR would have to be explored. Different design options based on maintaining core flow were evaluated by performing transient analysis of a depressurization accident using the system code RELAP5-3D. The study also reviewed the conceptual design of autonomous systems for shutdown decay heat removal and recommends that future work in this area should be focused on the potential for Brayton cycle DHRs.

  19. Overall system description and safety characteristics of Prototype Gen IV Sodium Cooled Fast Reactor in Korea

    International Nuclear Information System (INIS)

    Yoo, Jae Woon; Chang, Jin Wook; Lim, Jae Yong; Cheon, Jin Sik; Lee, Tae Ho; Kim, Sung Kyun; Lee, Kwi Lim; Joo, Hyung Kook

    2016-01-01

    The Prototype Gen IV sodium cooled fast reactor (PGSFR) has been developed for the last 4 years, fulfilling the technology demonstration of the burning capability of transuranic elements included in light water reactor spent nuclear fuel. The PGSFR design has been focused on the robustness of safety systems by enhancing inherent safety characteristics of metal fuel and strengthening passive safety features using natural circulation and thermal expansion. The preliminary safety information document as a major outcome of the first design phase of PGSFR development was issued at the end of 2015. The project entered the second design phase at the beginning of 2016. This paper summarizes the overall structures, systems, and components of nuclear steam supply system and safety characteristics of the PGSFR. The research and development activities to demonstrate the safety performance are also briefly introduced in the paper

  20. Overall System Description and Safety Characteristics of Prototype Gen IV Sodium Cooled Fast Reactor in Korea

    Directory of Open Access Journals (Sweden)

    Jaewoon Yoo

    2016-10-01

    Full Text Available The Prototype Gen IV sodium cooled fast reactor (PGSFR has been developed for the last 4 years, fulfilling the technology demonstration of the burning capability of transuranic elements included in light water reactor spent nuclear fuel. The PGSFR design has been focused on the robustness of safety systems by enhancing inherent safety characteristics of metal fuel and strengthening passive safety features using natural circulation and thermal expansion. The preliminary safety information document as a major outcome of the first design phase of PGSFR development was issued at the end of 2015. The project entered the second design phase at the beginning of 2016. This paper summarizes the overall structures, systems, and components of nuclear steam supply system and safety characteristics of the PGSFR. The research and development activities to demonstrate the safety performance are also briefly introduced in the paper.

  1. Mechanical Design Concept of Fuel Assembly for Prototype GEN-IV Sodium-cooled Fast Reactor

    International Nuclear Information System (INIS)

    Yoon, K. H.; Lee, C. B.

    2014-01-01

    The prototype GEN-IV sodium-cooled fast reactor (PGSFR) is an advanced fast reactor plant design that utilizes compact modular pool-type reactors sized to enable factory fabrication and an affordable prototype test for design certification at minimum cost and risk. The design concepts of the fuel assembly (FA) were introduced for a PGSFR. Unlike that for the pressurized water reactor, there is a neutron shielding concept in the FA and recycling metal fuel. The PGSFR core is a heterogeneous, uranium-10% zirconium (U-10Zr) metal alloy fuel design with 112 assemblies: 52 inner core fuel assemblies, 60 outer core fuel assemblies, 6 primary control assemblies, 3 secondary control assemblies, 90 reflector assemblies and 102 B4C shield assemblies. This configuration is shown in Fig. 1. The core is designed to produce 150 MWe with an average temperature rise of 155 .deg. C. The inlet temperature is 390 .deg. C and the bulk outlet temperature is 545 .deg. C. The core height is 900 mm and the gas plenum length is 1,250 mm. A mechanical design of a fuel assembly for a PGSFR was established. The mechanical design concepts are well realized in the design. In addition to this, the analytical and experimental works will be carries out for verifying the design soundness

  2. Mechanical Design Concept of Fuel Assembly for Prototype GEN-IV Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, K. H.; Lee, C. B. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    The prototype GEN-IV sodium-cooled fast reactor (PGSFR) is an advanced fast reactor plant design that utilizes compact modular pool-type reactors sized to enable factory fabrication and an affordable prototype test for design certification at minimum cost and risk. The design concepts of the fuel assembly (FA) were introduced for a PGSFR. Unlike that for the pressurized water reactor, there is a neutron shielding concept in the FA and recycling metal fuel. The PGSFR core is a heterogeneous, uranium-10% zirconium (U-10Zr) metal alloy fuel design with 112 assemblies: 52 inner core fuel assemblies, 60 outer core fuel assemblies, 6 primary control assemblies, 3 secondary control assemblies, 90 reflector assemblies and 102 B4C shield assemblies. This configuration is shown in Fig. 1. The core is designed to produce 150 MWe with an average temperature rise of 155 .deg. C. The inlet temperature is 390 .deg. C and the bulk outlet temperature is 545 .deg. C. The core height is 900 mm and the gas plenum length is 1,250 mm. A mechanical design of a fuel assembly for a PGSFR was established. The mechanical design concepts are well realized in the design. In addition to this, the analytical and experimental works will be carries out for verifying the design soundness.

  3. Fundamental Understanding of Crack Growth in Structural Components of Generation IV Supercritical Light Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Iouri I. Balachov; Takao Kobayashi; Francis Tanzella; Indira Jayaweera; Palitha Jayaweera; Petri Kinnunen; Martin Bojinov; Timo Saario

    2004-11-17

    This work contributes to the design of safe and economical Generation-IV Super-Critical Water Reactors (SCWRs) by providing a basis for selecting structural materials to ensure the functionality of in-vessel components during the entire service life. During the second year of the project, we completed electrochemical characterization of the oxide film properties and investigation of crack initiation and propagation for candidate structural materials steels under supercritical conditions. We ranked candidate alloys against their susceptibility to environmentally assisted degradation based on the in situ data measure with an SRI-designed controlled distance electrochemistry (CDE) arrangement. A correlation between measurable oxide film properties and susceptibility of austenitic steels to environmentally assisted degradation was observed experimentally. One of the major practical results of the present work is the experimentally proven ability of the economical CDE technique to supply in situ data for ranking candidate structural materials for Generation-IV SCRs. A potential use of the CDE arrangement developed ar SRI for building in situ sensors monitoring water chemistry in the heat transport circuit of Generation-IV SCWRs was evaluated and proved to be feasible.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-01

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

  5. ASAMPSA2 best-practices guidelines for L2 PSA development and applications. Volume 3 - Extension to Gen IV reactors

    International Nuclear Information System (INIS)

    Bassi, C.; Bonneville, H.; Brinkman, H.; Burgazzi, L.; Polidoro, F.; Vincon, L.; Jouve, S.

    2010-01-01

    The main objective assigned to the Work Package 4 (WP4) of the 'ASAMPSA2' project (EC 7. FPRD) consist in the verification of the potential compliance of L2PSA guidelines based on PWR/BWR reactors (which are specific tasks of WP2 and WP3) with Generation IV representative concepts. Therefore, in order to exhibit potential discrepancies between LWRs and new reactor types, the following work was based on the up-to-date designs of: - The European Fast Reactor (EFR) which will be considered as prototypical of a pool-type Sodium-cooled Fast Reactor (SFR); - The ELSY design for the Lead-cooled Fast Reactor (LFR) technology; - The ANTARES project which could be representative of a Very-High Temperature Reactor (VHTR); - The CEA 2400 MWth Gas-cooled Fast Reactor (GFR). (authors)

  6. Metal fuel development and verification for prototype generation- IV Sodium- Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chan Bock; Cheon, Jin Sik; Kim, Sung Ho; Park, Jeong Yong; Joo, Hyung Kook [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Metal fuel is being developed for the prototype generation-IV sodium-cooled fast reactor (PGSFR) to be built by 2028. U-Zr fuel is a driver for the initial core of the PGSFR, and U -transuranics (TRU)-Zr fuel will gradually replace U-Zr fuel through its qualification in the PGSFR. Based on the vast worldwide experiences of U-Zr fuel, work on U-Zr fuel is focused on fuel design, fabrication of fuel components, and fuel verification tests. U-TRU-Zr fuel uses TRU recovered through pyroelectrochemical processing of spent PWR (pressurized water reactor) fuels, which contains highly radioactive minor actinides and chemically active lanthanide or rare earth elements as carryover impurities. An advanced fuel slug casting system, which can prevent vaporization of volatile elements through a control of the atmospheric pressure of the casting chamber and also deal with chemically active lanthanide elements using protective coatings in the casting crucible, was developed. Fuel cladding of the ferritic-martensitic steel FC92, which has higher mechanical strength at a high temperature than conventional HT9 cladding, was developed and fabricated, and is being irradiated in the fast reactor.

  7. Metal Fuel Development and Verification for Prototype Generation IV Sodium-Cooled Fast Reactor

    Directory of Open Access Journals (Sweden)

    Chan Bock Lee

    2016-10-01

    Full Text Available Metal fuel is being developed for the prototype generation-IV sodium-cooled fast reactor (PGSFR to be built by 2028. U–Zr fuel is a driver for the initial core of the PGSFR, and U–transuranics (TRU–Zr fuel will gradually replace U–Zr fuel through its qualification in the PGSFR. Based on the vast worldwide experiences of U–Zr fuel, work on U–Zr fuel is focused on fuel design, fabrication of fuel components, and fuel verification tests. U–TRU–Zr fuel uses TRU recovered through pyroelectrochemical processing of spent PWR (pressurized water reactor fuels, which contains highly radioactive minor actinides and chemically active lanthanide or rare earth elements as carryover impurities. An advanced fuel slug casting system, which can prevent vaporization of volatile elements through a control of the atmospheric pressure of the casting chamber and also deal with chemically active lanthanide elements using protective coatings in the casting crucible, was developed. Fuel cladding of the ferritic–martensitic steel FC92, which has higher mechanical strength at a high temperature than conventional HT9 cladding, was developed and fabricated, and is being irradiated in the fast reactor.

  8. Light water reactor fuel analysis code FEMAXI-IV(Ver.2). Detailed structure and user`s manual

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, Motoe [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Saitou, Hiroaki

    1997-11-01

    A light water reactor fuel behavior analysis code FEMAXI-IV(Ver.2) was developed as an improved version of FEMAXI-IV. Development of FEMAXI-IV has been already finished in 1992, though a detailed structure and input manual of the code have not been open to users yet. Here, the basic theories and structure, the models and numerical solutions applied to FEMAXI-IV(Ver.2), and the material properties adopted in the code are described in detail. In FEMAXI-IV(Ver.2), programming bugs in previous FEMAXI-IV were eliminated, renewal of the pellet thermal conductivity was performed, and a model of thermal-stress restraint on FP gas release was incorporated. For facilitation of effective and wide-ranging application of the code, methods of input/output of the code are also described in detail, and sample output is included. (author)

  9. Progress reports for Gen IV sodium fast reactor activities FY 2007

    International Nuclear Information System (INIS)

    Cahalan, J. E.; Tentner, A. M.

    2007-01-01

    for prevention of progression into severe accident conditions (prevention of core melting) or for mitigation of severe accident consequences (mitigation of the impact of core melting to protect public health and safety). Because design measures for severe accident prevention and mitigation are beyond the normal design basis, established regulatory guidelines and codes do not provide explicit identification of the design performance requirements for severe accident accommodation. The treatment of severe accidents is one of the key issues of R and D plans for the Gen IV systems in general, and for the Sodium Fast Reactor (SFR) in particular. Despite the lack of an unambiguous definition of safety approach applicable for severe accidents, there is an emerging consensus on the need for their consideration for the design. The US SFR program and Argonne National Laboratory (ANL) in particular have actively studied the potential scenarios and consequences of Hypothetical Core Disruptive Accidents (HCDA) for SFRs with oxide fuel during the Fast Flux Test Facility (FFTF) and Clinch River Breeder Reactor Plant (CRBRP) programs in the 70s and 80s. Later, the focus of the US SFR safety R and D activities shifted to the prevention of all HCDAs through passive safety features of the SFRs with metal fuel in the Integral Fast Reactor (IFR) program, and the study of severe accident consequences was de-emphasized. The goal of this paper is to provide an overview of the current SFR safety approach and the role of severe accidents in Japan and France, in preparation for an expected and more active collaboration in this area between the US, Japan, and France

  10. Generation IV Nuclear Energy Systems Construction Cost Reductions through the Use of Virtual Environments - Task 5 Report: Generation IV Reactor Virtual Mockup Proof-of-Principle Study

    International Nuclear Information System (INIS)

    Timothy Shaw; Anthony Baratta; Vaughn Whisker

    2005-01-01

    Task 5 report is part of a 3 year DOE NERI-sponsored effort evaluating immersive virtual reality (CAVE) technology for design review, construction planning, and maintenance planning and training for next generation nuclear power plants. Program covers development of full-scale virtual mockups generated from 3D CAD data presented in a CAVE visualization facility. Created a virtual mockup of PBMR reactor cavity and discussed applications of virtual mockup technology to improve Gen IV design review, construction planning, and maintenance planning

  11. Generation IV Nuclear Energy Systems Construction Cost Reductions through the Use of Virtual Environments - Task 5 Report: Generation IV Reactor Virtual Mockup Proof-of-Principle Study

    Energy Technology Data Exchange (ETDEWEB)

    Timothy Shaw; Anthony Baratta; Vaughn Whisker

    2005-02-28

    Task 5 report is part of a 3 year DOE NERI-sponsored effort evaluating immersive virtual reality (CAVE) technology for design review, construction planning, and maintenance planning and training for next generation nuclear power plants. Program covers development of full-scale virtual mockups generated from 3D CAD data presented in a CAVE visualization facility. Created a virtual mockup of PBMR reactor cavity and discussed applications of virtual mockup technology to improve Gen IV design review, construction planning, and maintenance planning.

  12. Contribution of the IV generation fast reactors to the sustainable development

    International Nuclear Information System (INIS)

    Mendoza G, G.; Klapp E, J.L.

    2007-01-01

    During the XXI century all the energy forms are necessary for the sustainable development. A balanced energy politics has to use a mixture of energy sources that completes the objective of responding to the increase in the demand and that it uses non emitting gases sources of greenhouse effect like the nuclear one. It is evident the great existent difficulty to turn the objectives of emissions for the coming years without having the nuclear energy. Later on, the process continued outlining serious commitments among the development necessity, the improvement of the level of life and the competitiveness, and the execution from the established environmental requirements to world level. It is very foregone that the energy nuclear become the best energy source to improve the environmental conditions and that new initiatives are determined in those that this energy will have an important paper. The solution is to build a nuclear central of advanced design, using technologies that its help to brake the diffusion of the nuclear weapons. The nucleo electric energy at great scale should be developed on the base of designs of reactors and innovative processes of fuel that can lend technological support to the not nuclear proliferation regime, and that at the same time they contribute to satisfy the electricity demand in the world. In a scenario of increase of energy demand, mainly in the development countries, and of growing interest in the pollutants reduction originated by the use of fossil fuels, the nuclear reactors of IV Generation arise as proposal and challenge. Meanwhile the search of new technologies and innovations become imperative, translating an enormous evolution, not only in the conceptual projects, as well as in the fuel cycle so that, in a scenario of open economy, turn its more competitive. Inside the reactors of fourth generation, the quick reactors are configured as those that more assist to such demands and they will be, without a doubt, the reactors in

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

    International Nuclear Information System (INIS)

    Vierow, Karen; Aldemir, Tunc

    2009-01-01

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

  14. A Stochastic Proof of the Resonant Scattering Kernel and its Applications for Gen IV Reactors Type

    International Nuclear Information System (INIS)

    Becker, B.; Dagan, R.; Broeders, C.H.M.; Lohnert, G.

    2008-01-01

    Monte Carlo codes such as MCNP are widely accepted as almost-reference for reactor analysis. The Monte Carlo Code should therefore use as few as possible approximations in order to produce 'experimental-level' calculations. In this study we deal with one of the most problematic approximations done in MCNP in which the resonances are ignored for the secondary neutron energy distribution, namely the change of the energy and angular direction of the neutron after interaction with a heavy isotope with pronounced resonances. The endeavour of exploiting the influence of the resonances on the scattering kernel goes back to 1944 where E. Wigner and J. Wilkins developed the first temperature dependent scattering kernel. However only in 1998, the full analytical solution for the double differential resonant dependent scattering kernel was suggested by W. Rothenstein and R. Dagan. An independent stochastic approach is presented for the first time to confirm the above analytical kernel with a complete different methodology. Moreover, by manipulating in a subtle manner the scattering subroutine COLIDN of MCNP, it is proven that this very subroutine is, to some extent, inappropriate as well as the relevant explanation in the MCNP manual. The impact of this improved resonance dependent scattering kernel on diverse types of reactors, in particular for the Generation IV innovative core design HTR, is shown to be significant. (authors)

  15. Unfolding of neutron spectra from Godiva type critical assemblies

    International Nuclear Information System (INIS)

    Harvey, J.T.; Meason, J.L.; Wright, H.L.

    1976-01-01

    The results from three experiments conducted at the White Sands Missile Range Fast Burst Reactor Facility are discussed. The experiments were designed to measure the ''free-field'' neutron leakage spectrum and the neutron spectra from mildly perturbed environments. SAND-II was used to calculate the neutron spectrum utilizing several different trial input spectra for each experiment. Comparisons are made between the unfolded neutron spectrum for each trial input on the basis of the following parameters: average neutron energy (above 10 KeV), integral fluence (above 10 KeV), spectral index and the hardness parameter, phi/sub eq//phi

  16. Design and Selection of Innovative Primary Circulation Pumps for GEN-IV Lead Fast Reactors

    Directory of Open Access Journals (Sweden)

    Walter Borreani

    2017-12-01

    Full Text Available Although Lead-cooled Fast Reactor (LFR is not a new concept, it continues to be an example of innovation in the nuclear field. Recently, there has been strong interest in liquid lead (Pb or liquid lead–bismuth eutectic (LBE both critical and subcritical systems in a relevant number of Countries, including studies performed in the frame of GENERATION-IV initiative. In this paper, the theoretical and computational findings for three different designs of Primary Circulation Pump (PCP evolving liquid lead (namely the jet pump, the Archimedean pump and the blade pump are presented with reference to the ALFRED (Advanced Lead Fast Reactor European Demonstrator design. The pumps are first analyzed from the theoretical point of view and then modeled with a 3D CFD code. Required design performance of the pumps are approximatively around an effective head of 2 bar with a mass flow rate of 5000 kg/s. Taking into account the geometrical constraints of the reactor and the fluid dynamics characteristics of the molten lead, the maximum design velocity for molten lead fluid flow of 2 m/s may be exceeded giving rise to unacceptable erosion phenomena of the blade or rotating component of the primary pumping system. For this reason a deep investigation of non-conventional axial pumps has been performed. The results presented shows that the design of the jet pump looks like beyond the current technological feasibility while, once the mechanical challenges of the Archimedean (screw pump and the fluid-dynamic issues of the blade pump will be addressed, both could represent viable solutions as PCP for ALFRED. Particularly, the blade pump shows the best performance in terms of pressure head generated in normal operation conditions as well as pressure drop in locked rotor conditions. Further optimizations (mainly for what the geometrical configuration is concerned are still necessary.

  17. RA Reactor operation and maintenance (I-IX), Part IV, Task 3.08/04, Refurbishment of the RA reactor

    International Nuclear Information System (INIS)

    Zecevic, V.

    1963-12-01

    This volume contains reports describing maintenance and repair work of the RA reactor instrumentation, equipment of the reactor dosimetry control system, and equipment for regulation and control systems

  18. Application of a new cross section library based on ENDF/B-IV to reactor core analysis

    International Nuclear Information System (INIS)

    Lima Bezerra, J. de.

    1991-04-01

    The use of the ENDF/B-IV library in the LEOPARD code for the Angra-1 reactor simulation is presented. The results are compared to those obtained using the ENDF/B-II library and show better values for the power distribution but an underestimated global reactivity as compared to experimental results. (F.E.). 1 ref, 55 figs, 1 tab

  19. Level II Probabilistic Safety Analysis Methodology for the Application to GEN-IV Sodium-cooled Fast Reactor

    International Nuclear Information System (INIS)

    Park, S. Y.; Kim, T. W.; Han, S. H.; Jeong, H. Y.

    2010-03-01

    The Korea Atomic Energy Research Institute (KAERI) has been developing liquid metal reactor (LMR) design technologies under a National Nuclear R and D Program. Nevertheless, there is no experience of the probabilistic safety assessment (PSA) domestically for a fast reactor with the metal fuel. Therefore, the objective of this study is to establish the methodologies of risk assessment for the reference design of GEN-IV sodium fast reactor (SFR). An applicability of the PSA methodology of U. S. NRC and PRISM plant to the domestic GEN-IV SFR has been studied. The study contains a plant damage state analysis, a containment event tree analysis, and a source-term release category binning process

  20. Integral parameters for the Godiva benchmark calculated by using theoretical and adjusted fission spectra of 235U

    International Nuclear Information System (INIS)

    Caldeira, A.D.

    1987-05-01

    The theoretical and adjusted Watt spectrum representations for 235 U are used as weighting functions to calculate K eff and θ f 28 /θ f 25 for the benchmark Godiva. The results obtained show that the values of K eff and θ f 28 /θ f 25 are not affected by spectrum form change. (author) [pt

  1. Nuclear data uncertainty analysis for the generation IV gas-cooled fast reactor

    International Nuclear Information System (INIS)

    Pelloni, S.; Mikityuk, K.

    2012-01-01

    For the European 2400 MW Gas-cooled Fast Reactor (GoFastR), this paper summarizes a priori uncertainties, i.e. without any integral experiment assessment, of the main neutronic parameters which were obtained on the basis of the deterministic code system ERANOS (Edition 2.2-N). JEFF-3.1 cross-sections were used in conjunction with the newest ENDF/B-VII.0 based covariance library (COMMARA-2.0) resulting from a recent cooperation of the Brookhaven and Los Alamos National Laboratories within the Advanced Fuel Cycle Initiative. The basis for the analysis is the original GoFastR concept with carbide fuel pins and silicon-carbide ceramic cladding, which was developed and proposed in the first quarter of 2009 by the 'French alternative energies and Atomic Energy Commission', CEA. The main conclusions from the current study are that nuclear data uncertainties of neutronic parameters may still be too large for this Generation IV reactor, especially concerning the multiplication factor, despite the fact that the new covariance library is quite complete; These uncertainties, in relative terms, do not show the a priori expected increase with bum-up as a result of the minor actinide and fission product build-up. Indeed, they are found almost independent of the fuel depletion, since the uncertainty associated with 238 U inelastic scattering results largely dominating. This finding clearly supports the activities of Subgroup 33 of the Working Party on International Nuclear Data Evaluation Cooperation (WPEC), i.e. Methods and issues for the combined use of integral experiments and covariance data, attempting to reduce the present unbiased uncertainties on nuclear data through adjustments based on available experimental data. (authors)

  2. NOMAGE4 activities 2011. Part I, Nordic Nuclear Materials Forum for Generation IV Reactors: Status and activities in 2011

    International Nuclear Information System (INIS)

    Van Nieuwenhove, R.

    2012-01-01

    A network for materials issues has been initiated in 2009 within the Nordic countries. The original objectives of the Generation IV Nordic Nuclear Materials Forum (NOMAGE4) were to form the basis of a sustainable forum for Gen-IV issues, especially focusing on fuels, cladding, structural materials and coolant interaction. Over the last years, other issues such as reactor physics, thermal hydraulics, safety and waste have gained in importance (within the network) and therefore the scope of the forum has been enlarged and a more appropriate and more general name, NORDIC-GEN4, has been chosen for the forum. Further, the interaction with non-Nordic countries (such as The Netherlands (JRC, NRG) and Czech Republic (CVR)) will be increased. Within the NOMAGE4 project, a seminar was organized by IFE-Halden during 31 October - 1 November 2011. The seminar attracted 65 participants from 12 countries. The seminar provided a forum for exchange of information, discussion on future research reactor needs and networking of experts on Generation IV reactor concepts. The participants could also visit the Halden reactor site and the workshop. (Author)

  3. NOMAGE4 activities 2011. Part I, Nordic Nuclear Materials Forum for Generation IV Reactors: Status and activities in 2011

    Energy Technology Data Exchange (ETDEWEB)

    Van Nieuwenhove, R. (Institutt for Energiteknikk, OECD Halden Reactor Project (Norway))

    2012-01-15

    A network for materials issues has been initiated in 2009 within the Nordic countries. The original objectives of the Generation IV Nordic Nuclear Materials Forum (NOMAGE4) were to form the basis of a sustainable forum for Gen-IV issues, especially focusing on fuels, cladding, structural materials and coolant interaction. Over the last years, other issues such as reactor physics, thermal hydraulics, safety and waste have gained in importance (within the network) and therefore the scope of the forum has been enlarged and a more appropriate and more general name, NORDIC-GEN4, has been chosen for the forum. Further, the interaction with non-Nordic countries (such as The Netherlands (JRC, NRG) and Czech Republic (CVR)) will be increased. Within the NOMAGE4 project, a seminar was organized by IFE-Halden during 31 October - 1 November 2011. The seminar attracted 65 participants from 12 countries. The seminar provided a forum for exchange of information, discussion on future research reactor needs and networking of experts on Generation IV reactor concepts. The participants could also visit the Halden reactor site and the workshop. (Author)

  4. Optimisation of the mechanical alloying process for odsferritic steels for generation IV reactors application

    International Nuclear Information System (INIS)

    Stanciulescu, M.; Carlan, P.; Mihalache, M.; Abrudeanu, M.

    2016-01-01

    ODS ferritic steels appear as promising materials for fusion and Gen IV fission reactors, offering high temperature performance, corrosion and irradiation resistance and meeting low activation criteria. Mechanical alloying (MA) is a powder metallurgy technique efficient for fabricating advanced materials, and has been used for strengthening structural materials including Fe-Cr alloys. In this paper a high-energy ball mill is used to study the microstructural evolution of 14YW alloy during the mechanical alloying process. The elemental powders are milled at a rotation speed of 250rot/min in cycles of 10min milling and 5min pause, with a ball-to-powder ration of 10:1 and in argon protective atmosphere. After 72 hours milling, the morphology and element distribution of the MA powders is investigated by scanning electron microscopy (SEM) and energy dispersive X-ray (EDX) analysis, respectively. It is observed that the particles size increases in the first milling stages and then decreases with the milling time. Changes in the material composition are analysed by X-ray diffraction (DRX). It seems that after milling part of the W remains non-dissolved in the Fe-Cr matrix retarding the solid solution formation. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

    Vierow, Karen; Aldemir, Tunc

    2009-09-10

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

  6. R and D programme on generation IV nuclear energy systems: the high temperatures gas-cooled reactors

    International Nuclear Information System (INIS)

    Carre, F.; Fiorini, G.L.; Billot, P.; Anzieu, P.; Brossard, P.

    2005-01-01

    The Generation IV Technology Roadmap selected, among others, a sequenced development of advanced high temperature gas cooled reactors as one of the main focus for R and D on future nuclear energy systems. The selection of this research objective originates both from the significance of high temperature and fast neutrons for nuclear energy to meet the needs for a sustainable development for the medium-long term (2020/2030 and beyond), and from the significant common R and D pathway that supports both medium term industrial projects and more advanced versions of gas cooled reactors. The first step of the 'Gas Technology Path' aims to support the development of a modular HTR to meet specific international market needs around 2020. The second step is a Very High Temperature Reactor - VHTR (>950 C) - to efficiently produce hydrogen through thermo-chemical or electro-chemical water splitting or to generate electricity with an efficiency above 50%, among other applications of high temperature nuclear heat. The third step of the Path is a Gas Fast Reactor - GFR - that features a fast-spectrum helium-cooled reactor and closed fuel cycle, with a direct or indirect thermodynamic cycle for electricity production and full recycle of actinides. Hydrogen production is also considered for the GFR. The paper succinctly presents the R and D program currently under definition and partially launched within the Generation IV International Forum on this consistent set of advanced gas cooled nuclear systems. (orig.)

  7. RB Research nuclear reactor, Annual report for 1995, I-IV

    International Nuclear Information System (INIS)

    Stefanovic, D.; Milosevic, M.; Pesic, M.; Marinkovic, P.; Ilic, R.; Dasic, N.; Milovanovic, S.; Ljubenov, V.; Petronijevic, M.; Jevremovic, M.

    1995-12-01

    Report on RB reactor operation during 1995 contains 3 parts. Part one contains a brief description of reactor operation and reactor components, relevant dosimetry data and radiation protection issues, personnel and financial data. Part two is devoted to maintenance of the reactor components, namely, fuel, heavy water, reactor vessel, heavy water circulation system, absorption rods and heavy water level-meters, maintenance of electronic, mechanical, electrical and auxiliary equipment. Part three contains data concerned with reactor operation and utilization with a comprehensive list of publications resulting from experiments done at the RB reactor

  8. RB Research nuclear reactor, Annual report for 1996, I-IV

    International Nuclear Information System (INIS)

    Stefanovic, D.; Milosevic, M.; Pesic, M.; Marinkovic, P.; Ilic, R.; Dasic, N.; Milovanovic, S.; Ljubenov, V.; Petronijevic, M.; Jevremovic, M.

    1996-12-01

    Report on RB reactor operation during 1996 contains 3 parts. Part one contains a brief description of reactor operation and reactor components, relevant dosimetry data and radiation protection issues, personnel and financial data. Part two is devoted to maintenance of the reactor components, namely, fuel, heavy water, reactor vessel, heavy water circulation system, absorption rods and heavy water level-meters, maintenance of electronic, mechanical, electrical and auxiliary equipment. Part three contains data concerned with reactor operation and utilization with a list of publications resulting from experiments done at the RB reactor

  9. Improvement of the decay heat removal characteristics of the generation IV gas-cooled fast reactor

    International Nuclear Information System (INIS)

    Epiney, A. S.

    2010-09-01

    The majority of NPPs worldwide are currently light water reactors, using ordinary water as both coolant and moderator. (...) For the longer-term future, viz. beyond the year 2030, Research and Development is currently ongoing on Generation IV NPPs, aimed at achieving closure of the nuclear fuel cycle, and hence both drastically improved utilization of fuel resources and minimization of long-lived radioactive wastes. Since the very beginning of the international cooperation on Generation IV, viz. the year 2000, the main research interest in Europe as regards the advanced fast-spectrum systems needed for achieving complete fuel cycle closure, has been for the Sodium-cooled Fast Reactor (SFR). However, the Gas-cooled Fast Reactor (GFR) is currently considered as the main back-up solution. Like the SFR, the GFR is an efficient breeder, also able to work as iso-breeder using simply natural uranium as feed and producing waste which is predominantly in the form of fission products. The main drawback of the GFR is the difficulty to evacuate decay heat following a loss-of-coolant accident (LOCA) due to the low thermal inertia of the core, as well as to the low coolant density. The present doctoral research focuses on the improvement of decay heat removal (DHR) for the Generation-IV GFR. The reference GFR system design considered in the thesis is the 2006 CEA concept, with a power of 2400 MWth. The CEA 2006 DHR strategy foresees, in all accidental cases (independent of the system pressure), that the reactor is shut down. For high pressure events, dedicated DHR loops with blowers and heat exchangers are designed to operate when the power conversion system cannot be used to provide acceptable core temperatures under natural convection conditions. For depressurized events, the strategy relies on a dedicated small containment (called the guard containment) providing an intermediate back-up pressure. The DHR blowers, designed to work under these pressure conditions, need to be

  10. Directions for attractive tokamak reactors: The ARIES-II and ARIES-IV second-stability designs

    International Nuclear Information System (INIS)

    Najmabadi, F.; Conn, R.W.

    1993-01-01

    ARIES is a research program to develop several visions of tokamak reactors with enhanced economic, safety, and environmental features. The ARIES study has developed four visions for tokamaks. All four designs are steady-state, 1000-MWe (net) power reactors. The ARIES-II and ARIES-IV designs assume potential advances in plasma physics (such as second-stability operation) predicted by theory but not yet established experimentally. The two designs have the same fusion plasma but different fusion-power-core. There are only minor differences between the ARIES-II and ARIES-IV plasma parameters. ARIES-IV is a 1000-MWe reactor with an average neutron wall loading of 3 MW/m 2 , and a mass power density of about 120 kWe/tonne of fusion power core. The reactor major radius is 6.1 m, the plasma minor radius is 1.5 m and the plasma elongation is 2, and the plasma triangularity is 0.67. The plasma current is low (6.8 MA), B on-axis is 7.7 T (corresponding to a maximum field at the coil of 16T), and the toroidal beta is 3.4% (Troyon coefficient = 6). The operating regime is optimized such that most of the plasma current (∼ 90%) is provided by the bootstrap current. ARIES-II uses liquid lithium as the coolant and tritium breeder. V-5Cr-5Ti is used as the structural material so that the potential of low-activation metallic blankets can be studied. ARIES-IV uses helium as the coolant, a solid tritium-breeding material (Li 2 O), and silicon carbide composite as structural material. The waste produced by neutron activation in both designs is found to meet the criteria allowing shallow-land burial under U.S. regulations. The cost of electricity for the ARIES-II-IV class of reactors is estimated to be about 20% lower than comparable, steady-state first-stability reactors (e.g. ARIES-I). 25 refs, 2 figs, 1 tab

  11. Godiva, a European Project for Ozone and Trace Gas Measurements from GOME

    Science.gov (United States)

    Goede, A. P. H.; Tanzi, C. P.; Aben, I.; Burrows, J. P.; Weber, M.; Perner, D.; Monks, P. S.; Llewellyn-Jones, D.; Corlett, G. K.; Arlander, D. W.; Platt, U.; Wagner, T.; Pfeilsticker, K.; Taalas, P.; Kelder, H.; Piters, A.

    GODIVA (GOME Data Interpretation, Validation and Application) is a European Commission project aimed at the improvement of GOME (Global Ozone Monitoring Experiment) data products. Existing data products include global ozone, NO2 columns and (ir)radiances. Advanced data products include O3 profiles, BrO, HCHO and OCIO columns. These data are validated by ground-based and balloon borne instruments. Calibration issues are investigated by in-flight monitoring using several complementary calibration sources, as well as an on-ground replica of the GOME instrument. The results will lead to specification of operational processing of the EUMETSAT ozone Satellite Application Facility as well as implementation of the improved and new GOME data products in the NILU database for use in the European THESEO (Third European Stratospheric Experiment on Ozone) campaign of 1999

  12. Improvement of Sodium Neutronic Nuclear Data for the Computation of Generation IV Reactors

    International Nuclear Information System (INIS)

    Archier, P.

    2011-01-01

    The safety criteria to be met for Generation IV sodium fast reactors (SFR) require reduced and mastered uncertainties on neutronic quantities of interest. Part of these uncertainties come from nuclear data and, in the particular case of SFR, from sodium nuclear data, which show significant differences between available international libraries (JEFF-3.1.1, ENDF/B-VII.0, JENDL-4.0). The objective of this work is to improve the knowledge on sodium nuclear data for a better calculation of SFR neutronic parameters and reliable associated uncertainties. After an overview of existing 23 Na data, the impact of the differences is quantified, particularly on sodium void reactivity effects, with both deterministic and stochastic neutronic codes. Results show that it is necessary to completely re-evaluate sodium nuclear data. Several developments have been made in the evaluation code Conrad, to integrate new nuclear reactions models and their associated parameters and to perform adjustments with integral measurements. Following these developments, the analysis of differential data and the experimental uncertainties propagation have been performed with Conrad. The resolved resonances range has been extended up to 2 MeV and the continuum range begins directly beyond this energy. A new 23 Na evaluation and the associated multigroup covariances matrices were generated for future uncertainties calculations. The last part of this work focuses on the sodium void integral data feedback, using methods of integral data assimilation to reduce the uncertainties on sodium cross sections. This work ends with uncertainty calculations for industrial-like SFR, which show an improved prediction of their neutronic parameters with the new evaluation. (author) [fr

  13. A Qualitative Assessment of Diversion Scenarios for a GEN IV Example Sodium Fast Reactor Using the GEN IV PR and PP Methodology

    Energy Technology Data Exchange (ETDEWEB)

    Zentner, M.D.; Coles, G.A. [PNNL, P.O. Box 999, 902 Battelle Boulvard, Richland, WA 99336 (United States); Therios, I.U. [Argonne National Lab. - ANL (United States)

    2009-06-15

    An experts working group was created in 2002 by The Generation IV International Forum for the purpose of developing an internationally accepted methodology for assessing the proliferation resistance of a nuclear energy system (NES) and its individual elements. A two year case study was performed by the working group using this methodology to assess the proliferation resistance of a hypothetical NES called the Example Sodium Fast Reactor (ESFR). This work demonstrates how the PR and PP methodology can be used to provide important information to designers at various levels of details, including pre-conceptual design stage. The study analyzes the response of the ESFR entire nuclear energy system to different proliferation and theft strategies. The challenges considered comprise concealed diversion, concealed misuse and abrogation strategies. This paper describes the work done in performing a qualitative assessment of potential concealed diversion scenarios from the ESFR, and includes an evaluation of the potential effect of changes in the conversion ratio on diversion strategies. (authors)

  14. Improvement of the decay heat removal characteristics of the generation IV gas-cooled fast reactor

    International Nuclear Information System (INIS)

    Epiney, A.S.

    2010-01-01

    Gas cooling in nuclear power plants (NPPs) has a long history, the corresponding reactor types developed in France, the UK and the US having been thermal neutron spectrum systems using graphite as the moderator. The majority of NPPs worldwide, however, are currently light water reactors, using ordinary water as both coolant and moderator. These NPPs - of the so-called second generation - will soon need replacement, and a third generation is now being made available, offering increased safety while still based on light water technology. For the longer-term future, viz. beyond the year 2030, R and D is currently ongoing on Generation IV NPPs, aimed at achieving closure of the nuclear fuel cycle, and hence both drastically improved utilization of fuel resources and minimization of long-lived radioactive wastes. Like the SFR, the GFR is an efficient breeder, also able to work as iso-breeder using simply natural uranium as feed and producing waste which is predominantly in the form of fission products. The main drawback of the GFR is the difficulty to evacuate decay heat following a loss-of-coolant accident (LOCA) due to the low thermal inertia of the core, as well as to the low coolant density. The present doctoral research focuses on the improvement of decay heat removal (DHR) for the Generation-IV GFR. The reference GFR system design considered in the thesis is the 2006 CEA concept, with a power of 2400 MWth. The CEA 2006 DHR strategy foresees, in all accidental cases (independent of the system pressure), that the reactor is shut down. For high pressure events, dedicated DHR loops with blowers and heat exchangers are designed to operate when the power conversion system cannot be used to provide acceptable core temperatures under natural convection conditions. For de-pressurized events, the strategy relies on a dedicated small containment (called the guard containment) providing an intermediate back-up pressure. The DHR blowers, designed to work under these pressure

  15. Safety Research Experiment Facility Project. Conceptual design report. Volume IV. Reactor containment

    International Nuclear Information System (INIS)

    1975-12-01

    The principal purpose of the SAREF Reactor Containment Building (RCB) is to prevent the uncontrolled release of radioactive materials to the atmosphere as a result of accidental occurrences inside the containment. The RCB houses numerous reactor systems and components including the Prestressed Concrete Reactor Vessel (PCRV). The design of the RCB is of reinforced concrete (steel-lined). The containment building is embedded nearly 100 feet in lava rock. It has therefore been necessary to independently formulate an appropriate and conservative design approach

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

    International Nuclear Information System (INIS)

    Rosenbalm, K.F.

    1995-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Rosenbalm, K.F. [comp.

    1995-12-31

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

  18. Establishing a safety and licensing basis for generation IV advanced reactors. License by test

    International Nuclear Information System (INIS)

    Kadak, Andrew C.

    2001-01-01

    The license by test approach to licensing is a novel method of licensing reactors. It provides an opportunity to deal with innovative non-water reactors in a direct way on a time scale that could permit early certification based on tests of a demonstration reactor. The uncertainties in the design and significant contributors to risk would be identified in the PRA during the design. Deterministic analysis computer codes could be tested on a real reactor. Scaling effects and associated uncertainties would be minimized. License by test is an approach that has sufficient merit to be developed and tested

  19. Reactor technology. Progress report, January--March 1978

    International Nuclear Information System (INIS)

    Warren, J.L.

    1978-07-01

    Progress is reported in eight program areas. The nuclear Space Electric Power Supply Program examined safety questions in the aftermath of the COSMOS 954 incident, examined the use of thermoelectric converters, examined the neutronic effectiveness of various reflecting materials, examined ways of connecting heat pipes to one another, studied the consequences of the failure of one heat pipe in the reactor core, and did conceptual design work on heat radiators for various power supplies. The Heat Pipe Program reported progress in the design of ceramic heat pipes, new application of heat pipes to solar collectors, and final performance tests of two pipes for HEDL applications. Under the Nuclear Process Heat Program, work continues on computer codes to model a pebble bed high-temperature gas-cooled reactor, adaptation of a set of German reactor calculation codes to use on U.S. computers, and a parametric study of a certain resonance integral required in reactor studies. Under the Nonproliferation Alternative Sources Assessment Program LASL has undertaken an evaluation of a study of gaseous core reactors by Southern Science Applications, Inc. Independently LASL has developed a proposal for a comprehensive study of gaseous uranium-fueled reactor technology. The Plasma Core Reactor Program has concentrated on restacking the beryllium reflector and redesigning the nuclear control system. The status of and experiments on four critical assemblies, SKUA, Godiva IV, Big Ten, and Flattop, are reported. The Nuclear Criticality Safety Program carried out several tasks including conducting a course, doing several annual safety reviews and evaluating the safety of two Nevada test devices. During the quarter one of the groups involved in reactor technology has acquired responsibility for the operation of a Cockroft-Walton accelerator. The present report contains information on the use of machine and improvements being made in its operation

  20. Belgian Contribution to the IAEA CRP-IV Programme on Assuring Structural Integrity of Reactor Pressure Vessel

    International Nuclear Information System (INIS)

    Van Walle, E.; Chaouadi, R.; Scibetta, M.; Puzzolante, J.L.; Fabry, A.; Van de Velde, J.

    1997-10-01

    This report contains the actual status of the Belgian contribution to the IAEA CRP-IV program. Besides Charpy-V impact tests on as-received CRP-IV JRQ-specimens, fracture toughness tests were performed on two geometries: PCCV-specimens and CRB-specimens. The Charpy-V impact results correspond very well with the as-received CRP-III results. The fracture toughness data are also very consistent with identical tests recently performed on remaining as-received CRP-III material. Irradiated broken Charpy-V samples were reconstituted and tested in PCCV-mode. This was done in order to investigate the evolution of the ASME-curve versus the evolution of the mastercurve with irradiation. Initial results were reported. A new CHIVAS-irradiation in the CALLISTO-loop of the BR-2-reactor to support this investigation, is under preparation

  1. Safety Design Criteria (SDC) for Gen-IV Sodium-cooled Fast Reactor

    International Nuclear Information System (INIS)

    Nakai, Ryodai

    2013-01-01

    SDC Development Background & Objectives: • Safety Design Criteria (SDC) Development for Gen-IV SFR: – Proposed at the GIF Policy Group (PG) meeting in October 2010 –SDC “harmonization” is increasingly important for: • Realization of enhanced safety designs meeting to Gen-IV safety goals and safety approach common to SFR systems; • Preparation for the forthcoming licensing in the near future; • Because Gen-IV SFR are progressing into conceptual design stage. • The SDC is the Reference criteria: – Of the designs of safety-related Structures, Systems & Components that are specific to the SFR system; – For clarifying the requisites systematically & comprehensively; – When the technology developers apply the basic safety approach and use the codes & standards for conceptual design of the Gen-IV SFR system

  2. A Qualitative Assessment of Diversion Scenarios for an Example Sodium Fast Reactor Using the GEN IV PR and PP Methodology

    International Nuclear Information System (INIS)

    Zentner, Michael D.; Coles, Garill A.; Therios, Ike

    2012-01-01

    FAST REACTORS;NUCLEAR ENERGY;NUCLEAR MATERIALS MANAGEMENT;PROLIFERATION;SAFEGUARDS;THEFT; A working group was created in 2002 by the Generation IV International Forum (GIF) for the purpose of developing an internationally accepted methodology for assessing the Proliferation Resistance of a nuclear energy system (NES) and its individual elements. A two year case study is being performed by the experts group using this methodology to assess the proliferation resistance of a hypothetical NES called the Example Sodium Fast Reactor (ESFR). This work demonstrates how the PR and PP methodology can be used to provide important information at various levels of details to NES designers, safeguard administrators and decision makers. The study analyzes the response of the complete ESFR nuclear energy system to different proliferation and theft strategies. The challenges considered include concealed diversion, concealed misuse and 'break out' strategies. This paper describes the work done in performing a qualitative assessment of concealed diversion scenarios from the ESFR.

  3. Study of a fuel assembly for the nuclear reactor of IV generation cooled with supercritical water

    International Nuclear Information System (INIS)

    Barragan M, A.; Martin del Campo M, C.; Francois L, J. L.; Espinosa P, G.

    2011-11-01

    In this work a neutron study is presented about a square assembly design of double line of fuel rods, with moderator box to the center of the arrangement, for a nuclear reactor cooled with supercritical water (SCWR). The SCWR reactor was chosen by the characteristics of its design, mainly because is based in light water reactors (PWR and BWR), and the operational experience that has of them allow to use models and similar programs to simulate the fuel and the nucleus of this type of reactors. To develop the necessary models and to carry out the design and analysis of the SCWR reactor, the neutron codes MCNPX and Helios were used. The reason of using both codes, is because the code MCNPX used thoroughly in the neutron simulation of these reactors, it has been our reference code to analyze the results obtained with the Helios code which results are more efficient because its calculation times are minors. In the nucleus design the same parameters for both codes were considered. The results show that the design with Helios is a viable option to simulate these reactors since their values of the neutrons multiplication factor are very similar to those obtained with MCNPX. On the other hand, it could be corroborated that the CASMO-4 code is inadequate to simulate the fuel to the temperature conditions and water pressure in the SCWR. (Author)

  4. Physical aspects of the Canadian generation IV supercritical water-cooled pressure tube reactor plant design

    Energy Technology Data Exchange (ETDEWEB)

    Gaudet, M.; Yetisir, M.; Haque, Z. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

    2014-07-01

    The form of the containment building is a function of the requirements imposed by various systems. In order to provide sufficient driving force for naturally-circulated emergency cooling systems, as well as providing a gravity-driven core flooding pool function, the Canadian SCWR reactor design relies on elevation differences between the reactor and the safety systems. These elevation differences, the required cooling pool volumes and the optimum layout of safety-related piping are major factors influencing the plant design. As a defence-in-depth, the containment building and safety systems also provide successive barriers to the unplanned release of radioactive materials, while providing a path for heat flow to the ultimate heat sink, the atmosphere. Access to the reactor for refuelling is from the top of the reactor, with water used as shielding during the refuelling operations. The accessibility to the reactor and protection of the environment are additional factors influencing the plant design. This paper describes the physical implementation of the major systems of the Canadian SCWR within the reactor building, and the position of major plant services relative to the reactor building. (author)

  5. JSFR design progress related to development of safety design criteria for Generation IV sodium-cooled fast reactors. (1) Overview

    International Nuclear Information System (INIS)

    Kamide, Hideki; Ando, Masato; Ito, Takaya

    2015-01-01

    JAEA, JAPC and MFBR have been conducting design study for the Japan Sodium-cooled Fast Reactor (JSFR), which is a design concept aiming at future commercial use as sustainable electric power source. As the result of the design study and R and D activity related the innovative technologies incorporated in the design in the Fast Reactor Cycle Technology Development (FaCT) project up to 2010, basic design concept of JSFR was established and its development process to the commercialization including construction and operation of a demonstration version of JSFR was outlined. JSFR is a looptype next generation sodium-cooled fast reactor (SFR), which is aiming at achieving development targets of Generation IV reactors concerning sustainability, safety and reliability, economics and proliferation resistance and physical protection by introducing the innovative technologies such as shortened high-chromium steel piping. The output power is assumed for the design study as 1,500 MWe for the commercial version and 750 MWe for the demonstration version. In FaCT phase I up to 2010, in order to evaluate feasibility to achieve the development targets, the design study has been conducted on the main components and systems. Since 2011, in order to contribute to the development of safety design criteria (SDC) and safety design guideline (SDG), which include the lessons learned from the TEPCO's Fukushima Dai-ichi nuclear power plants accident, in the frame work of Generation IV International Forum (GIF), the design study is focusing on the design measures against severe external events such as earthquake and tsunami. At the same time, the design study is going into detail and paying much attention to the maintenance and repair to make surer its feasibility. This paper summarizes the design concept of the demonstration version of JSFR in which progress of design work was incorporated for the safety issues on SDC and SDG of a SFR. (author)

  6. 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)

  7. Drop performance test of conceptually designed control rod assembly for prototype generation IV sodium-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Young Kyu; Lee, Jae Han; Kim, Hoe Woong; KIm, Sung Kyun; Kim, Jong Bum [Sodium-cooled Fast Reactor NSSS Design Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2017-06-15

    The control rod assembly controls reactor power by adjusting its position during normal operation and shuts down chain reactions by its free drop under scram conditions. Therefore, the drop performance of the control rod assembly is important for the safety of a nuclear reactor. In this study, the drop performance of the conceptually designed control rod assembly for the prototype generation IV sodium-cooled fast reactor that is being developed at the Korea Atomic Energy Research Institute as a next-generation nuclear reactor was experimentally investigated. For the performance test, the test facility and test procedure were established first, and several free drop performance tests of the control rod assembly under different flow rate conditions were then carried out. Moreover, performance tests under several types and magnitudes of seismic loading conditions were also conducted to investigate the effects of seismic loading on the drop performance of the control rod assembly. The drop time of the conceptually designed control rod assembly for 0% of the tentatively designed flow rate was measured to be 1.527 seconds, and this agrees well with the analytically calculated drop time. It was also observed that the effect of seismic loading on the drop time was not significant.

  8. Service for maintenance of electronic equipment of the RA reactor, Report for 1977 - Annex IV

    International Nuclear Information System (INIS)

    Milosevic, M.

    1977-01-01

    This report describes the tasks and organizational structure of the Service for maintenance of electronic equipment of the RA reactor. The most important task of this Service is control and maintenance of the reactor control and protection instruments, operation control, and dosimetry system. Besides data about this basic instrumentation, the report includes data about control and maintenance of other electronic equipment related to the experimental devices [sr

  9. Solution properties of a heteropolysaccharide extracted from pumpkin (Cucurbita pepo, lady godiva).

    Science.gov (United States)

    Song, Yi; Zhao, Jing; Ni, Yuanying; Li, Quanhong

    2015-11-05

    A water-soluble galactoglucofucomannan was extracted from pumpkin (Cucurbita pepo, lady godiva variety). GC-MS analysis indicated that the polysaccharide was composed of 1,6-linked-glucosyl, 1,2,6-linked-mannosyl, 1,3,6-linked-mannosyl, 1,2,6-linked-galactosyl, 1,2,6-linked-galactosyl, terminal fucosyl and terminal glucose. The solution properties of the polysaccharide were studied systematically by using size-exclusion chromatography combined with multi-angle laser light scattering, viscometry and dynamic light scattering at 25 °C. The weight average molecular masses (Mw), intrinsic viscosity [η], radius of gyration (Rg) and hydrodynamic radius (Rh) were found to be 12.7 × 10(5)g/mol, 780 ml/g, 68 nm and 116 nm, respectively. The fraction dimension and value of ρ (Rg/Rh) of the polysaccharide revealed that it existed in a sphere-like conformation in distilled water. The dependence of zero shear specific viscosity on the coil overlap parameter was analyzed using different models. Furthermore, degradation of samples upon autoclaving has been observed and quantified by intrinsic viscosity determination and SEC-MALLS. Copyright © 2015 Elsevier Ltd. All rights reserved.

  10. Millimeter-Wave Thermal Analysis Development and Application to GEN IV Reactor Materials

    Energy Technology Data Exchange (ETDEWEB)

    Wosko, Paul; Sundram, S. K.

    2012-10-16

    New millimeter-wave thermal analysis instrumentation has been developed and studied for characterization of materials required for diverse fuel and structural needs in high temperature reactor environments such as the Next Generation Nuclear Plant (NGNP). A two-receiver 137 GHz system with orthogonal polarizations for anisotropic resolution of material properties has been implemented at MIT. The system was tested with graphite and silicon carbide specimens at temperatures up to 1300 ºC inside an electric furnace. The analytic and hardware basis for active millimeter-wave radiometry of reactor materials at high temperature has been established. Real-time, non contact measurement sensitivity to anisotropic surface emissivity and submillimeter surface displacement was demonstrated. The 137 GHz emissivity of reactor grade graphite (NBG17) from SGL Group was found to be low, ~ 5 %, in the 500 – 1200 °C range and increases by a factor of 2 to 4 with small linear grooves simulating fracturing. The low graphite emissivity would make millimeter-wave active radiometry a sensitive diagnostic of graphite changes due to environmentally induced stress fracturing, swelling, or corrosion. The silicon carbide tested from Ortek, Inc. was found to have a much higher emissivity at 137 GHz of ~90% Thin coatings of silicon carbide on reactor grade graphite supplied by SGL Group were found to be mostly transparent to millimeter-waves, increasing the 137 GHz emissivity of the coated reactor grade graphite to about ~14% at 1250 ºC.

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

    International Nuclear Information System (INIS)

    Breslow, M.

    1980-12-01

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

  12. Eugene P. Wigner's Visionary Contributions to Generations-I through IV Fission Reactors

    Science.gov (United States)

    Carré, Frank

    2014-09-01

    Among Europe's greatest scientists who fled to Britain and America in the 1930s, Eugene P. Wigner made instrumental advances in reactor physics, reactor design and technology, and spent nuclear fuel processing for both purposes of developing atomic weapons during world-war II and nuclear power afterwards. Wigner who had training in chemical engineering and self-education in physics first gained recognition for his remarkable articles and books on applications of Group theory to Quantum mechanics, Solid state physics and other topics that opened new branches of Physics.

  13. Eugene P. Wigner’s Visionary Contributions to Generations-I through IV Fission Reactors

    Directory of Open Access Journals (Sweden)

    Carré Frank

    2014-01-01

    Full Text Available Among Europe’s greatest scientists who fled to Britain and America in the 1930s, Eugene P. Wigner made instrumental advances in reactor physics, reactor design and technology, and spent nuclear fuel processing for both purposes of developing atomic weapons during world-war II and nuclear power afterwards. Wigner who had training in chemical engineering and self-education in physics first gained recognition for his remarkable articles and books on applications of Group theory to Quantum mechanics, Solid state physics and other topics that opened new branches of Physics.

  14. RB Research nuclear reactor, Annual report for 1995, I-IV; Istrazivacki nuklearni reaktor RB, Izvestaj o radu u 1995. godini, I-IV

    Energy Technology Data Exchange (ETDEWEB)

    Stefanovic, D; Milosevic, M; Pesic, M [Institute of Nuclear Sciences Vinca, Belgrade (Yugoslavia); Marinkovic, P [Elektrotehnicki fakultet, Beograd (Yugoslavia); Ilic, R; Dasic, N; Milovanovic, S; Ljubenov, V; Petronijevic, M; Jevremovic, M [Institute of Nuclear Sciences Vinca, Belgrade (Yugoslavia)

    1995-12-15

    Report on RB reactor operation during 1995 contains 3 parts. Part one contains a brief description of reactor operation and reactor components, relevant dosimetry data and radiation protection issues, personnel and financial data. Part two is devoted to maintenance of the reactor components, namely, fuel, heavy water, reactor vessel, heavy water circulation system, absorption rods and heavy water level-meters, maintenance of electronic, mechanical, electrical and auxiliary equipment. Part three contains data concerned with reactor operation and utilization with a comprehensive list of publications resulting from experiments done at the RB reactor.

  15. On a fast numerical tool for nuclear accidental dynamic phenomenology and application to the real time simulation of Lady Godiva

    International Nuclear Information System (INIS)

    Bindel, Laurent; Gamess, Andre; Jasserand, Frederic; Laporte, Sebastien

    2003-01-01

    This paper present a modern numerical method, implemented in a TUI-code named MacDSP, for solving any set of differential equations and in particular phenomenological accidental dynamic calculations. The speed efficiency of such an approach, thanks to the use of the hybrid-level power offered by C++ and an ad-hoc design, make it possible to construct the first kid of a family of real time simulator employing the video games technology DirectX TM : Lady Godiva Real Time Simulator. (author)

  16. Minor actinides transmutation potential: state of art for GEN IV sodium cooled fast reactors

    International Nuclear Information System (INIS)

    Buiron, Laurent

    2015-01-01

    In the frame of the R and D program relative to the 1991 French act on nuclear waste management, fast neutron systems have shown relevant characteristics that meet both requirements on sustainable resources management and waste minimization. They also offer flexibility by mean of burner or breeder configurations allowing mastering plutonium inventory without significant impact on core safety. From the technological point of view, sodium cooled fast reactor are considered in order to achieve mean term industrial deployment. The present document summaries the main results of R and D program on minor actinides transmutation in sodium fast reactor since 2006 following recommendation of the first part of the 1991 French act. Both homogeneous and heterogeneous management achievable performances are presented for 'evolutionary' SFR V2B core as well as low void worth CFV core for industrial scale configurations (1500 MWe). Minor actinides transmutation could be demonstrated in the ASTRID reactor with the following configurations: - a 2%vol Americium content for the homogeneous mode, - a 10%vol Americium content for the heterogeneous mode, without any substantial modification of the main core safety parameters and only limited impacts on the associated fuel cycle (manufacturing issues are not considered here). In order to achieve such goal, a wide range of experimental irradiations driven by transmutation scenarios have to be performed for both homogeneous and heterogeneous minor actinides management. (author) [fr

  17. A Qualitative Assessment Of Diversion Scenarios For A Example Sodium Fast Reactor Using The Gen IV PR And PP Methodology

    International Nuclear Information System (INIS)

    Zentner, Michael D.

    2008-01-01

    A working group was created in 2002 by the Generation IV International Forum (GIF) for the purpose of developing an internationally accepted methodology for assessing the Proliferation Resistance of a nuclear energy system (NES) and its individual elements. A two year case study is being performed by the experts group using this methodology to assess the proliferation resistance of a hypothetical NES called the Example Sodium Fast Reactor (ESFR). This work demonstrates how the PR and PP methodology can be used to provide important information at various levels of details to NES designers, safeguard administrators and decision makers. The study analyzes the response of the complete ESFR nuclear energy system to different proliferation and theft strategies. The challenges considered include concealed diversion, concealed misuse and 'break out' strategies. This paper describes the work done in performing a qualitative assessment of concealed diversion scenarios from the ESFR.

  18. Conversion of actinide solutions for the production of MA bearing fuels for Gen IV fast reactor systems

    International Nuclear Information System (INIS)

    Fernandez, A.; McGinley, J.; Somers, J.

    2008-01-01

    The conversion of the solution to solid for fuels containing minor actinides for accelerator driven systems or Gen IV fast reactors cannot be made by conventional ammonia or oxalate precipitation as is the case in today's reprocessing plant. The small particle size and concomitant dust that is produced in subsequent processing steps will not permit use of these processes on industrial scale. Innovation is needed to avoid dust generating powders, and indeed to simplify the processes themselves. Two such processing routes have been developed at the JRC-ITU. The sol gel route has been used to produce fuel containing Am and Np for the SUPERFACT, TRABANT and other irradiation experiments. The infiltration process has also been established and fuels have been produced for the FUTURIX and HELIOS experiments. (authors)

  19. Conversion of actinide solutions for the production of MA bearing fuels for Gen IV fast reactor systems

    Energy Technology Data Exchange (ETDEWEB)

    Fernandez, A.; McGinley, J.; Somers, J. [European Commission, Joint Research Centre, Institute for Transuranium Elements P.O.Box 2340, Karlsruhe, D-76125 (Germany)

    2008-07-01

    The conversion of the solution to solid for fuels containing minor actinides for accelerator driven systems or Gen IV fast reactors cannot be made by conventional ammonia or oxalate precipitation as is the case in today's reprocessing plant. The small particle size and concomitant dust that is produced in subsequent processing steps will not permit use of these processes on industrial scale. Innovation is needed to avoid dust generating powders, and indeed to simplify the processes themselves. Two such processing routes have been developed at the JRC-ITU. The sol gel route has been used to produce fuel containing Am and Np for the SUPERFACT, TRABANT and other irradiation experiments. The infiltration process has also been established and fuels have been produced for the FUTURIX and HELIOS experiments. (authors)

  20. Theoretical analysis of nuclear reactors (Phase I), I-V, Part IV, Nuclear fuel depletion; Razrada metoda teorijske analize nuklearnih reaktora (I faza) I-V, IV Deo, Promena izotopnog sastava goriva

    Energy Technology Data Exchange (ETDEWEB)

    Pop-Jordanov, J [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1962-07-15

    Nuclear fuel depletion is analyzed in order to estimate the qualitative and quantitative fuel property changes during irradiation and the influence of changes on the reactivity during long-term reactor operation. The changes of fuel properties are described by changes of neutron absorption and fission cross sections. Part one of this report covers the economic significance of fuel burnup and the review of fuel isotopic changes during depletion. Pat two contains the analysis of the U{sup 235} chain, analytical expressions for the concentrations of U{sup 235}, U{sup 236} and Np{sup 237} as a function of burnup. Part three contains the analysis of neutron spectrum influence on the Westcott method for calculating the cross sections. Part four contains the calculation method applied on Calder Hall type reactor. The results were obtained by applying ZUSE-22 R digital computer.

  1. Economic Analysis on Direct Use of Spent Pressurized Water Reactor Fuel in CANDU Reactors - IV: DUPIC Fuel Cycle Cost

    International Nuclear Information System (INIS)

    Ko, Won Il; Choi, Hangbok; Yang, Myung Seung

    2001-01-01

    This study examines the economics of the DUPIC fuel cycle using unit costs of fuel cycle components estimated based on conceptual designs. The fuel cycle cost (FCC) was calculated by a deterministic method in which reference values of fuel cycle components are used. The FCC was then analyzed by a Monte Carlo simulation to get the uncertainty of the FCC associated with the unit costs of the fuel cycle components. From the deterministic analysis on the equilibrium fuel cycle model, the DUPIC FCC was estimated to be 6.21 to 6.34 mills/kW.h for DUPIC fuel options, which is a little smaller than that of the once-through FCC by 0.07 to 0.27 mills/kW.h. Considering the uncertainty (0.40 to 0.44 mills/kW.h) of the FCC estimated by the Monte Carlo simulation method, the cost difference between the DUPIC and once-through fuel cycle is negligible. On the other hand, the material balance calculation has shown that the DUPIC fuel cycle can save natural uranium resources by ∼20% and reduce the spent fuel arising by ∼65% compared with the once-through fuel cycle. In conclusion, the DUPIC fuel cycle is comparable with the once-through fuel cycle from the viewpoint of FCC. In the future, it should be important to consider factors such as the environmental benefit owing to natural uranium savings, the capability of reusing spent pressurized water reactor fuel, and the safeguardability of the fuel cycle when deciding on an advanced nuclear fuel cycle option

  2. First wall/blanket/shield design and power conversion for the ARIES-IV tokamak fusion reactor

    International Nuclear Information System (INIS)

    Hasan, M.Z.; Conn, R.W.; Najmabadi, F.

    1994-01-01

    ARIES-IV is a conceptual, D-T burning, steady-state tokamak fusion reactor producing 1000 MWe net. It operates in the second plasma stability regime. The structural material is SiC composite and the primary coolant is helium at 10 MPa base pressure. The coolant flows poloidally in two loops, one inboard and one outboard. The coolant channels are circular tubes that form shells and are placed between two purge plates; the space between two adjacent tubes and the plate is purge gas flow area. The solid breeder is Li 2 O, and Be is used as neutron multiplier to ensure adequate TBR. Beryllium and Li 2 O are placed in between the adjacent tube shells. A computer code was developed to perform and optimize thermal-hydraulic design. Minimization of blanket thickness and the amount of Be, and the maximization of breeder zone thickness were done by iteration with neutronics. The gross thermal efficiency is 49%. The cost of electricity is 68 mills/kWh. The use of low activation SiC composite as the structural material, Li 2 O as the solid breeder, and avoidance of tungsten in the divertor has resulted in a good safety performance, and LSA rating of 1. Overall, SiC/He/Li 2 O ARIES-IV design is expected to have attractive economic and safety advantages

  3. The safety R and D for GEN-IV reactors in the European nuclear energy technology platform strategic research agenda

    International Nuclear Information System (INIS)

    Bruna, G.

    2009-01-01

    In the fall 2007 EC launched the Sustainable Nuclear Energy Technology Platform (SNE-TP). The SNE-TP governing board set-up three working groups (WG): 1) Strategic Research Agenda (SRA) WG, in charge of drafting road-maps to support research, development and demonstration for current and future NPPs; 2) Deployment Strategy (DS) WG, in charge of defining the research road-map implementation and 3) Education, Training and Knowledge management (ETKM) WG, which was aimed at issuing proposal to reinforce European education and attract young in the nuclear field. The SRA WG was mandated to prepare the SRA vision document based on the preliminary road-map sketched in the document published by the Commission earlier in 2007. The SRA WG was originally organized in 5 sub-groups covering specific topics (1) GEN II and III, III+, including Advanced LWR, 2) Advanced Fuel Cycle for waste minimization and resource optimization; 3) GEN IV Fast Systems (SFR, LFR, GFR, ADS); 4) GEN IV (V) HTR and non-electricity-production applications; 5) New Nuclear Large Research Infrastructures) and 5 other sub-groups dealing with more generic cross-cutting research activities applicable to many specific topics, namely: 1) Structural material research; 2) modeling, simulation and methods, including physical data and tools and means for qualification and validation; 3) Reactor Safety, including severe accidents and human factor; 4) Advanced Driver and Minor Actinide Fuels: science and properties; 5) Pre-normative Research, Codes and Standards.The present paper is mainly aimed at summarizing the content of the SRA Safety sub-chapter focusing on GEN-IV aspects

  4. Reprocessing techniques of LWR spent fuel for reutilization in hybrid systems and IV generation reactors

    Energy Technology Data Exchange (ETDEWEB)

    Aruquipa, Wilmer; Velasquez, Carlos E.; Pereira, Claubia; Veloso, Maria Auxiliadora F.; Costa, Antonella L. [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Barros, Graiciany de P. [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

    2017-07-01

    Since the era of nuclear technology begins, nuclear reactors have been produced spent fuel. This spent fuel contains material that could be recycle and reprocessed by different processes. All these processes aim to reduce the contribution to the final repository through the re-utilization of the nuclear material. Therefore, some new reprocessing options with non-proliferation characteristics have been proposed and the goal is to compare the different techniques used to maximize the effectiveness of the spent fuel utilization and to reduce the volume and long-term radiotoxicity of high-level waste by irradiation with neutron with high energy such as the ones created in hybrid reactors. In order to compare different recovery methods, the cross sections of fuels are calculated with de MCNP code, the first set consists of thorium-232 spiked with the reprocessed material and the second set in depleted uranium that containing 4.5% of U-235 spiked with the reprocessed material; These sets in turn are compared with the cross section of the UO{sub 2} in order to evaluate the efficiency of the reprocessed fuel as nuclear fuel. (author)

  5. {sup 20}F power measurement for generation IV sodium fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Coulon, R.; Normand, S.; Michel, M.; Barbot, L.; Domenech, T.; Boudergui, K.; Bourbotte, J.M.; Kondrasovs, V.; Frelin-Labalme, A.M.; Hamrita, H. [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France); BAN, G. [ENSICAEN, 6 Boulevard Marechal Juin, F-14050 Caen Cedex 4 (France); Barat, E.; Dautremer, T.; Montagu, T.; Carrel, F. [CEA, LIST, Laboratoire Modelisation Simulation et Systemes, F-91191 Gif-sur-Yvette (France); Brau, H.P. [ICSM, Centre de Marcoule, BP 17171 F-30207 Bagnols sur Ceze (France); Dumarcher, V. [AREVA NP, SET, F-84500 Bollene (France); Portier, J.L. [Centrale PHENIX, Centre de Marcoule, Groupe Essais Statistiques, F-30207 Bagnols sur Ceze (France); Jousset, P. [CEA, LIST, Laboratoire Capteurs Diamant, F-91191 Gif-sur-Yvette (France); Saurel, N. [CEA, DAM, Laboratoire Mesure de Dechets et Expertise, F-21120 Is-sur-Tille, France.F-84500 Bollene (France)

    2010-07-01

    The Phenix nuclear power plant has been a French Sodium Fast Reactor (SFR) prototype producing electrical power between 1973 and 2010. The power was monitored using ex-core neutron measurements. This kind of measurement instantly estimates the power but needs to be often calibrated with the heat balance thermodynamic measurement. Large safety and security margins have then been set not to derive above the nominal operating point. It is important for future SFR to reduce this margin and working closer to the nominal operating point. This work deals with the use of delayed gamma to measure the power. The main activation product contained in the primary sodium coolant is the {sup 24}Na which is not convenient for neutron flux measurement due to its long decay period. The experimental study done at the Phenix reactor shows that the use of {sup 20}F as power tagging agent gives a fast and accurate power measurement closed to the thermal balance measurement thanks to its high energy photon emission (1.634 MeV) and its short decay period (11 s). (authors)

  6. IV Training program for the staff of the laboratory for the RA reactor exploitation; IV Programi obuke osoblja Laboratorije za eksploataciju reaktora RA

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1960-07-01

    All the staff members of the laboratory for RA reactor exploitation are obliged to learn the following: fundamental properties of the RA reactor, the role and functionality of the reactor components, basic and auxiliary reactor systems, basics of radioactivity, measures for preventing contamination. The personnel working in shifts must be acquainted with the regulations and instructions for reactor operation. Training programs for reactor operators, mechanics, electricians, instrumentators and dosimetrysts are described separately. Svi saradnici Laboratorije za eksploataciju reaktora RA moraju poznavati sledece oblasti: Osnovne karakeristike reaktora RA, princip rada, ulogu i funkcionisanje komponenti reaktora, osnovnih i pomocnih sistema reaktora; osnovne pojmove o radioaktivnom zracenju, mere za sprecavanje kontaminacije. Osoblje koje radi u smenama mora dodatno poznavati propise i uputstva za rad reaktora. Posebno je naveden program obuke operatora reaktora, mehanicara, electricara, instrumentatora, dozimetrista.

  7. Multi-layered silicides coating for vanadium alloys for generation IV reactors

    International Nuclear Information System (INIS)

    Mathieu, S.; Chaia, N.; Vilasi, M.; Le Flem, M.

    2012-01-01

    The halide-activated pack-cementation technique was employed to fabricate a diffusion coating that is resistant both to isothermal and to cyclic oxidation in air at 650 degrees C on the surface of the V-4Cr-4Ti vanadium alloy that is a potential core component of future nuclear systems. A thermodynamic assessment determined the deposit conditions in terms of master alloy, activator, filler and temperature. The partial pressures of the main gaseous species (SiCl 4 , SiCl 2 and VCl 2 ) in the pack were calculated with the master alloy Si and the mixture VSi 2 + Si. The VSi 2 + Si master alloy was used to limit vanadium loss from the surface. The obtained coating consisted of multi-layered V x Si y silicides with an outer layer of VSi 2 . This silicide developed a protective layer of silica at 650 degrees C in air and was not susceptible to the pest phenomenon, unlike other refractory silicides (MoSi 2 , NbSi 2 ). We suggest that VSi 2 exhibits no risk of rapid degradation in the gas fast reactor (GFR) conditions. (authors)

  8. RA Reactor operation and maintenance (I-IX), Part IV, Task 3.08/04, Refurbishment of the RA reactor; Pogon i odrzavanje reaktora RA (I-IX), IV Deo, Zadatak 3.08/04 Remont reaktora RA

    Energy Technology Data Exchange (ETDEWEB)

    Zecevic, V [Institute of Nuclear Sciences Boris Kidric, Reaktor RA, Vinca, Beograd (Serbia and Montenegro)

    1963-12-15

    This volume contains reports describing maintenance and repair work of the RA reactor instrumentation, equipment of the reactor dosimetry control system, and equipment for regulation and control systems.

  9. Summary of ORSphere Critical and Reactor Physics Measurements

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, Margaret A.; Bess, John D.

    2016-09-01

    In the early 1970s Dr. John T. Mihalczo (team leader), J. J. Lynn, and J. R. Taylor performed experiments at the Oak Ridge Critical Experiments Facility (ORCEF) with highly enriched uranium (HEU) metal (called Oak Ridge Alloy or ORALLOY) to recreate GODIVA I results with greater accuracy than those performed at Los Alamos National Laboratory in the 1950s. The purpose of the Oak Ridge ORALLOY Sphere (ORSphere) experiments was to estimate the unreflected and unmoderated critical mass of an idealized sphere of uranium metal corrected to a density, purity, and enrichment such that it could be compared with the GODIVA I experiments. This critical configuration has been evaluated. Preliminary results were presented at ND2013. Since then, the evaluation was finalized and judged to be an acceptable benchmark experiment for the International Criticality Safety Benchmark Experiment Project (ICSBEP). Additionally, reactor physics measurements were performed to determine surface button worths, central void worth, delayed neutron fraction, prompt neutron decay constant, fission density and neutron importance. These measurements have been evaluated and found to be acceptable experiments and are discussed in full detail in the International Handbook of Evaluated Reactor Physics Benchmark Experiments. The purpose of this paper is summary summarize all the critical and reactor physics measurements evaluations and, when possible, to compare them to GODIVA experiment results.

  10. Uranium enrichment reduction in the Prototype Gen-IV sodium-cooled fast reactor (PGSFR) with PBO reflector

    Energy Technology Data Exchange (ETDEWEB)

    Hartanto, Donny; Kim, Chi Hyung; Kim, Yong Hee [Dept. of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology (KAIST), Daejeon (Korea, Republic of)

    2016-04-15

    The Korean Prototype Gen-IV sodium-cooled fast reactor (PGSFR) is supposed to be loaded with a relatively-costly low-enriched U fuel, while its envisaged transuranic fuels are not available for transmutation. In this work, the U-enrichment reduction by improving the neutron economy is pursued to save the fuel cost. To improve the neutron economy of the core, a new reflector material, PbO, has been introduced to replace the conventional HT9 reflector in the current PGSFR core. Two types of PbO reflectors are considered: one is the conventional pin-type and the other one is an inverted configuration. The inverted PbO reflector design is intended to maximize the PbO volume fraction in the reflector assembly. In addition, the core radial configuration is also modified to maximize the performance of the PbO reflector. For the baseline PGSFR core with several reflector options, the U enrichment requirement has been analyzed and the fuel depletion analysis is performed to derive the equilibrium cycle parameters. The linear reactivity model is used to determine the equilibrium cycle performances of the core. Impacts of the new PbO reflectors are characterized in terms of the cycle length, neutron leakage, radial power distribution, and operational fuel cost.

  11. Effect of acetylation on antioxidant and cytoprotective activity of polysaccharides isolated from pumpkin (Cucurbita pepo, lady godiva).

    Science.gov (United States)

    Song, Yi; Yang, Yang; Zhang, Yuyu; Duan, Liusheng; Zhou, Chunli; Ni, Yuanying; Liao, Xiaojun; Li, Quanhong; Hu, Xiaosong

    2013-10-15

    Acetylation of pumpkin (Cucurbita pepo, lady godiva variety) polysaccharide using acetic anhydride with pyridines as catalyst under different conditions was conducted to obtain different degrees of acetylation on a laboratory scale. Furthermore, antioxidant activities and cytoprotective effects of pumpkin polysaccharide and its acetylated derivatives were investigated employing various established in vitro systems. Results showed that addition of pyridine as catalyst could increase the degree of substitution, whereas volume of acetic anhydride had little effect. The acetylated polysaccharides in DPPH scavenging radical activity assay, superoxide anion radical activity assay and reducing power assay exhibited higher antioxidant activity than that of unmodified polysaccharide. H2O2-induced oxidative damages on rat thymic lymphocyte were also prevented by pumpkin polysaccharide and its acetylated derivatives and the derivatives presented higher protective effects. On the whole, acetylated polysaccharide showed relevant antioxidant activity both in vitro and in a cell system. Copyright © 2013 Elsevier Ltd. All rights reserved.

  12. The G4-ECONS Economic Evaluation Tool for Generation IV Reactor Systems and its Proposed Application to Deliberately Small Reactor Systems and Proposed New Nuclear Fuel Cycle Facilities. Annex IX

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-12-15

    At the outset of the international Generation IV programme, it was decided that the six candidate reactor systems will ultimately be evaluated on the basis of safety, sustainability, non-proliferation attributes, technical readiness and projected economics. It is likely that the same factors will influence the evaluation of deliberately small reactor systems1 and new fuel cycle facilities, such as reprocessing plants that are being considered under the more recent Global Nuclear Energy Partnership (GNEP). This annex describes how the development of an economic modelling system has evolved to address the issue of economic competitiveness for both the Generation IV and GNEP programmes. In 2004, the Generation IV Economic Modelling Working Group (EMWG) commissioned the development of a Microsoft Excel based model capable of calculating the levelized unit electricity cost (LUEC) in mills/kW.h (1 mill = $10{sup -3}) or $/MW.h for multiple types of reactor system being developed under the Generation IV programme. This overall modelling system is now called the Generation IV spreadsheet calculation of nuclear systems (G4-ECONS), and is being expanded to calculate costs of energy products in addition to electricity, such as hydrogen and desalinated water. A version has also been developed to evaluate the costs of products or services from fuel cycle facilities. The cost estimating methodology and algorithms are explained in detail in the Generation IV Cost Estimating Guidelines and in the G4-ECONS User's Manual. The model was constructed with relatively simple economic algorithms such that it could be used by almost any nation without regard to country specific taxation, cost accounting, depreciation or capital cost recovery methodologies. It was also designed with transparency to the user in mind (i.e. all algorithms and cell contents are visible to the user). A short description of version 1.0 G4-ECONS-R (reactor economics model) has also been published in the

  13. The G4-ECONS Economic Evaluation Tool for Generation IV Reactor Systems and its Proposed Application to Deliberately Small Reactor Systems and Proposed New Nuclear Fuel Cycle Facilities. Annex IX

    International Nuclear Information System (INIS)

    2013-01-01

    At the outset of the international Generation IV programme, it was decided that the six candidate reactor systems will ultimately be evaluated on the basis of safety, sustainability, non-proliferation attributes, technical readiness and projected economics. It is likely that the same factors will influence the evaluation of deliberately small reactor systems1 and new fuel cycle facilities, such as reprocessing plants that are being considered under the more recent Global Nuclear Energy Partnership (GNEP). This annex describes how the development of an economic modelling system has evolved to address the issue of economic competitiveness for both the Generation IV and GNEP programmes. In 2004, the Generation IV Economic Modelling Working Group (EMWG) commissioned the development of a Microsoft Excel based model capable of calculating the levelized unit electricity cost (LUEC) in mills/kW.h (1 mill = $10 -3 ) or $/MW.h for multiple types of reactor system being developed under the Generation IV programme. This overall modelling system is now called the Generation IV spreadsheet calculation of nuclear systems (G4-ECONS), and is being expanded to calculate costs of energy products in addition to electricity, such as hydrogen and desalinated water. A version has also been developed to evaluate the costs of products or services from fuel cycle facilities. The cost estimating methodology and algorithms are explained in detail in the Generation IV Cost Estimating Guidelines and in the G4-ECONS User's Manual. The model was constructed with relatively simple economic algorithms such that it could be used by almost any nation without regard to country specific taxation, cost accounting, depreciation or capital cost recovery methodologies. It was also designed with transparency to the user in mind (i.e. all algorithms and cell contents are visible to the user). A short description of version 1.0 G4-ECONS-R (reactor economics model) has also been published in the Proceedings of

  14. Development and characterization of the control assembly system for the large 2400 MWth Generation IV gas-cooled fast reactor

    International Nuclear Information System (INIS)

    Girardin, G.; Rimpault, G.; Morin, F.; Bosq, J.C.; Coddington, P.; Mikityuk, K.; Chawla, R.

    2008-01-01

    The present paper is related to the design and neutronic characterization of the principal control assembly system for the reference large (2400 MWth) Generation IV gas-cooled fast reactor (GFR), which makes use of ceramic-ceramic (CERCER) plate-type fuel-elements with (U-Pu) carbide fuel contained within a SiC inert matrix. For the neutronic calculations, the deterministic code system ERANOS-2.0 has been used, in association with a full core model including a European fast reactor (EFR)-type pattern for the control assemblies as a starting point. More specifically, the core contains a total of 33 control (control system device: CSD) and safety (diverse safety device: DSD) assemblies implemented in three banks. In the design of the new control assembly system, particular attention was given to the heat generation within the assemblies, so that both neutronic and thermal-hydraulic constraints could be appropriately accounted for. The thermal-hydraulic calculations have been performed with the code COPERNIC, significant coolant mass flow rates being found necessary to maintain acceptable cladding temperatures of the absorber pins. Complementary to the design study, neutronic investigations have been performed to assess the impact of the control assemblies in the GFR core in greater detail (rod interactions, shift of the flux, peaking factors, etc.). Thus, considerable shadowing effects have been observed between the first bank and the safety bank, as also between individual assemblies within the first bank. Large anti-shadowing effects also occur, the most prominent being that between the two CSD banks, where the total assembly worth is almost doubled in comparison to the sum of the individual values. Additional investigations have been performed and, in this context, it has been found that computation of the first-order eigenvalue and the eigenvalue separation is a robust tool to anticipate control assembly interactions in a large fast-spectrum core. One interesting

  15. Study of diluting and absorber materials to control the reactivity during a postulated core meltdown accident in generation IV reactors

    International Nuclear Information System (INIS)

    Plevacova, Kamila

    2010-01-01

    In order to limit the consequences of a hypothetical core meltdown accident in Generation IV Sodium Fast Reactors, absorber materials in or near the core, such as boron carbide B 4 C, and diluting materials in the core catcher will be used to prevent recriticality within the mixture of molten oxide fuel and molten structures called corium. The aim of the PhD thesis was to select materials of both types and to understand their behaviour during their interaction with corium, from chemical and thermodynamic points of view. Concerning B 4 C, thermodynamic calculations and experiments agree with the formation of two immiscible phases at high temperature in the B 4 C - UO 2 system: one oxide and one boride. This separation of phases can reduce the efficiency of the neutrons absorption inside the molten fuel contained in the oxide phase. Moreover, volatilization of a part of the boron element can occur. According to these results, the necessary quantity of B 4 C to be introduced should be reconsidered for postulated severe accident sequence. Other solution could be the use of Eu 2 O 3 or HfO 2 as absorber material. These oxides form a solid solution with the oxide fuel. Concerning the diluting materials, mixed oxides Al 2 O 3 - HfO 2 and Al 2 O 3 - Eu 2 O 3 were preselected. These systems being completely unknown to date at high temperature in association with UO 2 , first points on the corresponding ternary phase diagrams were researched. Contrary to Al 2 O 3 - Eu 2 O 3 - UO 2 system, the Al 2 O 3 - HfO 2 - UO 2 mixture presents only one eutectic and thus only one solidification path which makes easier forecasting the behaviour of corium in the core catcher. (author)

  16. Study of diluting and absorber materials to control reactivity during a postulated core melt down accident in Generation IV reactors

    International Nuclear Information System (INIS)

    Plevacova, K.

    2010-01-01

    In order to limit the consequences of a hypothetical core meltdown accident in Generation IV Sodium Fast Reactors, absorber materials in or near the core, such as boron carbide B 4 C, and diluting materials in the core catcher will be used to prevent recriticality within the mixture of molten oxide fuel and molten structures called corium. The aim of the PhD thesis was to select materials of both types and to understand their behaviour during their interaction with corium, from chemical and thermodynamic point of view. Concerning B 4 C, thermodynamic calculations and experiments agree with the formation of two immiscible phases at high temperature in the B 4 C - UO 2 system: one oxide and one boride. This separation of phases can reduce the efficiency of the neutrons absorption inside the molten fuel contained in the oxide phase. Moreover, a volatilization of a part of the boron element can occur. According to these results, the necessary quantity of B 4 C to be introduced should be reconsidered for postulated severe accident sequence. Other solution could be the use of Eu 2 O 3 or HfO 2 as absorber material. These oxides form a solid solution with the oxide fuel. Concerning the diluting materials, mixed oxides Al 2 O 3 - HfO 2 and Al 2 O 3 - Eu 2 O 3 were preselected. These systems being completely unknown to date at high temperature in association with UO 2 , first points on the corresponding ternary phase diagrams were researched. Contrary to Al 2 O 3 - Eu 2 O 3 - UO 2 system, the Al 2 O 3 - HfO 2 - UO 2 mixture presents only one eutectic and thus only one solidification path which makes easier forecasting the behaviour of corium in the core catcher. (author) [fr

  17. Theoretical analysis of nuclear reactors (Phase III), I-V, Part III, Reactor poisoning; Razrada metoda teorijske analize nuklearnih reaktora (III faza) I-IV, III Deo, Zatrovanje reaktora

    Energy Technology Data Exchange (ETDEWEB)

    Pop-Jordanov, J [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1963-01-15

    Report on calculation of poisoning in experimental and power reactor includes four parts. Part one describes the influence of poisoning on the physical parameters of a reactor. part two includes transformation of differential equations for iodine and xenon. It was needed for easier solution of of differential equation using the analog computer. This calculation was done for RA reactor operating at 5 MW power. The RA reactor was used an example of calculation by the proposed method. Part four shows the application of the method for calculating the Calder Hall power reactor.

  18. Development of the control assembly pattern and dynamic analysis of the Generation IV large gas-cooled fast reactor (GFR)

    Energy Technology Data Exchange (ETDEWEB)

    Girardin, G.

    2009-07-15

    During the past ten years, different independent factors, such as the rapidly increasing worldwide demand in energy, societal concerns about greenhouse gas emissions, and the high and volatile prices for fossil fuels, have contributed to the renewed interest in nuclear technology. In this context, the Generation IV International Forum (GIF) launched the initiative to collaborate on the research and development efforts needed for the next generation of nuclear reactors. A particular goal set for Generation IV systems is closure of the nuclear fuel cycle; they are expected to offer a better utilization of natural resources, as also a minimization of long-lived radioactive wastes. Among the systems selected by the GIF, the Gas-cooled Fast Reactor (GFR) is a highly innovative system with advanced fuel geometry and materials. The principal aim of the present research is to develop and qualify the control assembly (CA) pattern and corresponding CA implementation scheme for the 2400 MWth reference GFR design. The work has been carried out in three successive phases: (1) validation of the neutronics tools, (2) the CA pattern development and related static analysis, and (3) dynamic core behaviour studies for hypothetical CA driven transients. The deterministic code system ERANOS and its associated nuclear data libraries for fast reactors were developed and validated for sodium-cooled reactors. In order to validate ERANOS for GFR applications, a systematic reanalysis of the GFR-relevant integral data generated at PSI during the GCFR-PROTEUS experimental program of the 1970’s was undertaken. The reference PROTEUS test lattice has been analyzed with ERANOS-2.0 and its associated, adjusted nuclear data library ERALIB1. Benchmark calculations were performed with the Monte Carlo code MCNPX, allowing one to both check the deterministic results and to analyze the sensitivity to different modern data libraries. For the main reaction rate ratios, the new analysis of the GCFR

  19. Development of the control assembly pattern and dynamic analysis of the Generation IV large gas-cooled fast reactor (GFR)

    International Nuclear Information System (INIS)

    Girardin, G.

    2009-07-01

    During the past ten years, different independent factors, such as the rapidly increasing worldwide demand in energy, societal concerns about greenhouse gas emissions, and the high and volatile prices for fossil fuels, have contributed to the renewed interest in nuclear technology. In this context, the Generation IV International Forum (GIF) launched the initiative to collaborate on the research and development efforts needed for the next generation of nuclear reactors. A particular goal set for Generation IV systems is closure of the nuclear fuel cycle; they are expected to offer a better utilization of natural resources, as also a minimization of long-lived radioactive wastes. Among the systems selected by the GIF, the Gas-cooled Fast Reactor (GFR) is a highly innovative system with advanced fuel geometry and materials. The principal aim of the present research is to develop and qualify the control assembly (CA) pattern and corresponding CA implementation scheme for the 2400 MWth reference GFR design. The work has been carried out in three successive phases: (1) validation of the neutronics tools, (2) the CA pattern development and related static analysis, and (3) dynamic core behaviour studies for hypothetical CA driven transients. The deterministic code system ERANOS and its associated nuclear data libraries for fast reactors were developed and validated for sodium-cooled reactors. In order to validate ERANOS for GFR applications, a systematic reanalysis of the GFR-relevant integral data generated at PSI during the GCFR-PROTEUS experimental program of the 1970’s was undertaken. The reference PROTEUS test lattice has been analyzed with ERANOS-2.0 and its associated, adjusted nuclear data library ERALIB1. Benchmark calculations were performed with the Monte Carlo code MCNPX, allowing one to both check the deterministic results and to analyze the sensitivity to different modern data libraries. For the main reaction rate ratios, the new analysis of the GCFR

  20. Multi-criteria methodology to design a sodium-cooled carbide-fueled Gen-IV reactor

    International Nuclear Information System (INIS)

    Stauff, N.

    2011-01-01

    Compared with earlier plant designs (Phenix, Super-Phenix, EFR), Gen IV Sodium-cooled Fast Reactor requires improved economics while meeting safety and non-proliferation criteria. Mixed Oxide (U-Pu)O 2 fuels are considered as the reference fuels due to their important and satisfactory feedback experience. However, innovative carbide (U-Pu)C fuels can be considered as serious competitors for a prospective SFR fleet since carbide-fueled SFRs can offer another type of optimization which might overtake on some aspects the oxide fuel technology. The goal of this thesis is to reveal the potentials of carbide by designing an optimum carbide-fueled SFR with competitive features and a naturally safe behavior during transients. For a French nuclear fleet, a 1500 MW(e) break-even core is considered. To do so, a multi-physic approach was developed taking into account neutronics, fuel thermo-mechanics and thermal-hydraulic at a pre-design stage. Simplified modeling with the calculation of global neutronic feedback coefficients and a quasi-static evaluation was developed to estimate the behavior of a core during overpower transients, loss of flow and/or loss of heat removal transients. The breakthrough of this approach is to provide the designer with an overall view of the iterative process, emphasizing the well-suited innovations and the most efficient directions that can improve the SFR design project.This methodology was used to design a core that benefits from the favorable features of carbide fuels. The core developed is a large carbide-fueled SFR with high power density, low fissile inventory, break-even capability and forgiving behaviors during the un-scrammed transients studied that should prevent using expensive mitigate systems. However, the core-peak burnup is unlikely to significantly exceed 100 MWd/kg because of the large swelling of the carbide fuel leading to quick pellet-clad mechanical interaction and the low creep capacity of carbide. Moderate linear power fuel

  1. Evaluation of a sodium-water reaction event caused by steam generator tubes break in the prototype generation IV sodium-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Sang June; Ha, Kwi Seok; Chang, Won Pyo; Kang, Seok Hun; Lee, Kwi Lim; Choi, Chi Woong; Lee, Seung Won; Yoo, Jin; Jeong, Jae Ho; Jeong, Tae Kyeong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-08-15

    The prototype generation IV sodium-cooled fast reactor (PGSFR) has been developed by the Korea Atomic Energy Research Institute. This reactor uses sodium as a reactor coolant to transfer the core heat energy to the turbine. Sodium has chemical characteristics that allow it to violently react with materials such as a water or steam. When a sodium–water reaction (SWR) occurs due to leakage or breakage of steam generator tubes, high-pressure waves and corrosive reaction products are produced, which threaten the structural integrity of the components of the intermediate heat-transfer system (IHTS) and the safety of the primary heat-transfer system (PHTS). In the PGSFR, SWR events are included in the design-basis event. This event should be analyzed from the viewpoint of the integrities of the IHTS and fuel rods. To evaluate the integrity of the IHTS based on the consequences of the SWR, the behaviors of the generated high-pressure waves are analyzed at the major positions of a failed IHTS loop using a sodium–water advanced analysis method-II code. The integrity of the fuel rods must be consistently maintained below the safety acceptance criteria to avoid the consequences of the SWR. The integrity of the PHTS is evaluated using the multidimensional analysis of reactor safety-liquid metal reactor code to model the whole plant.

  2. Status of the design and safety project for the sodium-cooled fast reactor as a generation IV nuclear energy system

    International Nuclear Information System (INIS)

    Niwa, Hajime; Fiorini, Gian-Luigi; Sim, Yoon-Sub; Lennox, Tom; Cahalan, James E.

    2005-01-01

    The Design and Safety Project Management Board (DSPMB) was established under the Sodium Cooled Fast Reactor (SFR) System Steering Committee (SSC) in the Generation IV international Forum. The DSPMB will promote collaborative R and D activities on reactor core design, and safety assessment for candidate systems, and also integrate these results together with those from other PMBs such as advanced fuel and component to a whole fast reactor system in order to develop high performance systems that will satisfy the goals of Generation IV nuclear energy systems. The DSPMB has formulated the present R and D schedules for this purpose. Two SFR concepts were proposed: a loop-type system with primarily a MOX fuel core and a pool-type system with a metal fuel core. Study of innovative systems and their evaluation will also be included. The safety project will cover both the safety assessment of the design and the preparation of the methods/tools to be used for the assessment. After a rather short viability phase, the project will move to the performance phase for development of performance data and design optimization of conceptual designs. This paper describes the schedules, work packages and tasks for the collaborative studies of the member countries. (author)

  3. JSFR design progress related to development of safety design criteria for generation IV sodium-cooled fast reactors. (3) Progress of component design

    International Nuclear Information System (INIS)

    Enuma, Yasuhiro; Kawasaki, Nobuchika; Orita, Junichi; Eto, Masao; Miyagawa, Takayuki

    2015-01-01

    In the frame work of generation IV international forum (GIF), safety design criteria (SDC) and safety design guideline (SDG) for the generation IV sodium-cooled fast reactors have been developing in the circumstance of worldwide deployment of SFRs. JAEA, JAPC, MFBR have been investigating design study for JSFR to satisfy SDC in the feasibility study of SDG for Sodium-cooled Fast Reactor (SFR). In addition to the safety measures, maintainability, reparability and manufacturability are taken into account in the JSFR design study. This paper describes the design of main components. Enlargement of the access route for the inspection devices and addition of the access routes were carried out for the reactor structure. The pump-integrated IHX (pump/IHX) was modified for the primary heat exchanger (PHX), which was installed for the decay heat removal in the IHX at the upper plenum, to be removable for improved repair and maintenance. For the steam generator (SG), protective wall tube type design is under investigation as an option with less R and D risks. (author)

  4. Study of a fuel assembly for the nuclear reactor of IV generation cooled with supercritical water; Estudio de un ensamble de combustible para el reactor nuclear de generacion IV enfriado con agua supercritica

    Energy Technology Data Exchange (ETDEWEB)

    Barragan M, A.; Martin del Campo M, C.; Francois L, J. L. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Mexico D. F. (Mexico); Espinosa P, G., E-mail: albrm29@yahoo.com [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Area de Ingenieria en Recursos Energeticos, Av. San Rafael Atlixco 186, Col. Vicentina, 09340 Mexico D. F. (MX)

    2011-11-15

    In this work a neutron study is presented about a square assembly design of double line of fuel rods, with moderator box to the center of the arrangement, for a nuclear reactor cooled with supercritical water (SCWR). The SCWR reactor was chosen by the characteristics of its design, mainly because is based in light water reactors (PWR and BWR), and the operational experience that has of them allow to use models and similar programs to simulate the fuel and the nucleus of this type of reactors. To develop the necessary models and to carry out the design and analysis of the SCWR reactor, the neutron codes MCNPX and Helios were used. The reason of using both codes, is because the code MCNPX used thoroughly in the neutron simulation of these reactors, it has been our reference code to analyze the results obtained with the Helios code which results are more efficient because its calculation times are minors. In the nucleus design the same parameters for both codes were considered. The results show that the design with Helios is a viable option to simulate these reactors since their values of the neutrons multiplication factor are very similar to those obtained with MCNPX. On the other hand, it could be corroborated that the CASMO-4 code is inadequate to simulate the fuel to the temperature conditions and water pressure in the SCWR. (Author)

  5. Theoretical analysis of nuclear reactors (Phase II), I-V, Part III, Reactor poisoning; Razrada metoda teorijske analize nuklearnih reaktora (II faza) I-V, III Deo, Zatrovanje reaktora, II faza

    Energy Technology Data Exchange (ETDEWEB)

    Pop-Jordanov, J [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1962-10-15

    This phase is dealing with influence of all the fission products except Xe{sup 135} on the reactivity of a reactor, usually named as reactor poisoning. The first part of the report is a review of methods for calculation of reactor poisoning. The second part shows the most frequently used method for calculation of cross sections and yields of pseudo products (for thermal neutrons). The system of equations was adopted dependent on the conditions of the available computer system. It is described in part three. Detailed method for their application is described in part four and results obtained are presented in part five.

  6. Theoretical analysis of nuclear reactors (Phase I), I-V, Part V, Determining the fine flux distribution

    International Nuclear Information System (INIS)

    Pop-Jordanov, J.

    1962-07-01

    Mono energetic neutron transport equation was solved by Carlson numerical method in cylindrical geometry. S n code was developed for the digital computer ZUSE Z23. Neutron flux distribution was determined for the RA reactor cell by applying S 4 approximation. Reactor cell was treated as D 2 O-U-D 2 O system. Time of iteration was 185 s [sr

  7. Unprotected Accident Analyses of the 1200MWe GEN-IV Sodium-Cooled Fast Reactor Using the SSC-K Code

    International Nuclear Information System (INIS)

    Kwon, Young Min; Lee, Kwi Lim; Ha, Kwi Seok; Jeong, Hae Yong; Chang, Won Pyo; Seok, Su Dong; Lee, Yong Bum

    2010-02-01

    A conceptual design of an advanced breakeven sodium-cooled fast reactor (G4SFR) has recently been developed by KAERI under the national nuclear R and D plan. The G4SFR is a 1,200MWe metal-fueled pool-type sodium-cooled fast reactor adopting advanced safety design features. The G4SFR development plan focuses on particular technology development efforts to effectively meet the goals of the Generation-IV (GEN-IV) nuclear system such as efficient utilization of resources, economic competitiveness, a high standard of safety, and enhanced proliferation resistance. To enhance the safety of G4SFR, advanced design features of metal-fueled core, simple and large sodium-inventory primary heat transport system, and passive safety decay heat removal system are included in the reactor design. To evaluate potential safety characteristics of such advanced design features, the plant responses and safety margins were investigated using the system transient code SSC-K for three unprotected accidents of UTOP, ULOF, and ULOHS. It was shown that the G4SFR design has inherent and passive safety characteristics and is accommodating the selected ATWS events. The inherent safety mechanism of the reactor design makes the core shutdown with sufficient margin and passive removal of decay heat with matching the core power to heat sink by passive self-regulation. The self-regulation of power without scram is mainly due to the inherent negative reactivity feedback in conjunction with the large thermal inertia of the primary heat transport system and the passive decay heat removal. Such favorable inherent and passive safety behaviors of G4SFR are expected to virtually exclude the probability of severe accidents with potential for core damage

  8. Activity of the Service for maintenance of the electronic equipment of the RA reactor in 1976, report - Annex IV

    International Nuclear Information System (INIS)

    Milosevic, M.

    1976-01-01

    Within the organizational structure of the RA reactor staff, the Service for instrumentation maintenance has the following tasks: maintenance of the existing electronic equipment; participating in experiments planning and preparation of electronic equipment; purchasing new equipment, spare parts and components; construction of new equipment for internal needs; implementation of new equipment. Basic instrumentation of the reactor facility includes: control and protection system, and dosimetry system [sr

  9. Analysis of criticality using NJOY, AMPX-II and KENO-IV systems

    International Nuclear Information System (INIS)

    Ferreira, C.R.; Santos, A. dos

    1989-01-01

    Trough a linking between the nuclear data code systems NJOY and AMPX-II, performed at Divisao de Fisica de Reatores-IPEN, it was made criticality analysis with the KENO-IV in two fast multiplying systems: JEZEBEL and GODIVA. The ENDF/B-IV and JENDL-2 was used throughout the analysis. It was found that the linking between NJOY and AMPX-II systems can be suitable for a benchmark calculation of neutron cross sections. The KENO-IV code cannot be applied in systems where the degree of anisotropic scattering is too high because this computer code treats scattering up to first order. The XSDRNPM results show that the results obtained with ENDF/B-IV are of better quality than those obtained with JENDL-2. (author) [pt

  10. Nuclear data processing for cross-sections generation for fusion-fission, ADS, and IV generation reactors utilization

    Energy Technology Data Exchange (ETDEWEB)

    Velasquez, Carlos E.; Fernandes, Lorena C.; Pereira, Claubia; Veloso, Maria Auxiliadora F.; Costa, Antonella L. [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

    2017-11-01

    One of the mains topics about nuclear reactors is the microscopic cross section for incident neutrons. Therefore, in this work, it is evaluated the microscopic and macroscopic cross section for a nuclide and a material. One of the nuclides microscopic cross-section studied is the {sup 56}Fe which is the highest compound from the material macroscopic cross section studied SS316. On the other hand, it was studied the microscopic cross section of the {sup 242}Pu which is one of the nuclides that composes the nuclear fuel. The nuclear fuel chosen is a spent fuel reprocessed by UREX+ technique and spiked with thorium with 20% of fissile material. Therefore it was studied the macroscopic cross section from this nuclear fuel. Both of them were compared by using three different ways to reprocess the nuclides, one for LWR, another for ADS and the last one for Fusion reactors. The library used was JEFF-3.2 recommend for the reactors studied. The comparison was made at 1200 K for the nuclear fuel and 700K for the SS316.The results present differences due to the energy discretization, the number of groups chosen for each reactor and some nuclear reactions taken into consideration according to the neutron spectrum for each reactor. The nuclides were processed by NJOY99.364 and plotted with MCNP-Vised. (author)

  11. Nuclear data processing for cross-sections generation for fusion-fission, ADS, and IV generation reactors utilization

    International Nuclear Information System (INIS)

    Velasquez, Carlos E.; Fernandes, Lorena C.; Pereira, Claubia; Veloso, Maria Auxiliadora F.; Costa, Antonella L.

    2017-01-01

    One of the mains topics about nuclear reactors is the microscopic cross section for incident neutrons. Therefore, in this work, it is evaluated the microscopic and macroscopic cross section for a nuclide and a material. One of the nuclides microscopic cross-section studied is the 56 Fe which is the highest compound from the material macroscopic cross section studied SS316. On the other hand, it was studied the microscopic cross section of the 242 Pu which is one of the nuclides that composes the nuclear fuel. The nuclear fuel chosen is a spent fuel reprocessed by UREX+ technique and spiked with thorium with 20% of fissile material. Therefore it was studied the macroscopic cross section from this nuclear fuel. Both of them were compared by using three different ways to reprocess the nuclides, one for LWR, another for ADS and the last one for Fusion reactors. The library used was JEFF-3.2 recommend for the reactors studied. The comparison was made at 1200 K for the nuclear fuel and 700K for the SS316.The results present differences due to the energy discretization, the number of groups chosen for each reactor and some nuclear reactions taken into consideration according to the neutron spectrum for each reactor. The nuclides were processed by NJOY99.364 and plotted with MCNP-Vised. (author)

  12. Thermodynamic Data to Model the Interaction Between Coolant and Fuel in Gen IV Sodium Cooled Fast Reactors

    International Nuclear Information System (INIS)

    Dinsdale, Alan; Gisby, John; Davies, Hugh; Konings, Rudy; Benes, Ondrej

    2013-06-01

    Understanding the behaviour of nuclear fuels in various environments is vital to the design and safe operation of nuclear reactors. While this is true if the reactor is operating within its design specification, it is even more so if accidents occur and the fuel is exposed to unexpected temperatures, pressures or chemical environments. It is clearly hazardous and costly to explore all such scenarios experimentally and therefore it is necessary to undertake modelling where possible using well-grounded theoretical approaches. This paper will show examples of where calculations of chemical and phase equilibria have been applied successfully to the long term storage of nuclear waste, phase formation during core meltdown and prediction of fission product release into the atmosphere. It will also highlight the development of thermodynamic data carried out during the European Metrology Research Project Metrofission required to model the potential interaction between the coolant, nuclear fuel, containment materials and atmosphere of a sodium cooled fast reactor. (authors)

  13. The ENEN-III project: Technical Training on the Concepts and Design of GEN IV nuclear reactors

    International Nuclear Information System (INIS)

    Berkvens, T.; Renault, C.; Alonso, M.; Salomaa, R.; Schönfelder, C.

    2013-01-01

    Some conclusions: • Not enough training courses to cover the LO’s: – Especially GEN IV; – Many introductory courses, little specific courses; – Reach out to other partners for more courses. • Skills and Attitudes: – Much more difficult to train/measure; – To be treated in a separate project. • Use of Learning Outcomes must be promoted; • Involvement of human resources necessary for the successful implementation of the schemes: – End of project workshop

  14. Fission product data for thermal reactors. Final report. Part I. A data set for EPRI-CINDER using ENDF/B-IV

    International Nuclear Information System (INIS)

    England, T.R.; Wilson, W.B.; Stamatelatos, M.G.

    1976-12-01

    A four-group fission-product neutron absorption library, appropriate for use in thermal reactors, is described. All decay parameters are taken from ENDF/B-IV. The absorption cross sections are also processed from ENDF/B-IV files, first into a 154-group set and subsequently collapsed into the 4-group set described in this report. The decay and cross section data were used to form 84 linear chains in the CINDER code format. These chains contain all significant fission products having half-lives exceeding 4 hours--a total of 186 nuclides. A 12-chain set containing one pseudo-chain for use in spatial depletion calculations is described. This set accurately reproduces the aggregate absorption buildup of the 84 chains. This report describes the chains and processed data, results of comparison calculations for various fuels, and a comparison of calculated temporal fission-product absorption buildup with corresponding results from a long-term fuel irradiation and cooling integral experiment

  15. Theoretical analysis of nuclear reactors (Phase I), I-V, Part III, Reactor poisoning; Razrada metoda teorijske analize nuklearnih reaktora (I faza) I-V, III Deo, Zatrovanje reaktora, I faza

    Energy Technology Data Exchange (ETDEWEB)

    Pop-Jordanov, J [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1962-07-15

    Method was developed for calculation of Xe{sup 135} static effect and kinetic effects of Xe{sup 135} and Sm{sup 149} with separate treatment of iodine effect and influence of reactor poisoning during power increase. Mentioned effects are treated first for uranium fuel and then the basic formulae were generalized for a mixture of fissile material. The annex contains a table with data needed for calculations and the Xe{sup 135} microscopic capture cross section dependent on temperature. Razradjen je metod proracuna statickog efekta Xe{sup 135} zatim kinetickog efekta Xe{sup 135} i Sm{sup 149} sa posebnim tretiranjem jodne jame i promene zatrovanja pri prelazu sa jedne snage na drugu. Navedeni efekti su tretirani prvo za uransko gorivo, a zatim su glavni obrasci uopsteni za smesu fisibilnih materijala. U prilogu su dati u vidu tabele, podaci potrebni za proracun i grafik zavisnosti mikroskopskog preseka zahvata Xe{sup 135} od temperature.

  16. On the safety and performance demonstration tests of Prototype Gen-IV Sodium-Cooled Fast Reactor and validation and verification of computational codes

    International Nuclear Information System (INIS)

    Kim, Jong Bum; Jeong, Ji Young; Lee, Tae Ho; Kim, Sung Kyun; Euh, Dong Jin; Joo, Hyung Kook

    2016-01-01

    The design of Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR) has been developed and the validation and verification (V and V) activities to demonstrate the system performance and safety are in progress. In this paper, the current status of test activities is described briefly and significant results are discussed. The large-scale sodium thermal-hydraulic test program, Sodium Test Loop for Safety Simulation and Assessment-1 (STELLA-1), produced satisfactory results, which were used for the computer codes V and V, and the performance test results of the model pump in sodium showed good agreement with those in water. The second phase of the STELLA program with the integral effect tests facility, STELLA-2, is in the detailed design stage of the design process. The sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger performance test, the intermediate heat exchanger test facility, and the test facility for the reactor flow distribution are underway. Flow characteristics test in subchannels of a wire-wrapped rod bundle has been carried out for safety analysis in the core and the dynamic characteristic test of upper internal structure has been performed for the seismic analysis model for the PGSFR. The performance tests for control rod assemblies (CRAs) have been conducted for control rod drive mechanism driving parts and drop tests of the CRA under scram condition were performed. Finally, three types of inspection sensors under development for the safe operation of the PGSFR were explained with significant results

  17. On the safety and performance demonstration tests of Prototype Gen-IV Sodium-Cooled Fast Reactor and validation and verification of computational codes

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jong Bum; Jeong, Ji Young; Lee, Tae Ho; Kim, Sung Kyun; Euh, Dong Jin; Joo, Hyung Kook [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The design of Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR) has been developed and the validation and verification (V and V) activities to demonstrate the system performance and safety are in progress. In this paper, the current status of test activities is described briefly and significant results are discussed. The large-scale sodium thermal-hydraulic test program, Sodium Test Loop for Safety Simulation and Assessment-1 (STELLA-1), produced satisfactory results, which were used for the computer codes V and V, and the performance test results of the model pump in sodium showed good agreement with those in water. The second phase of the STELLA program with the integral effect tests facility, STELLA-2, is in the detailed design stage of the design process. The sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger performance test, the intermediate heat exchanger test facility, and the test facility for the reactor flow distribution are underway. Flow characteristics test in subchannels of a wire-wrapped rod bundle has been carried out for safety analysis in the core and the dynamic characteristic test of upper internal structure has been performed for the seismic analysis model for the PGSFR. The performance tests for control rod assemblies (CRAs) have been conducted for control rod drive mechanism driving parts and drop tests of the CRA under scram condition were performed. Finally, three types of inspection sensors under development for the safe operation of the PGSFR were explained with significant results.

  18. 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.)

  19. NUCLEAR ENERGY RESEARCH INITIATIVE (NERI) PROGRAM GRANT NUMBER DE-FG03-00SF22168 TECHNICAL PROGRESS REPORT (Aug 15, 2002 to Nov. 15, 2002) - DESIGN AND LAYOUT CONCEPTS FOR COMPACT, FACTORY-PRODUCED, TRANSPORTABLE GENERATION IV REACTOR SYSTEMS

    International Nuclear Information System (INIS)

    Fred R. Mynatt; Andy Kadak; Marc Berte; Larry Miller; Lawrence Townsend; Martin Williamson; Rupy Sawhney; Jacob Fife

    2002-01-01

    The objectives of this project are to develop and evaluate nuclear power plant designs and layout concepts to maximize the benefits of compact modular Generation IV reactor concepts including factory fabrication and packaging for optimal transportation and siting. This report covers the ninth quarter of the project. The three reactor concept teams have completed initial plant concept development, evaluation and layout. A significant design effort has proceeded with substantial change and evolution from original ideas. The concepts have been reviewed by the industry participants and improvements have been implemented. The third phase, industrial engineering simulation of reactor fabrication has begun

  20. Present status of reactor physics in the United States and Japan-IV. 2. Micro-Reactor Physics of MOX-Fueled Core

    International Nuclear Information System (INIS)

    Takeda, Toshikazu

    2001-01-01

    Recently, fuel assemblies of light water reactors have become complicated because of the extension of fuel burnup and the use of high-enriched Gd and mixed-oxide (MOX) fuel, etc. In conventional assembly calculations, the detailed flux distribution, spectrum distribution, and space dependence of self-shielding within a fuel pellet are not directly taken into account. The experimental and theoretical study of investigating these microscopic properties is named micro-reactor physics. The purpose of this work is to show the importance of micro-reactor physics in the analysis of MOX fuel assemblies. Several authors have done related studies; however, their studies are limited to fuel pin cells, and they are never mentioned with regard to burnup effect, which is important for actual core design. We used the subgroup method to treat the space dependence of the self-shielding effect of heavy nuclides, and we used the characteristics method to treat the angular dependence of neutron flux in a fuel pellet. Figure 1 compares the power distributions in MOX and UO 2 fuel cells at the beginning of burnup. The power is calculated with and without considering the space dependence of the self-shielding effect of the cross sections. For the MOX cell, the power distribution has a peak at the cell edge because of large Pu absorption especially when considering the spatial self-shielding effect. When a MOX rod is adjacent to UO 2 fuel rods, the flux distribution has an azimuthal dependence in addition to the radial dependence within a rod. For example, consider a 2x2 fuel assembly composed of three UO 2 rods and one MOX rod, with the mirror reflection boundary condition. A burnup calculation was done with the condition; the radius of the MOX pellet is divided into two regions, and the azimuthal angle is divided into eight. The number density of 239 Pu at 44 000 MWd/t for the MOX rod shows azimuthal dependence by 20%. The maximum burnup occurs in the direction of the UO 2 rods. This is

  1. 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.)

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

    Science.gov (United States)

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

    2011-06-01

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

  3. Propagation of epithermal neutrons in the empty mesh of the EDF3 reactor. Application of Zeus IV

    International Nuclear Information System (INIS)

    Brisbois, Jacques; Sonnet, Alain; Rafalovitch, Armand

    1969-10-01

    The authors report the verification of the Zeus 4 Monte Carlo code by applying it to an actual example of neutron propagation in a large enclosure, the empty mesh of the EDF3 reactor, supplied through the box holes (for the passage of heat transfer fluid ducts). This enclosure is represented by 105 parallelepipedic meshes in which are placed 4 plane and square sources of same cross section as the holes, and emitting inwards in cos θ. Source intensity is determined by flow rate measurements during tests in the air, interpreted by a Zeus-based calculation and channel power measurements during a 387 MW level operation. The presence of ducts is taken into account by applying a global attenuation coefficient to the sources. Calculation and experiment results agree very well in view of geometry complexity [fr

  4. System and safety studies of accelerator driven systems and generation IV reactors for transmutation of minor actinides. Annual report 2009

    International Nuclear Information System (INIS)

    Bergloef, Calle; Fokau, Andrei; Jolkkonen, Mikael; Tesinsky, Milan; Wallenius, Janne; Youpeng Zhang

    2010-03-01

    During 2009, the reactor physics division has made a design study of a source efficient ADS with nitride fuel and 15/15Ti cladding, based on the EFIT design made within the EUROTRANS project. It was shown that the source efficiency may be doubled as compared to the reference design with oxide fuel and T91 cladding. Transient analysis of a medium sized sodium cooled reactor with MOX fuel allowed to define criteria in terms of power penalty for americium introduction. It was shown that for each percent of americium added to the fuel, the linear rating must be reduced by 6% in order for the fuel to survive postulated unprotected transients. The Sjoestrand area ratio method for reactivity determination has been evaluated experimentally in the strongly heterogeneous subcritical facility YALINA-Booster. Surprisingly, it has been found that the area ratio reactivity estimates may differ by a factor of two depending on detector position. It is shown that this strong spatial dependence can be explained based on a two-region point kinetics model and rectified by means of correction factors obtained through Monte Carlo simulations. For the purpose of measuring high energy neutron cross sections at the SCANDAL facility in Uppsala, Monte Carlo simulations of neutron to proton conversion efficiencies in CsI detectors have been performed. A uranium fuel fabrication laboratory has been taken into operation at KTH in 2009. Uranium and zirconium nitride powders have been fabricated by hydridation/nitridation of metallic source materials. Sample pellets have been pressed and ZrN discs have been sintered to 93% density by means of spark plasma sintering methods

  5. System and safety studies of accelerator driven systems and generation IV reactors for transmutation of minor actinides. Annual report 2009

    Energy Technology Data Exchange (ETDEWEB)

    Bergloef, Calle; Fokau, Andrei; Jolkkonen, Mikael; Tesinsky, Milan; Wallenius, Janne; Youpeng Zhang (Div. of Reactor Physics, Royal Institute of Technology, Stockholm (Sweden))

    2010-03-15

    During 2009, the reactor physics division has made a design study of a source efficient ADS with nitride fuel and 15/15Ti cladding, based on the EFIT design made within the EUROTRANS project. It was shown that the source efficiency may be doubled as compared to the reference design with oxide fuel and T91 cladding. Transient analysis of a medium sized sodium cooled reactor with MOX fuel allowed to define criteria in terms of power penalty for americium introduction. It was shown that for each percent of americium added to the fuel, the linear rating must be reduced by 6% in order for the fuel to survive postulated unprotected transients. The Sjoestrand area ratio method for reactivity determination has been evaluated experimentally in the strongly heterogeneous subcritical facility YALINA-Booster. Surprisingly, it has been found that the area ratio reactivity estimates may differ by a factor of two depending on detector position. It is shown that this strong spatial dependence can be explained based on a two-region point kinetics model and rectified by means of correction factors obtained through Monte Carlo simulations. For the purpose of measuring high energy neutron cross sections at the SCANDAL facility in Uppsala, Monte Carlo simulations of neutron to proton conversion efficiencies in CsI detectors have been performed. A uranium fuel fabrication laboratory has been taken into operation at KTH in 2009. Uranium and zirconium nitride powders have been fabricated by hydridation/nitridation of metallic source materials. Sample pellets have been pressed and ZrN discs have been sintered to 93% density by means of spark plasma sintering methods

  6. Multi-objective technico-economic optimization of energy conversion systems: hydrogen and electricity cogeneration from Generation IV nuclear reactor

    International Nuclear Information System (INIS)

    Gomez, A.

    2008-01-01

    With the increase in environmental considerations, such as the control of greenhouse emissions, and with the decrease in the fossil energy resources, hydrogen is currently considered as a promising energy vector. One of the main technological challenges of a future hydrogen economy is its large scale production without fossil fuel emissions. Under this context, nuclear energy is particularly adapted for hydrogen massive production by thermochemical cycles or high temperature electrolysis. One of the selected nuclear systems is the Very High Temperature Reactor (950 C/1200 C), cooled with helium, and dedicated to hydrogen production or to hydrogen electricity cogeneration. The main objective of this investigation, within the framework of a collaboration between CEA, French Atomic Agency (Cadarache) and LGC (Toulouse), consists in defining a technico-economic optimization methodology of electricity-hydrogen cogeneration systems, in order to identify and propose promising development strategies. Among the massive production processes of hydrogen, the thermochemical cycle Iodine-Sulphur has been considered. Taking into account the diversity of the used energies (i.e., heat and electricity) on the one hand and of the produced energies (hydrogen and electricity) on the other hand of the studied cogeneration system, an exergetic approach has been developed due to its ability to consider various energy forms on the same thermodynamical basis. The CYCLOP software tool (CEA) is used for the thermodynamic modelling of these systems. The economic criterion, calculated using the SEMER software tool (CEA), is based on the minimization of the total production site cost over its lifespan i.e., investment, operating costs and nuclear fuel cost. Capital investment involves the development of cost functions adapted to specific technologies and their specific operating conditions. The resulting optimization problems consist in maximizing the energy production, while minimizing the

  7. Change of I-V characteristics of SiC diodes upon reactor irradiation; Modification des caracteristiques I-V de jonctions p-n au SiC du fait d'une irradiation dans un reacteur; Izmeneniya kharakteristik I-V vyrashchennogo v SiC perekhoda tipa p-n posle oblucheniya ego v reaktore; Modificaciones que sufren por irradiacion en un reactor las caracteristicas I-V de uniones p-n en SiC

    Energy Technology Data Exchange (ETDEWEB)

    Heerschap, M; De Coninck, R [Solid State Physics Dept., SCK-CEN, Mol (Belgium)

    1962-04-15

    In search for semiconductors, which can be used in high-flux reactors in order to measure flux distributions, we irradiated SiC p-n junctions in the Belgium BR-1 reactor. Two types of SiC-diodes of different origin have been irradiated. These junctions are grown in the Lely-furnace. The change in forward and reverse characteristics have been measured during and after irradiation up to temperatures of 150{sup o}C, while measurements up to a temperature of 500{sup o}C are in progress. It has been found that one type resists BR-1 neutrons up to an integrated flux of 10{sup 15} n/cm{sup 2}, while the other resists irradiation up to a flux of 10{sup 17} n/cm{sup 2}. The changes in characteristics are given as well as the result of some annealing experiments. (author) [French] En recherchant des semi-conducteurs pouvant servir a mesurer les distributions de flux dans les reacteurs a haut flux de neutrons, les auteurs ont irradie des jonctions p-n au SiC dans le reacteur belge BR-1. Deux types de diodes a SiC d'origines differentes ont ete ainsi irradies. Les jonctions en question sont preparees par etirage dans le four Lely. Les auteurs ont mesure les modifications subies par les caracteristiques I-V apres et pendant l'irradiation a des temperatures allant jusqu'a 150{sup o}C; ils poursuivent leurs mesures dans la gamme des temperatures allant de 150{sup o}C a 500{sup o}C. Us ont constate que l'un des types de diode a SiC resiste aux neutrons du reacteur BR-1 jusqu'a 10{sup 15} n/cm{sup 2}, tandis que l'autre type resiste a l'irradiation jusqu'a 10{sup 17} n/cm{sup 2}. Les auteurs indiquent les modifications subies par les caracteristiques, ainsi que le resultat de certaines experiences de recuit. (author) [Spanish] Los autores estan tratando de encontrar semiconductores con los que sea posible medir distribuciones de flujo en reactores de flujo elevado, y con este fin irradiaron uniones p-n del SiC en el reactor BR-1 de Belgica. Irradiaron dos tipos de diodos de SiC de

  8. Summary of ORSphere critical and reactor physics measurements

    Directory of Open Access Journals (Sweden)

    Marshall Margaret A.

    2017-01-01

    Full Text Available In the early 1970s Dr. John T. Mihalczo (team leader, J.J. Lynn, and J.R. Taylor performed experiments at the Oak Ridge Critical Experiments Facility (ORCEF with highly enriched uranium (HEU metal (called Oak Ridge Alloy or ORALLOY to recreate GODIVA I results with greater accuracy than those performed at Los Alamos National Laboratory in the 1950s. The purpose of the Oak Ridge ORALLOY Sphere (ORSphere experiments was to estimate the unreflected and unmoderated critical mass of an idealized sphere of uranium metal corrected to a density, purity, and enrichment such that it could be compared with the GODIVA I experiments. This critical configuration has been evaluated. Preliminary results were presented at ND2013. Since then, the evaluation was finalized and judged to be an acceptable benchmark experiment for the International Criticality Safety Benchmark Experiment Project (ICSBEP. Additionally, reactor physics measurements were performed to determine surface button worths, central void worth, delayed neutron fraction, prompt neutron decay constant, fission density and neutron importance. These measurements have been evaluated and found to be acceptable experiments and are discussed in full detail in the International Handbook of Evaluated Reactor Physics Benchmark Experiments. The purpose of this paper is to summarize all the evaluated critical and reactor physics measurements evaluations.

  9. Summary of ORSphere critical and reactor physics measurements

    Science.gov (United States)

    Marshall, Margaret A.; Bess, John D.

    2017-09-01

    In the early 1970s Dr. John T. Mihalczo (team leader), J.J. Lynn, and J.R. Taylor performed experiments at the Oak Ridge Critical Experiments Facility (ORCEF) with highly enriched uranium (HEU) metal (called Oak Ridge Alloy or ORALLOY) to recreate GODIVA I results with greater accuracy than those performed at Los Alamos National Laboratory in the 1950s. The purpose of the Oak Ridge ORALLOY Sphere (ORSphere) experiments was to estimate the unreflected and unmoderated critical mass of an idealized sphere of uranium metal corrected to a density, purity, and enrichment such that it could be compared with the GODIVA I experiments. This critical configuration has been evaluated. Preliminary results were presented at ND2013. Since then, the evaluation was finalized and judged to be an acceptable benchmark experiment for the International Criticality Safety Benchmark Experiment Project (ICSBEP). Additionally, reactor physics measurements were performed to determine surface button worths, central void worth, delayed neutron fraction, prompt neutron decay constant, fission density and neutron importance. These measurements have been evaluated and found to be acceptable experiments and are discussed in full detail in the International Handbook of Evaluated Reactor Physics Benchmark Experiments. The purpose of this paper is to summarize all the evaluated critical and reactor physics measurements evaluations.

  10. On the Safety and Performance Demonstration Tests of Prototype Gen-IV Sodium-Cooled Fast Reactor and Validation and Verification of Computational Codes

    Directory of Open Access Journals (Sweden)

    Jong-Bum Kim

    2016-10-01

    Full Text Available The design of Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR has been developed and the validation and verification (V&V activities to demonstrate the system performance and safety are in progress. In this paper, the current status of test activities is described briefly and significant results are discussed. The large-scale sodium thermal-hydraulic test program, Sodium Test Loop for Safety Simulation and Assessment-1 (STELLA-1, produced satisfactory results, which were used for the computer codes V&V, and the performance test results of the model pump in sodium showed good agreement with those in water. The second phase of the STELLA program with the integral effect tests facility, STELLA-2, is in the detailed design stage of the design process. The sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger performance test, the intermediate heat exchanger test facility, and the test facility for the reactor flow distribution are underway. Flow characteristics test in subchannels of a wire-wrapped rod bundle has been carried out for safety analysis in the core and the dynamic characteristic test of upper internal structure has been performed for the seismic analysis model for the PGSFR. The performance tests for control rod assemblies (CRAs have been conducted for control rod drive mechanism driving parts and drop tests of the CRA under scram condition were performed. Finally, three types of inspection sensors under development for the safe operation of the PGSFR were explained with significant results.

  11. Multi-scale approach to the modeling of fission gas discharge during hypothetical loss-of-flow accident in gen-IV sodium fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Behafarid, F.; Shaver, D. R. [Rensselaer Polytechnic Inst., Troy, NY (United States); Bolotnov, I. A. [North Carolina State Univ., Raleigh, NC (United States); Jansen, K. E. [Univ. of Colorado, Boulder, CO (United States); Antal, S. P.; Podowski, M. Z. [Rensselaer Polytechnic Inst., Troy, NY (United States)

    2012-07-01

    The required technological and safety standards for future Gen IV Reactors can only be achieved if advanced simulation capabilities become available, which combine high performance computing with the necessary level of modeling detail and high accuracy of predictions. The purpose of this paper is to present new results of multi-scale three-dimensional (3D) simulations of the inter-related phenomena, which occur as a result of fuel element heat-up and cladding failure, including the injection of a jet of gaseous fission products into a partially blocked Sodium Fast Reactor (SFR) coolant channel, and gas/molten sodium transport along the coolant channels. The computational approach to the analysis of the overall accident scenario is based on using two different inter-communicating computational multiphase fluid dynamics (CMFD) codes: a CFD code, PHASTA, and a RANS code, NPHASE-CMFD. Using the geometry and time history of cladding failure and the gas injection rate, direct numerical simulations (DNS), combined with the Level Set method, of two-phase turbulent flow have been performed by the PHASTA code. The model allows one to track the evolution of gas/liquid interfaces at a centimeter scale. The simulated phenomena include the formation and breakup of the jet of fission products injected into the liquid sodium coolant. The PHASTA outflow has been averaged over time to obtain mean phasic velocities and volumetric concentrations, as well as the liquid turbulent kinetic energy and turbulence dissipation rate, all of which have served as the input to the core-scale simulations using the NPHASE-CMFD code. A sliding window time averaging has been used to capture mean flow parameters for transient cases. The results presented in the paper include testing and validation of the proposed models, as well the predictions of fission-gas/liquid-sodium transport along a multi-rod fuel assembly of SFR during a partial loss-of-flow accident. (authors)

  12. Development and application of an advanced fuel model for the safety analysis of the generation IV gas-cooled fast reactor

    International Nuclear Information System (INIS)

    Petkevich, P.

    2008-09-01

    Until about the year 2030, current-day nuclear power plants (NPPs) will be replaced by so-called Gen-III or Gen-III+ units, which are mainly based on light water reactor technology. The principal new features are increased safety and improved economical effectiveness. However, these systems use the same fuel forms and are based on the same fuel cycle. Beyond 2030, the interest is likely to shift towards fourth generation NPPs, which offer the possibility of complete fuel cycle closure. Generation-IV reactor concepts include both thermal and fast systems, and involve a wide range of fuel forms and compositions. The present research has been focused on the development of a thermo-mechanical model for the innovative fuel design of the Generation-IV Gas-cooled Fast Reactor (GFR). The principal distinctive feature of the fuel is that the fuel pellets are arranged within plates which enclose an inner honeycomb structure. Apart from the geometry, the usage of new materials is foreseen. Thus, the fuel pellets are of mixed uranium-plutonium carbide, and the cladding is bulk or fiber-reinforced SiC. The setting up of an appropriate materials database was thus the very first task which had to be carried out in the current work. The main purpose of the currently developed model is to provide reliable data, in the context of transient analysis, for the calculation of the principal neutronic feedbacks in the GFR core, viz. the fuel temperature for the Doppler effect and the fuel plate deformation for the axial core expansion effect. None of the available fuel modeling codes is suitable for a realistic simulation of the GFR fuel, as the inner honeycomb structure cannot be explicitly taken into account. The development work has been carried out largely in the context of PSI’s generic code system for fast reactor safety analysis, FAST. Thereby, it has mainly involved extension of the thermo-mechanical code FRED, developed originally for the modeling of traditional rodded fuel

  13. Development of the control assembly pattern and dynamic analysis of the generation IV large gas-cooled fast reactor (GFR)

    International Nuclear Information System (INIS)

    Girardin, G.

    2009-01-01

    Among the systems selected by the GIF, the Gas-cooled Fast Reactor (GFR) is a highly innovative system with advanced fuel geometry and materials. It is in the context of the large, 2400 MWth reference GFR design that the present doctoral research has been conducted, the principal aim having been to develop and qualify the control assembly (CA) pattern and corresponding CA implementation scheme for this system. The work has been carried out in three successive and complementary phases: (1) validation of the neutronics tools, (2) the CA pattern development and related static analysis, and (3) dynamic core behavior studies for hypothetical CA driven transients. During the first phase of the thesis, the reference PROTEUS test lattice from these experiments has been analyzed with ERANOS-2.0 and its associated, adjusted nuclear data library ERALIB1. Additionally, benchmark calculations were performed with the Monte Carlo code MCNPX, allowing one to both check the deterministic results and to analyze the sensitivity to different modern data libraries. It has been found that, for the main reaction rate ratios, the new analysis of the GCFR-PROTEUS reference lattice generally yields good agreement - within 1σ measurement uncertainty - with experimental values and with the Monte Carlo simulations. As shown by the analysis, the predictions were in somewhat better agreement in the case of the adjusted ERALIB1 library. The applicability of ERANOS-2.0/ERALIB1 as the reference neutronics tool for the GFR analysis could thus be demonstrated. Furthermore, neutronics aspects related to the novel features of the GFR, for which new experimental investigations are needed, were highlighted. In the second phase of the research, the CA pattern was developed for the GFR, based on iterative neutronics and thermal-hydraulics calculations, 2D and 3D neutronics models for the reactor core having first been set up using the reference ERANOS-2.0/ERALIB1 computational scheme. For the thermal

  14. Applying the PR and PP Methodology for a qualitative assessment of a misuse scenario in a notional Generation IV Example Sodium Fast Reactor. Assessing design variations

    Energy Technology Data Exchange (ETDEWEB)

    Cojazzi, G.G.M.; Renda, G. [European Commission, Joint Research Centre, Institute for the Protection and Security of the Citizen, TP 210, Via E. Fermi 2749, I-21027, Ispra - Va (Italy); Hassberger, J. [Lawrence Livermore National Laboratory (United States)

    2009-06-15

    The Generation IV International Forum (GIF) Proliferation Resistance and Physical Protection (PR and PP) Working Group has developed a methodology for the PR and PP evaluation of advanced nuclear energy systems. The methodology is organised as a progressive approach applying alternative methods at different levels of thoroughness as more design information becomes available and research improves the depth of technical knowledge. The GIF Proliferation Resistance and Physical Protection (PR and PP) Working Group developed a notional sodium cooled fast neutron nuclear reactor, named the Example Sodium Fast Reactor (ESFR), for use in developing and testing the methodology. The ESFR is a hypothetical nuclear energy system consisting of four sodium-cooled fast reactors of medium size, co-located with an on-site dry fuel storage facility and a Fuel Cycle Facility with pyrochemical processing of the spent fuel and re-fabrication of new ESFR fuel elements. The baseline design is an actinide burner, with LWR spent fuel elements as feed material processed on the site. In the years 2007 and 2008 the GIF PR and PP Working Group performed a case study designed to both test the methodology and demonstrate how it can provide useful feedback to designers even during pre-conceptual design. The Study analysed the response of the entire ESFR system to different proliferation and theft strategies. Three proliferation threats were considered: Concealed diversion, Concealed Misuse and Abrogation. An overt theft threat was also studied. One of the objectives of the case study is to confirm the capability of the methodology to capture PR and PP differences among varied design configurations. To this aim Design Variations (DV) have been also defined corresponding respectively to a) a small variation of the baseline design (DV0), b) a deep burner configuration (DV1), c) a self sufficient core (DV2), and c) a breeder configuration (DV3). This paper builds on the approach followed for the

  15. Actions needed for RA reactor exploitation - I-IV, Part II, Design project VI-SA 1, Experimental loop for testing the EL-4 reactor fuel elements in the central vertical experimental channel of the RA reactor in Vinca

    International Nuclear Information System (INIS)

    Novakovic, M.

    1961-12-01

    The objective of installing the VISA-1 loop was testing the fuel elements of the EL-4 reactor. The fuel elements planned for testing are natural UO 2 with beryllium cladding, cooled by CO 2 under nominal pressure of 60 at and temperature 600 deg C. central vertical experimental channel of the RA reactor was chosen for installing a test loop cooled by CO 2 . This report contains the detailed design project of the testing loop with the control system and safety analysis of the planned experiment

  16. Condensation and homogenization of cross sections for the deterministic transport codes with Monte Carlo method: Application to the GEN IV fast neutron reactors

    International Nuclear Information System (INIS)

    Cai, Li

    2014-01-01

    In the framework of the Generation IV reactors neutronic research, new core calculation tools are implemented in the code system APOLLO3 for the deterministic part. These calculation methods are based on the discretization concept of nuclear energy data (named multi-group and are generally produced by deterministic codes) and should be validated and qualified with respect to some Monte-Carlo reference calculations. This thesis aims to develop an alternative technique of producing multi-group nuclear properties by a Monte-Carlo code (TRIPOLI-4). At first, after having tested the existing homogenization and condensation functionalities with better precision obtained nowadays, some inconsistencies are revealed. Several new multi-group parameters estimators are developed and validated for TRIPOLI-4 code with the aid of itself, since it has the possibility to use the multi-group constants in a core calculation. Secondly, the scattering anisotropy effect which is necessary for handling neutron leakage case is studied. A correction technique concerning the diagonal line of the first order moment of the scattering matrix is proposed. This is named the IGSC technique and is based on the usage of an approximate current which is introduced by Todorova. An improvement of this IGSC technique is then presented for the geometries which hold an important heterogeneity property. This improvement uses a more accurate current quantity which is the projection on the abscissa X. The later current can represent the real situation better but is limited to 1D geometries. Finally, a B1 leakage model is implemented in the TRIPOLI-4 code for generating multi-group cross sections with a fundamental mode based critical spectrum. This leakage model is analyzed and validated rigorously by the comparison with other codes: Serpent and ECCO, as well as an analytical case.The whole development work introduced in TRIPOLI-4 code allows producing multi-group constants which can then be used in the core

  17. 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...

  18. Theoretical analysis of nuclear reactors (Phase III), I-V, Part IV, Influence of isotopic composition of nuclear fuel on the reactivity with constant flux; Razrada metoda teorijske analize nuklearnih reaktora (III faza) I-IV, IV Deo, Uticaj promene izotopnog sastava goriva na reaktivnost uz konstantan fluks

    Energy Technology Data Exchange (ETDEWEB)

    Pop-Jordanov, J [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1963-01-15

    Part one of this report presents a series of differential equations describing the nuclear fuel depletion during reactor operation. This series of differential equations is extended to describe the fission products. This part includes equations for effective multiplication factor k{sub eff} and reactivity {rho} as a function of irradiation {tau}. Part two includes results obtained on the analog computer PACE 231 R, and related to Calder Hall type reactor. Part three covers detailed preparation of the series of equations for solution by using the analog computer. Part four includes the list of references related to this task.

  19. Improvement of computer programs 'BAMBOO' and 'ASFRE-IV' for coupling analysis of deformation and thermal-hydraulics in a high burn-up fuel subassembly of fast reactor

    International Nuclear Information System (INIS)

    Uwaba, Tomoyuki; Ohshima, Hiroyuki; Imai, Yasutomo

    2003-04-01

    A simulation system of a deformed fuel subassembly is being developed for the structure integrity of high burn-up wire-spacer-type fuel subassemblies of sodium-cooled fast breeder reactors. This report describes a computer program improvement work for coupling analyses of deformation and thermal-hydraulics in a fuel subassembly as part of the simulation system development. In this work, a function of data conversion as an interface between a bundle deformation analysis program BAMBOO and a thermal hydraulic analysis program ASFRE-IV was incorporated to each program. BAMBOO was improved to accept the coolant temperature data from ASFRE-IV and to offer bundle deformation data to ASFRE-IV. ASFRE-IV was also improved to offer the coolant temperature data to BAMBOO and to obtain the bundle deformation data from BAMBOO. Improved BAMBOO and ASFRE-IV were applied to an analysis of 169-pin bundle for the program verification. It was confirmed that the coupling analysis gave the physically reasonable results on both deformation and thermal hydraulic behaviors in the fuel subassembly. (author)

  20. 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.)

  1. Activities needed for exploitation of the RA reactor I-IV, Part III, IZ-093-0103-1961; Radovi za potrebe eksploatacije raktora RA - I-IV - III deo, IZ-093-0103-1961

    Energy Technology Data Exchange (ETDEWEB)

    Jurida, V [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1961-12-15

    For the purpose of designing the low-temperature loop at the RA reactor, a list of materials that would be irradiated in the reactor was made. This document includes a review of the radiochemistry data for the listed materials and data about thermal stability of the mentioned materials. Calculation of heat generated in listed samples was based on the theoretical and experimental results of radiation characteristics of the RA reactor. Maximum temperature values in the samples without forced cooling are calculated. [Serbo-Croat] Za potrebe projektovanja niskotemperaturne petlje na reaktoru RA nacinjen je pregled materijala za ozracivanje u reaktoru. Ovaj dokument sadrzi pregled radiajaciono hemijskih podataka za navedene materijale i podatke o termickoj stabilnosti pomenutih materijala. Toplota generisana u navedenim uzorcima racunata je uzimajuci u obzir osnovne podatke teorijske i eksperimentalne rezultate odredjivanja karakteristika zracenja u reaktoru. Izracunate su maksimalne temperature u uzorcima bez prinudnog hladjenja.

  2. Expected value of finite fission chain lengths of pulse reactors

    International Nuclear Information System (INIS)

    Liu Jianjun; Zhou Zhigao; Zhang Ben'ai

    2007-01-01

    The average neutron population necessary for sponsoring a persistent fission chain in a multiplying system, is discussed. In the point reactor model, the probability function θ(n, t 0 , t) of a source neutron at time t 0 leading to n neutrons at time t is dealt with. The non-linear partial differential equation for the probability generating function G(z; t 0 , t) is derived. By solving the equation, we have obtained an approximate analytic solution for a slightly prompt supercritical system. For the pulse reactor Godiva-II, the mean value of finite fission chain lengths is estimated in this work and shows that the estimated value is reasonable for the experimental analysis. (authors)

  3. Theoretical analysis of nuclear reactors (Phase I), I-V, Part V, Determining the fine flux distribution; Razrada metoda teorijske analize nuklearnih reaktora (I faza) I-V, V Deo, Odredjivanje fine raspodele fluksa

    Energy Technology Data Exchange (ETDEWEB)

    Pop-Jordanov, J [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1962-07-15

    Mono energetic neutron transport equation was solved by Carlson numerical method in cylindrical geometry. S{sub n} code was developed for the digital computer ZUSE Z23. Neutron flux distribution was determined for the RA reactor cell by applying S{sub 4} approximation. Reactor cell was treated as D{sub 2}O-U-D{sub 2}O system. Time of iteration was 185 s. Resena je transportna monoenergetska jednacina numerickom metodom Carlsona u cilindricnoj geometriji. Razvijanje S{sub n} kod za digitalnu masinu ZUSE-Z 23. Odredjena je raspodela fluksa u celiji reaktora RA S{sub 4} aproksimacijom. Celija je tretirana kao D{sub 2}O-U-D{sub 2}O. Vreme iteracije je 185 sec (author)

  4. Activities of the Service for maintenance of the RA reactor electronic equipment in 1979 - Report - Annex IV; Prilog IV Rad sluzbe za odrzavanje elektronske opreme reaktora RA u 1979. godini - Izvestaj -

    Energy Technology Data Exchange (ETDEWEB)

    Milosevic, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1979-12-15

    Within the organizational structure of the RA reactor staff, the Service for instrumentation maintenance has the following tasks: maintenance of the existing electronic equipment; participating in experiments planning and preparation of electronic equipment; purchasing new equipment, spare parts and components; construction of new equipment for internal needs; implementation of new equipment. Basic instrumentation of the reactor facility includes: control and protection system, and dosimetry system. [Serbo-Croat] U okviru organizacione seme OOUR nuklearni reaktor RA sluzba za odrzavanje instrumentacije postrojenja ima sledece zadatke: odrzavanje postojece elektronske opreme; ucesce u planiranju eksperimenata i priprema elektronske opreme; nabavka nove opreme, rezervnih delova i komponenti; rad na izradi elektronske opreme i uredjaja za sopstvene potrebe i izrada projekata u domenu reaktorske opreme i ucesce u ugradnji nove opreme. Osnovna instrumentacija postrojenja obuhvata: sistem za upravljanje i zastitu, sistem za tehnolosku kontrolu i sistem za dozimetrijsku kontrolu.

  5. Actions needed for RA reactor exploitation - I-IV, Part I, Thermotechnical experiments related to RA reactor hot start-up; Radovi za potrebe eksploatacije reaktora RA - I-IV, I Deo, Termotehnicki eksperimenti u vezi pustanja rektora RA na snagu

    Energy Technology Data Exchange (ETDEWEB)

    Novakovic, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1961-12-15

    Heavy water coolant loop of the RA reactor includes the reactor, circulation pumps, heat exchangers and pipes. The objective of this task was measuring the thermal parameters of the RA reactor during operation. This report contains the results of the experiment, calculations of thermal regime for the outer and inner tubes, maximum temperature of the fuel element, fluid flow rate in the reactor channels, temperature of the coolant and fuel element cladding.

  6. An Innovative Three-Dimensional Heterogeneous Coarse-Mesh Transport Method for Advanced and Generation IV Reactor Core Analysis and Design

    International Nuclear Information System (INIS)

    Rahnema, Farzad

    2009-01-01

    This project has resulted in a highly efficient method that has been shown to provide accurate solutions to a variety of 2D and 3D reactor problems. The goal of this project was to develop (1) an accurate and efficient three-dimensional whole-core neutronics method with the following features: based solely on transport theory, does not require the use of cross-section homogenization, contains a highly accurate and self-consistent global flux reconstruction procedure, and is applicable to large, heterogeneous reactor models, and to (2) create new numerical benchmark problems for code cross-comparison.

  7. Theoretical analysis of nuclear reactors (Phase II), I-V, Part V, Determining the fine neutron flux distribution by Pn method; Razrada metoda teorijske analize nuklearnih reaktora (II faza) I-V, V Deo, Odredjivanje fine raspodele fluksa Pn metodom

    Energy Technology Data Exchange (ETDEWEB)

    Pop-Jordanov, J [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1962-10-15

    Expression for spherical harmonic moments were applied. They were obtained by spherical harmonics expansion of monoenergetic transport equation. This report presents the procedure for calculating the neutron flux distribution in the nine-zone reactor cell of the RA reactor in Vinca. The procedure was modelled for digital computer ZUSE Z-23 by expansion of the diagram of the automated P{sub 3} code, which is adequate for P{sub n} code with minor changes. The needed subroutines were developed. The most important ones were those for modified first and second order Bessel functions of n-th order. Computer Z-23 was operating only 15 hours during three months, and thus only the subroutines for modified Bessel functions could be tested and the obtained results were excellent. For the mentioned reason the neutron flux distribution will be calculated in the forthcoming period. Koriscen je izraz za svernoharmonicne momente koji se dobija razvijanjem monoenergetske transportne jednacine u sverne harmonike. Dat je postupak za odredjivanje raspodele neutronskog fluksa u devet medijalnoj celiji reaktora RA u Vinci. Gornji postupak je logicki organizovan na digitalnoj masini ZUSE Z-23 razvijanjem tekuceg dijagrama automatskog P{sub 3} koda, koji sa malim izmenama odgovara Pn kodu. Razvijene su potrebne osnovne subrutine, medju kojima su najznacajnije one za modificirane Beselove funkcije prve i druge vrste n - tog reda. Masina ZUSE Z-23 bila je u radu svega 15 zasova u toku tri meseca, te sam stigao da testiram samo subrutine za modificirane Beselove funkcije koje su dale odlicne rezultate. Iz ovoga razloga nije se mogla dobiti trazena raspodela fluksa. To ce biti ucinjeno u narednom periodu posto cemo moci koristiti masinu Z-23 (author)

  8. Gen IV. Technical and economical aspects

    International Nuclear Information System (INIS)

    Kaluzny, Y.; Legee, F.

    2010-01-01

    In this presentation author deals with development of nuclear reactor type of Generation IV. He concluded that: - Nuclear energy is competitive with regards to the other generation sources; Its competitiveness also increases with CO 2 cost. Considering the nuclear cost breakdown of LWR reactors, it turns out that the uranium is currently not in the range of a threshold for FBR deployment; - The global balance of uranium supply and demand and also innovation required to fulfil GEN IV objectives would probably imply the emergence of fast reactor competitiveness after the turn of the mid-century; - We shall need fast reactors in the coming decade.

  9. Theoretical analysis of nuclear reactors (Phase II), I-V, Part V, Determining the fine neutron flux distribution by Pn method

    International Nuclear Information System (INIS)

    Pop-Jordanov, J.

    1962-10-01

    Expression for spherical harmonic moments were applied. They were obtained by spherical harmonics expansion of monoenergetic transport equation. This report presents the procedure for calculating the neutron flux distribution in the nine-zone reactor cell of the RA reactor in Vinca. The procedure was modelled for digital computer ZUSE Z-23 by expansion of the diagram of the automated P 3 code, which is adequate for P n code with minor changes. The needed subroutines were developed. The most important ones were those for modified first and second order Bessel functions of n-th order. Computer Z-23 was operating only 15 hours during three months, and thus only the subroutines for modified Bessel functions could be tested and the obtained results were excellent. For the mentioned reason the neutron flux distribution will be calculated in the forthcoming period [sr

  10. Positive Experience with Small and Medium Sized Reactors in India, Lessons Learned in the Previous Two Decades and Future Plans. Annex IV

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-12-15

    India has limited uranium resources and about one third of the world's thorium resources. With a view to utilizing these resources in the country for electricity generation, a long term three stage programme has been evolved as a strategy. At the second International Conference on Peaceful Uses of Atomic Energy in Geneva in September 1957, H. Bhabha and N.B. Prasad presented a paper on a study of the contribution of atomic energy to a power programme in India. This paper elaborated the three stage programme of natural uranium fuelled pressurized heavy water reactors (PHWRs) in the first stage, fast breeder reactors (FBRs) using the spent fuel of the second stage, and thorium based reactor systems in the third stage. This three stage programme has been accepted over the years and still holds good, even today. The focus in India, from the very beginning of the programme, has been to make nuclear power self-reliant. Towards this objective, development of human resources, research and development for all aspects of the nuclear fuel cycle, and establishment of an infrastructure for the manufacturing of nuclear components within the country were targeted well before launching the nuclear power programme. Some of the initial activities, over a period of time, have been shaped as industrial units of the government, as well as industries in the country. Such an approach has greatly contributed to the self-reliance of the nuclear power programme, in addition to the benefits to the overall industrial infrastructure in the country. The establishment of research reactors, facilities and laboratories to support the nuclear power programme and related fuel cycle activities at the Bhabha Atomic Research Centre, and the Indira Gandhi Centre for Atomic Research for the FBR programme has made a significant contribution to the national capacity in the frontier areas of nuclear science and technology.

  11. Contribution of nonlinear acoustic to the characterization of micro-bubbles clouds in liquid sodium. Application to the generation IV nuclear reactors

    International Nuclear Information System (INIS)

    Cavaro, M.

    2010-11-01

    The SFR system chosen (Sodium Fast Reactor: fast neutron reactors cooled by liquid sodium) by France led to a fourth-generation prototype named ASTRID. The development of this kind of reactors presents several challenges, particularly in terms of improving the safety and monitoring operation. This involves, among other things, characterization of the bubbles presence in liquid sodium. The characterization of the bubbles presence is the subject of this thesis. It involves the determination of void fraction (gas volume fraction) and histogram of the radii of bubbles. The bibliographic work done has shown that linear acoustic techniques for the characterization of bubble clouds are inadequate to achieve this. However promising leads have been identified by studying nonlinear acoustic techniques. This last idea has therefore been explored. An experimental water bench for the generation and optical control of micro-bubbles cloud allowed us to validate finely the reconstruction of histograms of radii through a technique of nonlinear mixing of a high frequency with a low frequency. The potential of the mixing of two high frequencies, more interesting for the industrial point of view has also been demonstrated. Finally, the bases of the transposition of an original technique of nonlinear resonance spectroscopy applied to a bubbles cloud were explored through the introduction of acoustic resonators. The results offer many interesting opportunities, both in terms of industrial applications and for more fundamental understanding of non-linear behavior of a bubble excited by multiple frequencies and of bubbles clouds excited at low frequency. (author)

  12. IVS Organization

    Science.gov (United States)

    2004-01-01

    International VLBI Service (IVS) is an international collaboration of organizations which operate or support Very Long Baseline Interferometry (VLBI) components. The goals are: To provide a service to support geodetic, geophysical and astrometric research and operational activities. To promote research and development activities in all aspects of the geodetic and astrometric VLBI technique. To interact with the community of users of VLBI products and to integrate VLBI into a global Earth observing system.

  13. Current status of NPP generation IV

    International Nuclear Information System (INIS)

    Yohanes Dwi Anggoro; Dharu Dewi; Nurlaila; Arief Tris Yuliyanto

    2013-01-01

    Today development of nuclear technology has reached the stage of research and development of Generation IV nuclear power plants (advanced reactor systems) which is an innovative development from the previous generation of nuclear power plants. There are six types of power generation IV reactors, namely: Very High Temperature Reactor (VHTR), Sodium-cooled Fast Reactor (SFR), Gas-cooled Fast Reactor (GFR), Lead-cooled Fast Reactor (LFR), Molten Salt Reactor (MSR), and Super Critical Water-cooled Reactor (SCWR). The purpose of this study is to know the development of Generation IV nuclear power plants that have been done by the thirteen countries that are members of the Gen IV International Forum (GIF). The method used is review study and refers to various studies related to the current status of research and development of generation IV nuclear power. The result of this study showed that the systems and technology on Generation IV nuclear power plants offer significant advances in sustainability, safety and reliability, economics, and proliferation resistance and physical protection. In addition, based on the research and development experience is estimated that: SFR can be used optimally in 2015, VHTR in 2020, while NPP types GFR, LFR, MSR, and SCWR in 2025. Utilization of NPP generation IV said to be optimal if fulfill the goal of NPP generation IV, such as: capable to generate energy sustainability and promote long-term availability of nuclear fuel, minimize nuclear waste and reduce the long term stewardship burden, has an advantage in the field of safety and reliability compared to the previous generation of NPP and VHTR technology have a good prospects in Indonesia. (author)

  14. Behaviour of palladium(II), platinum(IV), and rhodium(III) in artificial and natural waters: Influence of reactor surface and geochemistry on metal recovery

    Energy Technology Data Exchange (ETDEWEB)

    Cobelo-Garcia, Antonio [School of Earth, Ocean and Environmental Sciences, University of Plymouth, Drake Circus, Plymouth PL4 8AA (United Kingdom)]. E-mail: antonio.cobelo-garcia@plymouth.ac.uk; Turner, Andrew [School of Earth, Ocean and Environmental Sciences, University of Plymouth, Drake Circus, Plymouth PL4 8AA (United Kingdom); Millward, Geoffrey E. [School of Earth, Ocean and Environmental Sciences, University of Plymouth, Drake Circus, Plymouth PL4 8AA (United Kingdom); Couceiro, Fay [School of Earth, Ocean and Environmental Sciences, University of Plymouth, Drake Circus, Plymouth PL4 8AA (United Kingdom)

    2007-03-07

    The recovery of dissolved platinum group elements (PGE: Pd(II), Pt(IV) and Rh(III)) added to Milli-Q[reg] water, artificial freshwater and seawater and filtered natural waters has been studied, as a function of pH and PGE concentration, in containers of varying synthetic composition. The least adsorptive and/or precipitative loss was obtained for borosilicate glass under most of the conditions employed, whereas the greatest loss was obtained for low-density polyethylene. Of the polymeric materials tested, the adsorptive and/or precipitative loss of PGE was lowest for fluorinated ethylene propylene (Teflon[reg]). The loss of Pd(II) in freshwater was significant due to its affinity for surface adsorption and its relatively low solubility. The presence of natural dissolved organic matter increases the recovery of Pd(II) but enhances the loss of Pt(IV). The loss of Rh(III) in seawater was significant and was mainly due to precipitation, whereas Pd(II) recovery was enhanced, compared to freshwater, because of its complexation with chloride. The results have important implications regarding protocols employed for sample preservation and controlled laboratory experiments used in the study of the speciation and biogeochemical behaviour of PGE.

  15. JSFR design progress related to development of safety design criteria for generation IV sodium-cooled fast reactors. (4) Balance of plant

    International Nuclear Information System (INIS)

    Chikazawa, Yoshitaka; Katoh, Atsushi; Nabeshima, Kunihiko; Ohtaka, Masahiko; Uzawa, Masayuki; Ikari, Risako; Iwasaki, Mikinori

    2015-01-01

    In this paper, design study and evaluation related with safety design criteria (SDC) and safety design guideline (SDG) on the balance of plant (BOP) of the demonstration JSFR including fuel handling system, power supply system, component cooling water system, building arrangement are reported. For the fuel handling system, enhancement of storage cooling system has been investigated adding diversified cooling systems. For the power supply, existing emergency power supply system has been reinforced and alternative emergency power supply system is added. For the component cooling system, requirements and relation with safety grade components such investigated. Additionally for the component cooling system, design impact when adding decay heat removal system by sea water has been investigated. For reactor building, over view of evaluation on the external events and design policy for distributed arrangement is reported. Those design study and evaluation provides background information of SDC and SDG. (author)

  16. Simulators IV

    International Nuclear Information System (INIS)

    Fairchild, B.T.

    1987-01-01

    These proceedings contain papers on simulators with artificial intelligence, and the human decision making process; visuals for simulators: human factors, training, and psycho-physical impacts; the role of institutional structure on simulation projects; maintenance trainers for economic value and safety; biomedical simulators for understanding nature, for medical benefits, and the physiological effects of simulators; the mathematical models and numerical techniques that drive today's simulators; and the demography of simulators, with census papers identifying the population of real-time simulator training devices; nuclear reactors

  17. Irradiated uranium reprocessing, Final report I-VI, IV Deo IV - Separation of uranium, plutonium and fission products from the irradiated fuel of the reactor in Vinca; Prerada ozracenog urana. Zavrsni izvestaj - I-VI, IV Deo - Odvajanje urana, plutonijuma i fisionih produkata iz isluzenog goriva reaktora u Vinci

    Energy Technology Data Exchange (ETDEWEB)

    Gal, I [Institute of Nuclear Sciences Boris Kidric, Laboratorija za visoku aktivnost, Vinca, Beograd (Serbia and Montenegro)

    1961-12-15

    This study describes the technology for separation of uranium, plutonium and fission products from the radioactive water solution which is obtained by dissolving the spent uranium fuel from the reactor in Vinca. The procedure should be completed in a hot cell, with the maximum permitted activity of 10 Ci.

  18. 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.

  19. Safety approach and research and development presentation for the selected systems of the International forum Generation IV

    International Nuclear Information System (INIS)

    Fiorini, G.L.

    2003-01-01

    This paper deals with the six projects of the Generation IV forum: Sodium Fast reactor, lead fast reactor, gas fast reactor, very high temperature reactor, supercritical water reactor, molten salt reactor. The technical objectives of the reactor safety and the design/evaluation approach are discussed. (A.L.B.)

  20. Asteroids IV

    Science.gov (United States)

    Michel, Patrick; DeMeo, Francesca E.; Bottke, William F.

    . Asteroids, like planets, are driven by a great variety of both dynamical and physical mechanisms. In fact, images sent back by space missions show a collection of small worlds whose characteristics seem designed to overthrow our preconceived notions. Given their wide range of sizes and surface compositions, it is clear that many formed in very different places and at different times within the solar nebula. These characteristics make them an exciting challenge for researchers who crave complex problems. The return of samples from these bodies may ultimately be needed to provide us with solutions. In the book Asteroids IV, the editors and authors have taken major strides in the long journey toward a much deeper understanding of our fascinating planetary ancestors. This book reviews major advances in 43 chapters that have been written and reviewed by a team of more than 200 international authorities in asteroids. It is aimed to be as comprehensive as possible while also remaining accessible to students and researchers who are interested in learning about these small but nonetheless important worlds. We hope this volume will serve as a leading reference on the topic of asteroids for the decade to come. We are deeply indebted to the many authors and referees for their tremendous efforts in helping us create Asteroids IV. We also thank the members of the Asteroids IV scientific organizing committee for helping us shape the structure and content of the book. The conference associated with the book, "Asteroids Comets Meteors 2014" held June 30-July 4, 2014, in Helsinki, Finland, did an outstanding job of demonstrating how much progress we have made in the field over the last decade. We are extremely grateful to our host Karri Muinonnen and his team. The editors are also grateful to the Asteroids IV production staff, namely Renée Dotson and her colleagues at the Lunar and Planetary Institute, for their efforts, their invaluable assistance, and their enthusiasm; they made life as

  1. Electromagnetic Near Field Measurements of Two Critical Assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Goettee, Jeffrey David

    2015-11-03

    The reactors employed, Godiva IV and WSMR Fast Burst Reactor, are described first. Then the point reactor kinetics model, electromagnetic potential, and the measurement of kinetics quantities are successively discussed. In summary, reactor power produces measurable electric energy. The electric signal mimics power curve for prompt burst operations - features in logarithmic derivatives match. The electric signature should be dependent on the power and not the derivative; therefore, steady-state modes should be measurable.

  2. Actions needed for RA reactor exploitation - I-IV, Part II, Design project VI-SA 1, Experimental loop for testing the EL-4 reactor fuel elements in the central vertical experimental channel of the RA reactor in Vinca; Radovi za potrebe eksploatacije reaktora RA - I-IV, II Deo, Predprojekat VI-SA 1, Petlja za ispitivanje gorivnih elemenata reaktora EL-4 u centralnom vertikalnom eksperimentalnom kanalu reaktora RA u Vinci

    Energy Technology Data Exchange (ETDEWEB)

    Novakovic, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1961-12-15

    The objective of installing the VISA-1 loop was testing the fuel elements of the EL-4 reactor. The fuel elements planned for testing are natural UO{sub 2} with beryllium cladding, cooled by CO{sub 2} under nominal pressure of 60 at and temperature 600 deg C. central vertical experimental channel of the RA reactor was chosen for installing a test loop cooled by CO{sub 2}. This report contains the detailed design project of the testing loop with the control system and safety analysis of the planned experiment.

  3. Environmental, financial and energy feasibility of the electrolysis of the water steam of a generation IV reactor cooling system during the moments of low consumption ef eletrical energy: challenges and perspectives

    International Nuclear Information System (INIS)

    Stefanelli, Eduardo J.; Vargas, Miltom; Garcia, Pedro L.; Seo, Emilia S.M.; Oliveira, Wagner de S.

    2009-01-01

    Our civilization is in an inflection moment. Our current decisions will lead us to drastic changes in the planet climate or to cleaner and more sustainable energy generation models. In a certain moment in the evolution of our society, we have privileged the productivity instead of the reasonable use of the planet's resources. This option has been leading us to a situation in which these resources are seriously jeopardized. New models of energy generation and use should be discussed and adopted in order to reverse this process. The electric-power consumption is not constant through time and it must be generated at the moment it is going to be used. There are moments of great idleness in the electric-power generation system, counterbalanced by high demand moments. This characteristic has induced us to the construction of a model of great generation capacity that remains without use most of the time, producing huge financial and environmental impacts. In this article, we discuss the environmental, financial and energy viability of using the idle capacity of the electric-power system to, through water steam electrolysis, produce hydrogen, which would be reconverted into electric power in peak moments by a fuel cell. In this study, we aim at investigating the viability of associating a SOFC (Solid Oxide Fuel Cell), acting as an electrolysis bow, to a generation IV reactor, in order to produce hydrogen from superheated water steam in the cooling of the reactor, which will be converted into electric power via SOFC (Solid Oxide Fuel Cell) in peak moments. The method used in this investigation was to study the electric charge variation consumed in a day, randomly selected in relation to the hour of the day, to launch a curve into a diagram 'Demand x hour of the day', to establish the peak moments, the minimum moments, and the average consumption, and, based on these data and geometrically, predict the viability of using the energetic potential of the moments in which the

  4. Environmental, financial and energy feasibility of the electrolysis of the water steam of a generation IV reactor cooling system during the moments of low consumption ef eletrical energy: challenges and perspectives

    Energy Technology Data Exchange (ETDEWEB)

    Stefanelli, Eduardo J.; Vargas, Miltom; Garcia, Pedro L.; Seo, Emilia S.M.; Oliveira, Wagner de S. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)], e-mail: ipen@stefanelli.eng.br

    2009-07-01

    Our civilization is in an inflection moment. Our current decisions will lead us to drastic changes in the planet climate or to cleaner and more sustainable energy generation models. In a certain moment in the evolution of our society, we have privileged the productivity instead of the reasonable use of the planet's resources. This option has been leading us to a situation in which these resources are seriously jeopardized. New models of energy generation and use should be discussed and adopted in order to reverse this process. The electric-power consumption is not constant through time and it must be generated at the moment it is going to be used. There are moments of great idleness in the electric-power generation system, counterbalanced by high demand moments. This characteristic has induced us to the construction of a model of great generation capacity that remains without use most of the time, producing huge financial and environmental impacts. In this article, we discuss the environmental, financial and energy viability of using the idle capacity of the electric-power system to, through water steam electrolysis, produce hydrogen, which would be reconverted into electric power in peak moments by a fuel cell. In this study, we aim at investigating the viability of associating a SOFC (Solid Oxide Fuel Cell), acting as an electrolysis bow, to a generation IV reactor, in order to produce hydrogen from superheated water steam in the cooling of the reactor, which will be converted into electric power via SOFC (Solid Oxide Fuel Cell) in peak moments. The method used in this investigation was to study the electric charge variation consumed in a day, randomly selected in relation to the hour of the day, to launch a curve into a diagram 'Demand x hour of the day', to establish the peak moments, the minimum moments, and the average consumption, and, based on these data and geometrically, predict the viability of using the energetic potential of the moments

  5. Mesures de bruit neutronique auprès des réacteurs Caliban, Godiva IV, Planet et Flattop, Détermination des paramètres cinétiques et des incertitudes associées

    OpenAIRE

    Chapelle , Amaury

    2014-01-01

    The evolution of the neutron power is characterized by several key parameters for safety: the effective multiplication factor from a neutron generation to another, the mean lifetime of neutrons, the delayed neutrons fraction, and the total and leakage multiplications. Neutron noise measurements allow the deduction of the kinetic parameters. They are divided into frequency methods (not described in the PhD work), assessments of probability counts or variance-to-mean measurements. They complete...

  6. Implementation and qualification of MCNP 5 through the intercomparison with the benchmark for the calculation of critical systems Godiva and Jezebel

    International Nuclear Information System (INIS)

    Lara, Rafael G.; Maiorino, Jose R.

    2013-01-01

    This work aimed at the implementation and qualification of MCNP code in a supercomputer of the Universidade Federal do ABC, so that may be available a next-generation simulation tool for precise calculations of nuclear reactors and systems subject to radiation. The implementation of this tool will have multidisciplinary applications, covering various areas of engineering (nuclear, aerospace, biomedical), radiation physics and others

  7. Development of generation IV nuclear energy systems

    International Nuclear Information System (INIS)

    Matsui, Kazuaki; Oka, Yoshiaki; Ogawa, Masuro; Ichimiya, Masakazu; Noda, Hiroshi

    2003-01-01

    The fifth 'Generation IV International Forum (GIF), Policy Group Meetings' was held at the Zen-Nikku Hotel in Tokyo, on September 19-20, 2002, under participations of Abraham, Secretary of DOE in U.S.A., Columbani, Secretary of CEA in France, Fujiie, Chairman of CAE in Japan, Kano, Parliamental Minister of MIS in Japan, and so on. Ten nations entering GIF (Argentina, Brazil, Canada, France, Japan, Korea, South Africa, Switzerland, U.K., and U.S.A.) selected six next generation nuclear energy concepts for objects of international cooperative research and development aiming at its practice by 2030. These concepts applicable to not only power generation, but also hydrogen production, sea water purification, and so on, are sodium liquid metal cooled reactor (Japan), high temperature gas cooled reactor (France), Super-critical pressure water cooled reactor (SCWR: Canada), Lead metal cooled reactor (Switzerland), Gas cooled fast reactor (U.S.A.), and molten salts reactor. On the generation IV nuclear reactor systems aiming to further upgrade their sustainability, safety, economical efficiency, and nuclear non proliferation, the 'Plans on Technical Development' (Road-map) to decide priority of their R and Ds has been cooperatively discussed under frameworks of international research cooperation by the GIF members nations. Here were shared descriptions on nuclear fuel cycle as a remise of technical evaluation and adopted concepts by Japanese participants contributing to making up the Road-map. (G.K.)

  8. Scylla IV-P theta pinch

    International Nuclear Information System (INIS)

    Bailey, A.G.; Chandler, G.I.; Ekdahl, C.A. Jr.; Lillberg, J.W.; Machalek, M.D.; Seibel, F.T.

    1976-01-01

    Scylla IV-P is a flexible, linear theta pinch designed to investigate high-density linear concepts, end-stoppering, alternate heating methods, and plasma injection techniques relevant to a pure fusion reactor and/or a fusion-fission hybrid system. The construction and experimental arrangement of the device are briefly described

  9. IV treatment at home

    Science.gov (United States)

    ... Other IV treatments you may receive after you leave the hospital include: Treatment for hormone deficiencies Medicines for severe nausea that cancer chemotherapy or pregnancy may cause Patient-controlled analgesia (PCA) for pain (this is IV ...

  10. Argentinean activities related to Fast Reactors

    International Nuclear Information System (INIS)

    Azpitarte, Osvaldo

    2012-01-01

    CNEA objectives in the area of Generation IV nuclear reactors: Implement a programme for the monitoring of the global progress of new technologies for Generation IV nuclear reactors and their fuel cycles, in order to generate and assess associated lines of R&D. – Perform studies and evaluations for defining the Generation IV line or lines on which CNEA would be interested; – Promote the participation on specific international projects; – Implementation of experimental facilities

  11. New Materials for NGNP/Gen IV

    International Nuclear Information System (INIS)

    Swindeman, Robert W.; Marriott, Douglas L.

    2009-01-01

    The bounding conditions were briefly summarized for the Next Generation Nuclear Plant (NGNP) that is the leading candidate in the Department of Energy Generation IV reactor program. Metallic materials essential to the successful development and proof of concept for the NGNP were identified. The literature bearing on the materials technology for high-temperature gas-cooled reactors was reviewed with emphasis on the needs identified for the NGNP. Several materials were identified for a more thorough study of their databases and behavioral features relative to the requirements ASME Boiler and Pressure Vessel Code, Section III, Division 1, Subsection NH.

  12. Implementation and qualification of MCNP 5 through the intercomparison with the benchmark for the calculation of critical systems Godiva and Jezebel; Implementacao e qualificacao do MCNP5 atraves da intercomparacao com o benchmark para o calculo dos sistemas criticos Godiva e Jezebel

    Energy Technology Data Exchange (ETDEWEB)

    Lara, Rafael G.; Maiorino, Jose R., E-mail: rafael.lara@aluno.ufabc.edu.br, E-mail: joserubens.maiorino@ufabc.edu.br [Universidade Federal do ABC (UFABC), Santo Andre, SP (Brazil). Centro de Engenharia, Modelagem e Ciencias Sociais Aplicadas

    2013-07-01

    This work aimed at the implementation and qualification of MCNP code in a supercomputer of the Universidade Federal do ABC, so that may be available a next-generation simulation tool for precise calculations of nuclear reactors and systems subject to radiation. The implementation of this tool will have multidisciplinary applications, covering various areas of engineering (nuclear, aerospace, biomedical), radiation physics and others.

  13. Generation IV national program

    International Nuclear Information System (INIS)

    Preville, M.; Sadhankar, R.; Brady, D.

    2007-01-01

    This paper outlines the Generation IV National Program. This program involves evolutionary and innovative design with significantly higher efficiencies (∼50% compared to present ∼30%) - sustainable, economical, safe, reliable and proliferation resistant - for future energy security. The Generation IV Forum (GIF) effectively leverages the resources of the participants to meet these goals. Ten countries signed the GIF Charter in 2001

  14. Gen IV Materials Handbook Implementation Plan

    International Nuclear Information System (INIS)

    Rittenhouse, P.; Ren, W.

    2005-01-01

    A Gen IV Materials Handbook is being developed to provide an authoritative single source of highly qualified structural materials information and materials properties data for use in design and analyses of all Generation IV Reactor Systems. The Handbook will be responsive to the needs expressed by all of the principal government, national laboratory, and private company stakeholders of Gen IV Reactor Systems. The Gen IV Materials Handbook Implementation Plan provided here addresses the purpose, rationale, attributes, and benefits of the Handbook and will detail its content, format, quality assurance, applicability, and access. Structural materials, both metallic and ceramic, for all Gen IV reactor types currently supported by the Department of Energy (DOE) will be included in the Gen IV Materials Handbook. However, initial emphasis will be on materials for the Very High Temperature Reactor (VHTR). Descriptive information (e.g., chemical composition and applicable technical specifications and codes) will be provided for each material along with an extensive presentation of mechanical and physical property data including consideration of temperature, irradiation, environment, etc. effects on properties. Access to the Gen IV Materials Handbook will be internet-based with appropriate levels of control. Information and data in the Handbook will be configured to allow search by material classes, specific materials, specific information or property class, specific property, data parameters, and individual data points identified with materials parameters, test conditions, and data source. Details on all of these as well as proposed applicability and consideration of data quality classes are provided in the Implementation Plan. Website development for the Handbook is divided into six phases including (1) detailed product analysis and specification, (2) simulation and design, (3) implementation and testing, (4) product release, (5) project/product evaluation, and (6) product

  15. Generation IV nuclear plant design strategies

    International Nuclear Information System (INIS)

    Altin, V.

    2007-01-01

    In this presentation Generation IV nuclear reactor design criteria are examined under the light of known nuclear properties of fissile and fertile nuclei. Their conflicting nature is elucidated along with the resulting inevitability of a multitude of designs. The designs selected as candidates for further development are evaluated with respect to their potential to serve the different design criteria, thereby revealing their more difficult aspects of realization and the strong research challenges lying ahead

  16. Iris reactor conceptual design

    International Nuclear Information System (INIS)

    Carelli, M.D.; Conway, L.E.; Petrovic, B.; Paramonov, D.V.; Galvin, M.; Todreas, N.E.; Lombardi, C.V.; Maldari, F.; Ricotti, M.E.; Cinotti, L.

    2001-01-01

    IRIS (International Reactor Innovative and Secure) is a modular, integral, light water cooled, low-to-medium power (100-350 MWe) reactor which addresses the requirements defined by the US DOE for Generation IV reactors, i.e., proliferation resistance, enhanced safety, improved economics and fuel cycle sustainability. It relies on the proven technology of light water reactors and features innovative engineering, but it does not require new technology development. This paper discusses the current reference IRIS design, which features a 1000 MWt thermal core with proven 5%-enriched uranium oxide fuel and five-year long straight burn fuel cycle, integral reactor vessel housing helical tube steam generators and immersed spool pumps. Other major contributors to the high level of safety and economic attractiveness are the safety by design and optimized maintenance approaches, which allow elimination of some classes of accidents, lower capital cost, long operating cycle, and high capacity factors. (author)

  17. The nuclear reactor systems

    International Nuclear Information System (INIS)

    Bacher, P.

    2008-01-01

    This paper describes the various nuclear reactor systems, starting with the Generation II, then the present development of the Generation III and the stakes and challenges of the future Generation IV. Some have found appropriate to oppose reactor systems or generations one to another, especially by minimizing the enhancements of generation III compared to generation II or by expecting the earth from generation IV (meaning that generation III is already obsolete). In the first part of the document (chapter 2), some keys are given to the reader to develop its proper opinion. Chapter 3 describes more precisely the various reactor systems and generations. Chapter 4 discusses the large industrial manoeuvres around the generation III, and the last chapter gives some economical references, taking into account, for the various means of power generation, the impediments linked to climate protection

  18. Neptunium (IV) oxalate solubility

    International Nuclear Information System (INIS)

    Luerkens, D.W.

    1983-07-01

    The equilibrium solubility of neptunium (IV) oxalate in nitric/oxalic acid solutions was determined at 22 0 C, 45 0 C, and 60 0 C. The concentrations of nitric/oxalic acid solutions represented a wide range of free oxalate ion concentration. A mathematical solubility model was developed which is based on the formation of the known complexes of neptunium (IV) oxalate. the solubility model uses a simplified concentration parameter which is proportional to the free oxalate ion concentration. The solubility model can be used to estimate the equilibrium solubility of neptunium (IV) oxalate over a wide range of oxalic and nitric acid concentrations at each temperature

  19. NNDSS - Table IV. Tuberculosis

    Data.gov (United States)

    U.S. Department of Health & Human Services — NNDSS - Table IV. Tuberculosis - 2016.This Table includes total number of cases reported in the United States, by region and by states, in accordance with the...

  20. NNDSS - Table IV. Tuberculosis

    Data.gov (United States)

    U.S. Department of Health & Human Services — NNDSS - Table IV. Tuberculosis - 2014.This Table includes total number of cases reported in the United States, by region and by states, in accordance with the...

  1. NNDSS - Table IV. Tuberculosis

    Data.gov (United States)

    U.S. Department of Health & Human Services — NNDSS - Table IV. Tuberculosis - 2015.This Table includes total number of cases reported in the United States, by region and by states, in accordance with the...

  2. SAGE IV Pathfinder

    Data.gov (United States)

    National Aeronautics and Space Administration — Utilizing a unique, new occultation technique involving imaging, the SAGE IV concept will meet or exceed the quality of previous SAGE measurements at a small...

  3. Super critical water reactors

    International Nuclear Information System (INIS)

    Dumaz, P.; Antoni, O; Arnoux, P.; Bergeron, A; Renault, C.; Rimpault, G.

    2005-01-01

    Water is used as a calori-porter and moderator in the most major nuclear centers which are actually in function. In the pressurized water reactor (PWR) and boiling water reactor (BWR), water is maintained under critical point of water (21 bar, 374 Centigrade) which limits the efficiency of thermodynamic cycle of energy conversion (yield gain of about 33%) Crossing the critical point, one can then use s upercritical water , the obtained pressure and temperature allow a significant yield gains. In addition, the supercritical water offers important properties. Particularly there is no more possible coexistence between vapor and liquid. Therefore, we don't have more boiling problem, one of the phenomena which limits the specific power of PWR and BWR. Since 1950s, the reactor of supercritical water was the subject of studies more or less detailed but neglected. From the early 1990s, this type of conception benefits of some additional interests. Therefore, in the international term G eneration IV , the supercritical water reactors had been considered as one of the big options for study as Generation IV reactors. In the CEA, an active city has engaged from 1930 with the participation to a European program: The HPWR (High Performance Light Water Reactor). In this contest, the R and D studies are focused on the fields of neutrons, thermodynamic and materials. The CEA intends to pursue a limited effort of R and D in this field, in the framework of international cooperation, preferring the study of versions of rapid spectrum. (author)

  4. Development of an ultrahigh-temperature process for the enzymatic hydrolysis of lactose. IV. Immobilization of two thermostable beta-glycosidases and optimization of a packed-bed reactor for lactose conversion.

    Science.gov (United States)

    Petzelbauer, Inge; Kuhn, Bernhard; Splechtna, Barbara; Kulbe, Klaus D; Nidetzky, Bernd

    2002-03-20

    Recombinant hyperthermostable beta-glycosidases from the archaea Sulfolobus solfataricus (Ss beta Gly) and Pyrococcus furiosus (CelB) were covalently attached onto the insoluble carriers chitosan, controlled pore glass (CPG), and Eupergit C. For each enzyme/carrier pair, the protein-binding capacity, the immobilization yield, the pH profiles for activity and stability, the activity/temperature profile, and the kinetic constants for lactose hydrolysis at 70 degrees C were determined. Eupergit C was best among the carriers in regard to retention of native-like activity and stability of Ss beta Gly and CelB over the pH range 3.0-7.5. Its protein binding capacity of approximately 0.003 (on a mass basis) was one-third times that of CPG, while immobilization yields were typically 80% in each case. Activation energies for lactose conversion by the immobilized enzymes at pH 5.5 were in the range 50-60 kJ/mol. This is compared to values of approximately 75 kJ/mol for the free enzymes. Immobilization expands the useful pH range for CelB and Ss beta Gly by approximately 1.5 pH units toward pH 3.5 and pH 4.5, respectively. A packed-bed enzyme reactor was developed for the continuous conversion of lactose in different media, including whey and milk, and operated over extended reaction times of up to 14 days. The productivities of the Eupergit C-immobilized enzyme reactor were determined at dilution rates between 1 and 12 h(-1), and using 45 and 170 g/L initial lactose. Results of kinetic modeling for the same reactor, assuming plug flow and steady state, suggest the presence of mass-transfer limitation of the reaction rate under the conditions used. Formation of galacto-oligosaccharides in the continuous packed-bed reactor and in the batch reactor using free enzyme was closely similar in regard to yield and individual saccharide components produced. Copyright 2002 John Wiley & Sons, Inc. Biotechnol Bioeng 77: 619-631, 2002; DOI 10.1002/bit.10110

  5. Fourth Generation Reactor Concepts

    International Nuclear Information System (INIS)

    Furtek, A.

    2008-01-01

    Concerns over energy resources availability, climate changes and energy supply security suggest an important role for nuclear energy in future energy supplies. So far nuclear energy evolved through three generations and is still evolving into new generation that is now being extensively studied. Nuclear Power Plants are producing 16% of the world's electricity. Today the world is moving towards hydrogen economy. Nuclear technologies can provide energy to dissociate water into oxygen and hydrogen and to production of synthetic fuel from coal gasification. The introduction of breeder reactors would turn nuclear energy from depletable energy supply into an unlimited supply. From the early beginnings of nuclear energy in the 1940s to the present, three generations of nuclear power reactors have been developed: First generation reactors: introduced during the period 1950-1970. Second generation: includes commercial power reactors built during 1970-1990 (PWR, BWR, Candu, Russian RBMK and VVER). Third generation: started being deployed in the 1990s and is composed of Advanced LWR (ALWR), Advanced BWR (ABWR) and Passive AP600 to be deployed in 2010-2030. Future advances of the nuclear technology designs can broaden opportunities for use of nuclear energy. The fourth generation reactors are expected to be deployed by 2030 in time to replace ageing reactors built in the 1970s and 1980s. The new reactors are to be designed with a view of the following objectives: economic competitiveness, enhanced safety, minimal radioactive waste production, proliferation resistance. The Generation IV International Forum (GIF) was established in January 2000 to investigate innovative nuclear energy system concepts. GIF members include Argentina, Brazil, Canada, Euratom, France Japan, South Africa, South Korea, Switzerland, United Kingdom and United States with the IAEA and OECD's NEA as permanent observers. China and Russia are expected to join the GIF initiative. The following six systems

  6. Regenerative Heater Optimization for Steam Turbo-Generation Cycles of Generation IV Nuclear Power Plants with a Comparison of Two Concepts for the Westinghouse International Reactor Innovative and Secure (IRIS)

    International Nuclear Information System (INIS)

    Williams, W.C.

    2002-01-01

    The intent of this study is to discuss some of the many factors involved in the development of the design and layout of a steam turbo-generation unit as part of a modular Generation IV nuclear power plant. Of the many factors involved in the design and layout, this research will cover feed water system layout and optimization issues. The research is arranged in hopes that it can be generalized to any Generation IV system which uses a steam powered turbo-generation unit. The research is done using the ORCENT-II heat balance codes and the Salisbury methodology to be reviewed herein. The Salisbury methodology is used on an original cycle design by Famiani for the Westinghouse IRIS and the effects due to parameter variation are studied. The vital parameters of the Salisbury methodology are the incremental heater surface capital cost (S) in $/ft 2 , the value of incremental power (I) in $/kW, and the overall heat transfer coefficient (U) in Btu/ft 2 -degrees Fahrenheit-hr. Each is varied in order to determine the effects on the cycles overall heat rate, output, as well as, the heater surface areas. The effects of each are shown. Then the methodology is then used to compare the optimized original Famiani design consisting of seven regenerative feedwater heaters with an optimized new cycle concept, INRC8, containing four regenerative heaters. The results are shown. It can be seen that a trade between the complexity of the seven stage regenerative Famiani cycle and the simplicity of the INRC8 cycle can be made. It is desired that this methodology can be used to show the ability to evaluate modularity through the value of size a complexity of the system as well as the performance. It also shows the effectiveness of the Salisbury methodology in the optimization of regenerative cycles for such an evaluation

  7. Nuclear reactors

    International Nuclear Information System (INIS)

    Barre, Bertrand

    2015-10-01

    After some remarks on the nuclear fuel, on the chain reaction control, on fuel loading and unloading, this article proposes descriptions of the design, principles and operations of different types of nuclear reactors as well as comments on their presence and use in different countries: pressurized water reactors (design of the primary and secondary circuits, volume and chemistry control, backup injection circuits), boiling water reactors, heavy water reactors, graphite and boiling water reactors, graphite-gas reactors, fast breeder reactors, and fourth generation reactors (definition, fast breeding). For these last ones, six concepts are presented: sodium-cooled fast reactor, lead-cooled fast reactor, gas-cooled fast reactor, high temperature gas-cooled reactor, supercritical water-cooled reactor, and molten salt reactor

  8. IV access in dental practice.

    LENUS (Irish Health Repository)

    Fitzpatrick, J J

    2009-04-01

    Intravenous (IV) access is a valuable skill for dental practitioners in emergency situations and in IV sedation. However, many people feel some apprehension about performing this procedure. This article explains the basic principles behind IV access, and the relevant anatomy and physiology, as well as giving a step-by-step guide to placing an IV cannula.

  9. Safety approach and research and development presentation for the selected systems of the International forum Generation IV; Elements de l'approche de surete et themes de R et D specifiques pour les systemes selectionnes par le Forum International Generation 4

    Energy Technology Data Exchange (ETDEWEB)

    Fiorini, G.L. [CEA Saclay, Dir. du Developpement et de l' Innovation Nucleares (DEN/DDIN), 91 - Gif Sur Yvette (France)

    2003-07-01

    This paper deals with the six projects of the Generation IV forum: Sodium Fast reactor, lead fast reactor, gas fast reactor, very high temperature reactor, supercritical water reactor, molten salt reactor. The technical objectives of the reactor safety and the design/evaluation approach are discussed. (A.L.B.)

  10. Internet Economics IV

    Science.gov (United States)

    2004-08-01

    edts.): Internet Economics IV Technical Report No. 2004-04, August 2004 Information Systems Laboratory IIS, Departement of Computer Science University of...level agreements (SLA), Information technology (IT), Internet address, Internet service provider 16. PRICE CODE 17. SECURITY CLASSIFICATION 18... technology and its economic impacts in the Internet world today. The second talk addresses the area of AAA protocol, summarizing authentication

  11. Uranium (IV) carboxylates - I

    Energy Technology Data Exchange (ETDEWEB)

    Satpathy, K C; Patnaik, A K [Sambalpur Univ. (India). Dept. of Chemistry

    1975-11-01

    A few uranium(IV) carboxylates with monochloro and trichloro acetic acid, glycine, malic, citric, adipic, o-toluic, anthranilic and salicylic acids have been prepared by photolytic methods. The I.R. spectra of these compounds are recorded and basing on the spectral data, structure of the compounds have been suggested.

  12. PLATO IV Accountancy Index.

    Science.gov (United States)

    Pondy, Dorothy, Comp.

    The catalog was compiled to assist instructors in planning community college and university curricula using the 48 computer-assisted accountancy lessons available on PLATO IV (Programmed Logic for Automatic Teaching Operation) for first semester accounting courses. It contains information on lesson access, lists of acceptable abbreviations for…

  13. Advanced Computational Thermal Fluid Physics (CTFP) and Its Assessment for Light Water Reactors and Supercritical Reactors

    International Nuclear Information System (INIS)

    D.M. McEligot; K. G. Condie; G. E. McCreery; H. M. McIlroy; R. J. Pink; L.E. Hochreiter; J.D. Jackson; R.H. Pletcher; B.L. Smith; P. Vukoslavcevic; J.M. Wallace; J.Y. Yoo; J.S. Lee; S.T. Ro; S.O. Park

    2005-01-01

    Background: The ultimate goal of the study is the improvement of predictive methods for safety analyses and design of Generation IV reactor systems such as supercritical water reactors (SCWR) for higher efficiency, improved performance and operation, design simplification, enhanced safety and reduced waste and cost. The objective of this Korean/US/laboratory/university collaboration of coupled fundamental computational and experimental studies is to develop the supporting knowledge needed for improved predictive techniques for use in the technology development of Generation IV reactor concepts and their passive safety systems. The present study emphasizes SCWR concepts in the Generation IV program

  14. Advanced Computational Thermal Fluid Physics (CTFP) and Its Assessment for Light Water Reactors and Supercritical Reactors

    Energy Technology Data Exchange (ETDEWEB)

    D.M. McEligot; K. G. Condie; G. E. McCreery; H. M. McIlroy; R. J. Pink; L.E. Hochreiter; J.D. Jackson; R.H. Pletcher; B.L. Smith; P. Vukoslavcevic; J.M. Wallace; J.Y. Yoo; J.S. Lee; S.T. Ro; S.O. Park

    2005-10-01

    Background: The ultimate goal of the study is the improvement of predictive methods for safety analyses and design of Generation IV reactor systems such as supercritical water reactors (SCWR) for higher efficiency, improved performance and operation, design simplification, enhanced safety and reduced waste and cost. The objective of this Korean / US / laboratory / university collaboration of coupled fundamental computational and experimental studies is to develop the supporting knowledge needed for improved predictive techniques for use in the technology development of Generation IV reactor concepts and their passive safety systems. The present study emphasizes SCWR concepts in the Generation IV program.

  15. Enhanced Design Alternative IV

    International Nuclear Information System (INIS)

    Kramer, N.E.

    1999-01-01

    This report evaluates Enhanced Design Alternative (EDA) IV as part of the second phase of the License Application Design Selection (LADS) effort. The EDA IV concept was compared to the VA reference design using criteria from the Design Input Request for LADS Phase II EDA Evaluations (CRWMS M and O 1999b) and (CRWMS M and O 1999f). Briefly, the EDA IV concept arranges the waste packages close together in an emplacement configuration known as line load. Continuous pre-closure ventilation keeps the waste packages from exceeding their 350 C cladding and 200 C (4.3.6) drift wall temperature limits. This EDA concept keeps relatively high, uniform emplacement drift temperatures (post-closure) to drive water away from the repository and thus dry out the pillars between emplacement drifts. The waste package is shielded to permit human access to emplacement drifts and includes an integral filler inside the package to reduce the amount of water that can contact the waste form. Closure of the repository is desired 50 years after first waste is emplaced. Both backfill and drip shields will be emplaced at closure to improve post-closure performance. The EDA IV concept includes more defense-in-depth layers than the VA reference design because of its backfill, drip shield, waste package shielding, and integral filler features. These features contribute to the low dose-rate to the public achieved during the first 10,000 years of repository life as shown in Figure 3. Investigation of the EDA IV concept has led to the following general conclusions: (1) The total life cycle cost for EDA IV is about $21.7 billion which equates to a $11.3 billion net present value (both figures rounded up). (2) The incidence of design basis events for EDA IV is similar to the VA reference design. (3) The emplacement of the waste packages in drifts will be similar to the VA reference design. However, heavier equipment may be required because the shielded waste package will be heavier. (4) The heavier

  16. Improving fuel cycle design and safety characteristics of a gas cooled fast reactor

    NARCIS (Netherlands)

    van Rooijen, W.F.G.

    2006-01-01

    This research concerns the fuel cycle and safety aspects of a Gas Cooled Fast Reactor, one of the so-called "Generation IV" nuclear reactor designs. The Generation IV Gas Cooled Fast Reactor uses helium as coolant at high temperature. The goal of the GCFR is to obtain a "closed nuclear fuel cycle",

  17. Development of technologies for nuclear reactors of small and medium sized; Desarrollo de Tecnologias para Reactores Nucleares de pequeno y medio tamano

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2011-08-15

    This meeting include: countries presentations, themes and objectives of the training course, reactor types, design, EPR, APR1400, A P 1000, A PWR, ATMEA 1, VVER-1000, A PWR, ATMEA 1, VVER 1200, Boiling Water Reactor, A BWR, A BWR -II, ESBUR, Ke ren, AREVA, Heavy Water Reactor, Candu 6, Acr-1000, HWR, Bw, Iris, CAREM NuCcale, Smart, KLT-HOS, Westinghouse small modular Reactor, Gas Cooled Reactors, PBMR, React ores enfriados con metales liquidos, Hs, Prism,Terra Power, Hyper ion, appliance's no electric as de energia, Generation IV Reactors,VHTR, Gas Fast Reactor, Sodium Fast Reactor, Molten salt Reactor, Lfr, Water Cooled Reactor, Technology Assessment Process, Fukushima accident.

  18. H Reactor

    Data.gov (United States)

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

  19. History of critical experiments at Pajarito Site

    Energy Technology Data Exchange (ETDEWEB)

    Paxton, H.C.

    1983-03-01

    This account describes critical and subcritical assemblies operated remotely at the Pajarito Canyon Site at the Los Alamos National Laboratory. Earliest assemblies, directed exclusively toward the nuclear weapons program, were for safety tests. Other weapon-related assemblies provided neutronic information to check detailed weapon calculations. Topsy, the first of these critical assemblies, was followed by Lady Godiva, Jezebel, Flattop, and ultimately Big Ten. As reactor programs came to Los Alamos, design studies and mockups were tested at Pajarito Site. For example, nearly all 16 Rover reactors intended for Nevada tests were preceded by zero-power mockups and proof tests at Pajarito Site. Expanded interest and capability led to fast-pulse assemblies, culminating in Godiva IV and Skua, and to the Kinglet and Sheba solution assemblies.

  20. History of critical experiments at Pajarito Site

    International Nuclear Information System (INIS)

    Paxton, H.C.

    1983-03-01

    This account describes critical and subcritical assemblies operated remotely at the Pajarito Canyon Site at the Los Alamos National Laboratory. Earliest assemblies, directed exclusively toward the nuclear weapons program, were for safety tests. Other weapon-related assemblies provided neutronic information to check detailed weapon calculations. Topsy, the first of these critical assemblies, was followed by Lady Godiva, Jezebel, Flattop, and ultimately Big Ten. As reactor programs came to Los Alamos, design studies and mockups were tested at Pajarito Site. For example, nearly all 16 Rover reactors intended for Nevada tests were preceded by zero-power mockups and proof tests at Pajarito Site. Expanded interest and capability led to fast-pulse assemblies, culminating in Godiva IV and Skua, and to the Kinglet and Sheba solution assemblies

  1. A sputnik IV saga

    Science.gov (United States)

    Lundquist, Charles A.

    2009-12-01

    The Sputnik IV launch occurred on May 15, 1960. On May 19, an attempt to deorbit a 'space cabin' failed and the cabin went into a higher orbit. The orbit of the cabin was monitored and Moonwatch volunteer satellite tracking teams were alerted to watch for the vehicle demise. On September 5, 1962, several team members from Milwaukee, Wisconsin made observations starting at 4:49 a.m. of a fireball following the predicted orbit of Sputnik IV. Requests went out to report any objects found under the fireball path. An early morning police patrol in Manitowoc had noticed a metal object on a street and had moved it to the curb. Later the officers recovered the object and had it dropped off at the Milwaukee Journal. The Moonwarch team got the object and reported the situation to Moonwatch Headquarters at the Smithsonian Astrophysical Observatory. A team member flew to Cambridge with the object. It was a solid, 9.49 kg piece of steel with a slag-like layer attached to it. Subsequent analyses showed that it contained radioactive nuclei produced by cosmic ray exposure in space. The scientists at the Observatory quickly recognized that measurements of its induced radioactivity could serve as a calibration for similar measurements of recently fallen nickel-iron meteorites. Concurrently, the Observatory directorate informed government agencies that a fragment from Sputnik IV had been recovered. Coincidently, a debate in the UN Committee on Peaceful Uses of Outer Space involved the issue of liability for damage caused by falling satellite fragments. On September 12, the Observatory delivered the bulk of the fragment to the US Delegation to the UN. Two days later, the fragment was used by US Ambassador Francis Plimpton as an exhibit that the time had come to agree on liability for damage from satellite debris. He offered the Sputnik IV fragment to USSR Ambassador P.D. Morozov, who refused the offer. On October 23, Drs. Alla Massevitch and E.K. Federov of the USSR visited the

  2. Corrosion of structural materials for Generation IV systems

    International Nuclear Information System (INIS)

    Balbaud-Celerier, F.; Cabet, C.; Courouau, J.L.; Martinelli, L.; Arnoux, P.

    2009-01-01

    The Generation IV International Forum aims at developing future generation nuclear energy systems. Six systems have been selected for further consideration: sodium-cooled fast reactor (SFR), gas-cooled fast reactor (GFR), lead-cooled fast reactor (LFR), molten salt reactor (MSR), supercritical water-cooled reactor (SCWR) and very high temperature reactor (VHTR). CEA, in the frame of a national program, of EC projects and of the GIF, contributes to the structural materials developments and research programs. Particularly, corrosion studies are being performed in the complex environments of the GEN IV systems. As a matter of fact, structural materials encounter very severe conditions regarding corrosion concerns: high temperatures and possibly aggressive chemical environments. Therefore, the multiple environments considered require also a large diversity of materials. On the other hand, the similar levels of working temperatures as well as neutron spectrum imply also similar families of materials for the various systems. In this paper, status of the research performed in CEA on the corrosion behavior of the structural material in the different environments is presented. The materials studied are either metallic materials as austenitic (or Y, La, Ce doped) and ferrito-martensitic steels, Ni base alloys, ODS steels, or ceramics and composites. In all the environments studied, the scientific approach is identical, the objective being in all cases the understanding of the corrosion processes to establish recommendations on the chemistry control of the coolant and to predict the long term behavior of the materials by the development of corrosion models. (author)

  3. Thermo-hydraulic simulations of the experimental fast reactor core

    International Nuclear Information System (INIS)

    Silveira Luz, M. da; Braz Filho, F.A.; Borges, E.M.

    1985-01-01

    A study of the core and performance of metallic fuel of the experimental fast reactor, from the thermal-hydraulic point of view, was carried out employing the COBRA IV-I code. The good safety characteristics of this reactor and the feasibility of using metallic fuel in experimental fast reactor were demonstrated. (Author) [pt

  4. Development of technologies for nuclear reactors of small and medium sized

    International Nuclear Information System (INIS)

    2011-08-01

    This meeting include: countries presentations, themes and objectives of the training course, reactor types, design, EPR, APR1400, A P 1000, A PWR, ATMEA 1, VVER-1000, A PWR, ATMEA 1, VVER 1200, Boiling Water Reactor, A BWR, A BWR -II, ESBUR, Ke ren, AREVA, Heavy Water Reactor, Candu 6, Acr-1000, HWR, Bw, Iris, CAREM NuCcale, Smart, KLT-HOS, Westinghouse small modular Reactor, Gas Cooled Reactors, PBMR, React ores enfriados con metales liquidos, Hs, Prism,Terra Power, Hyper ion, appliance's no electric as de energia, Generation IV Reactors,VHTR, Gas Fast Reactor, Sodium Fast Reactor, Molten salt Reactor, Lfr, Water Cooled Reactor, Technology Assessment Process, Fukushima accident.

  5. Work related to increasing the exploitation and experimental possibilities of the RA reactor, 05. Independent CO2 loop for cooling the samples irradiated in the RA vertical experimental channels (I-IV), Part II, IZ-240-0379-1963, Vol. II Head of the low temperature RA reactor coolant loop

    International Nuclear Information System (INIS)

    Pavicevic, M.

    1963-07-01

    The objective of the project was to design the head of the CO 2 coolant loop for cooling the materials during irradiation in the RA reactor. Six heads of coolant loops will be placed in the RA reactor, two in the region of heavy water in the experimental channels VEK-6 and four in the graphite reflector in the channels VEK-G. maximum generated heat in the heads of the coolant loop is 10500 kcal/h and minimum generated heat is 1500 kcal/h. The loops are cooled by CO 2 gas, coolant flow is 420 kg/h, and the pressure is 4.5 atu. There is a need to design and construct the secondary coolant loop for the low temperature coolant loop. This volume includes technical specifications of the secondary CO 2 loop with instructions for construction and testing; needed calculations; specification of materials; cost estimation for materials, equipment and construction; and graphical documentation [sr

  6. Thermochemical investigation of molten fluoride salts for Generation IV nuclear applications - an equilibrium exercise

    NARCIS (Netherlands)

    van der Meer, J.P.M.

    2006-01-01

    The concept of the Molten Salt Reactor, one of the so-called Generation IV future reactors, is that the fuel, a fissile material, which is dissolved in a molten fluoride salt, circulates through a closed circuit. The heat of fission is transferred to a second molten salt coolant loop, the heat of

  7. Neutron lifetime and generation time by KENO IV

    International Nuclear Information System (INIS)

    Hayashi, Masatoshi

    1991-01-01

    It is believed that Monte Carlo method is suitable to the calculation of neutron lifetime and generation time with reference to the life cycle viewpoint. This paper illustrates that those times obtained by Monte Carlo method are quite different from the results by perturbation method. The neutron lifetime and the generation time for bare and reflected reactors were investigated by the Monte Carlo program, KENO IV. the Monte Carlo procedure is based on tracking and recording the life history of neutrons in a realistic fashion in a fissionable system with minimum nuclear and geometric approximations. The KENO IV provides the multiplication factor, neutron lifetime and generation time simultaneously. The thermal spherical reactors for both bare and reflected reactors were studied using the KENO IV. The reflected reactor is surrounded with 30 cm thick light water. The atomic densities in the regions and the calculated results of the multiplication factor, neutron lifetime and generation time are given. The different definitions of these times between the Monte Carlo method and perturbation theory caused the difference of the results. (K.I.)

  8. Hepatic imaging in stage IV-S neuroblastoma

    International Nuclear Information System (INIS)

    Franken, E.A. Jr.; Smith, W.L.; Iowa Univ., Iowa City; Cohen, M.D.; Kisker, C.T.; Platz, C.E.

    1986-01-01

    Stage IV-S neuroblastoma describes a group of infants with tumor spread limited to liver, skin, or bone marrow. Such patients, who constitute about 25% of affected infants with neuroblastoma, may expect spontaneous tumor remission. We report 18 infants with Stage IV-S neuroblastoma, 83% of whom had liver involvement. Imaging investigations included Technetium 99m sulfur colloid scan, ultrasound, and CT. Two patterns of liver metastasis were noted: ill-defined nodules or diffuse tumor throughout the liver. Distinction of normal and abnormal liver with diffuse type metastasis could be quite difficult, particularly with liver scans. We conclude that patients with Stage IV-S neuroblastoma have ultrasound or CT examination as an initial workup, with nuclear medicine scans reserved for followup studies. (orig.)

  9. Diaquatetrabromidotin(IV trihydrate

    Directory of Open Access Journals (Sweden)

    Fei Ye

    2012-09-01

    Full Text Available The title compound, [SnBr4(H2O2]·3H2O, forms large colourless crystals in originally sealed samples of tin tetrabromide. It constitutes the first structurally characterized hydrate of SnBr4 and is isostructural with the corresponding hydrate of SnCl4. It is composed of SnIV atoms octahedrally coordinated by four Br atoms and two cis-related water molecules. The octahedra exhibit site symmetry 2. They are arranged into columns along [001] via medium–strong O—H...O hydrogen bonds involving the two lattice water molecules (one situated on a twofold rotation axis while the chains are interconnected via longer O—H...Br hydrogen bonds, forming a three-dimensional network.

  10. Cyclopentadienyluranium(IV) acetylacetonates

    International Nuclear Information System (INIS)

    Bagnall, K.W.; Edwards, J.; Rickard, C.E.F.; Tempest, A.C.

    1979-01-01

    Cyclopentadienyluranium(IV) acetylacetonate complexes, (eta 5 C 5 H 5 )UClsub(3-x)(acac)sub(x), where x = 1 or 2, and the corresponding bis triphenylphosphine oxide (tppo) complexes have been prepared. The bis cyclopentadienyl complexes, (eta 5 C 5 H 5 ) 2 U(acac) 2 and (eta 5 C 5 H 5 ) 2 UCl(acac)(tppo) 2 have also been prepared and are stable with respect to disproportionation, whereas (eta 5 C 5 H 5 ) 2 UCl(acac) is not. The IR and UV/visible spectra of the complexes are reported, together with some additional information on the UCl 2 (acac) 2 thf and -tppo systems. (author)

  11. Reactor Physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, A.

    2002-01-01

    SCK-CEN's Reactor Physics and MYRRHA Department offers expertise in various areas of reactor physics, in particular in neutron and gamma calculations, reactor dosimetry, reactor operation and control, reactor code benchmarking and reactor safety calculations. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 materials testing reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2001 are summarised

  12. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2001-04-01

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

  13. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2002-04-01

    SCK-CEN's Reactor Physics and MYRRHA Department offers expertise in various areas of reactor physics, in particular in neutron and gamma calculations, reactor dosimetry, reactor operation and control, reactor code benchmarking and reactor safety calculations. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 materials testing reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2001 are summarised.

  14. Reactor Physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, A.

    2001-01-01

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

  15. JRC-IE's research of safety of Gen IV systems

    International Nuclear Information System (INIS)

    Tsige-Tamirat, H.; Ranguelova, V.; Feutterer, M.; Ammirabile, L.; Carlsson, J.; D'Agata, E.; Laurie, M.; Magallon, D.

    2010-01-01

    The Institute for Energy (IE), one of the seven scientific Institutes of the Joint Research Centre (JRC) of the European Commission, has the mission to provide scientific and technical support for the conception, development, implementation and monitoring of community policies related to energy. To accomplish its mission, IE performs research in the areas of renewable energies, safety and sustainability of nuclear energy for current and future reactor systems, energy technic/economic assessment, and security of energy supply. The Generation IV International Forum (GIF) is a cooperative international endeavour organized to carry out R and D needed to establish the feasibility and performance capabilities of the next generation nuclear energy systems and support the progress towards their realization. The EU, represented by EURATOM and with the JRC as implementing agent, is working together with other GIF partners to perform pre-competitive R and D on key technologies to be implemented in future nuclear systems. IE is engaged in experimental research, simulation and modeling, scientific, feasibility and engineering studies on innovative nuclear reactor systems needed to support the EURATOM contribution to GEN IV initiative, in particular in assessment of innovative fuels and materials, development of new reactor core concepts and safety solutions and knowledge management and preservation. IE's research activities on Generation IV reactor systems are focused on the assessment of the potential of such systems to meet long term EU energy needs with respect to economical advantages, enhanced safety, sustainability, and proliferation resistance. IE participates in international collaborations and has bilateral research cooperation both with European and non-European partners. This paper gives an overview of IE's current research activities on the Gen IV reactor systems related to safety. (authors)

  16. 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

  17. Reactor safeguards

    CERN Document Server

    Russell, Charles R

    1962-01-01

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

  18. Nuclear reactors

    International Nuclear Information System (INIS)

    Middleton, J.E.

    1977-01-01

    Reference is made to water cooled reactors and in particular to the cooling system of steam generating heavy water reactors (SGHWR). A two-coolant circuit is described for the latter. Full constructural details are given. (U.K.)

  19. Reactor decommissioning

    International Nuclear Information System (INIS)

    Lawton, H.

    1984-01-01

    A pioneering project on the decommissioning of the Windscale Advanced Gas-cooled Reactor, by the UKAEA, is described. Reactor data; policy; waste management; remote handling equipment; development; and recording and timescales, are all briefly discussed. (U.K.)

  20. Temperature capability for DTF-IV (DTF-71 version)

    International Nuclear Information System (INIS)

    Varela, D.W.; Philbin, J.S.

    1977-05-01

    A subroutine has been developed for calculating adiabatic temperature distributions as part of the activity edit in the DTF-IV S/sub N/ neutron transport code. The specific heat input is simple and versatile. The subroutine integrates the specific heat functions and solves for the temperatures which match the energy deposition supplied by the activity edit. There is no heat transfer modeling in the subroutine. The temperatures are only valid for cases where the energy deposition time is compared to the thermal relaxation time of the media. The code can be used, for example, to calculate the initial temperature distributions in certain pulsed reactors or pulsed reactor experiments

  1. Congenital bilateral neuroblastoma (stage IV-S): case report

    International Nuclear Information System (INIS)

    Lee, Jeong Hee; Lee, Hee Jung; Woo, Seong Ku; Lee, Sang Rak; Kim, Heung Sik

    2002-01-01

    Congenital neonatal neuroblastoma is not uncommon but bilateral adrenal neuroblastoma is rare, accounting for about ten percent of neuroblastomas in children. We report the US the MR findings of a stage IV-S congenital bilateral neuroblastoma occurring in a one-day-old neonate

  2. RA Reactor

    International Nuclear Information System (INIS)

    1978-02-01

    In addition to basic characteristics of the RA reactor, organizational scheme and financial incentives, this document covers describes the state of the reactor components after 18 years of operation, problems concerned with obtaining the licence for operation with 80% fuel, problems of spent fuel storage in the storage pool of the reactor building and the need for renewal of reactor equipment, first of all instrumentation [sr

  3. Multiregion reactors

    International Nuclear Information System (INIS)

    Moura Neto, C. de; Nair, R.P.K.

    1979-08-01

    The study of reflected reactors can be done employing the multigroup diffusion method. The neutron conservation equations, inside the intervals, can be written by fluxes and group constants. A reflected reactor (one and two groups) for a slab geometry is studied, aplying the continuity of flux and current in the interface. At the end, the appropriated solutions for a infinite cylindrical reactor and for a spherical reactor are presented. (Author) [pt

  4. GIF (Gen-IV International Forum) Symposium 2009. Proceedings

    International Nuclear Information System (INIS)

    2009-01-01

    The objective of this symposium is to give a well documented state of the art of the initiative and to report and discuss the most significant technical progress and evolution in the different areas during these last ten years. Another significant objective is to provide a forum for an open and hopefully lively discussion of the perspectives, priorities and challenges for the next few years, accounting for a rapidly evolving environment. The symposium has been organized into three sessions that have dealt with the following issues: -) Generation IV International Forum (GIF): 10 years of achievements and the path forward, -) Methodology Overviews and Focus on Applications, -) Very High Temperature Reactor (VHTR), -) Gas-cooled Fast Reactor (GFR), -) Super-Critical Water-cooled Reactor (SCWR), -) Lead-cooled Fast Reactor (LFR), -) Molten Salt Reactor (MSR), -) Sodium-cooled Fast Reactor (SFR), -) International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) and its potential synergy with GIF, and -) GIF priority objectives for the next 5 years

  5. Nuclear reactor

    International Nuclear Information System (INIS)

    Hattori, Sadao; Sato, Morihiko.

    1994-01-01

    Liquid metals such as liquid metal sodium are filled in a reactor container as primary coolants. A plurality of reactor core containers are disposed in a row in the circumferential direction along with the inner circumferential wall of the reactor container. One or a plurality of intermediate coolers are disposed at the inside of an annular row of the reactor core containers. A reactor core constituted with fuel rods and control rods (module reactor core) is contained at the inside of each of the reactor core containers. Each of the intermediate coolers comprises a cylindrical intermediate cooling vessels. The intermediate cooling vessel comprises an intermediate heat exchanger for heat exchange of primary coolants and secondary coolants and recycling pumps for compulsorily recycling primary coolants at the inside thereof. Since a plurality of reactor core containers are thus assembled, a great reactor power can be attained. Further, the module reactor core contained in one reactor core vessel may be small sized, to facilitate the control for the reactor core operation. (I.N.)

  6. The finite element structural analysis code SAP IV conversion from CDC to IBM

    International Nuclear Information System (INIS)

    Harrop, L.P.

    1977-02-01

    SAP IV is a general three dimensional, linear, static and dynamic finite element structural analysis program. The program which was obtained from the Earthquake Engineering Research Center, University of California, Berkeley, was written in FORTRAM for a CDC 6400. Its main use was anticipated to be the seismic analysis of reactor structures. SAP IV may also prove useful for fracture mechanics studies as well as the usual elastic stress analysis of structures. A brief description of SAP IV and a more detailed account of the FORTRAN conversion required to make SAP IV run successfully on the UKAEA Harwell IBM 370/168 are given. (author)

  7. Supercritical-pressure light water cooled reactors

    CERN Document Server

    Oka, Yoshiaki

    2014-01-01

    This book focuses on the latest reactor concepts, single pass core and experimental findings in thermal hydraulics, materials, corrosion, and water chemistry. It highlights research on supercritical-pressure light water cooled reactors (SCWRs), one of the Generation IV reactors that are studied around the world. This book includes cladding material development and experimental findings on heat transfer, corrosion and water chemistry. The work presented here will help readers to understand the fundamental elements of reactor design and analysis methods, thermal hydraulics, materials and water

  8. Euratom contributions in Fast Reactor research programmes

    International Nuclear Information System (INIS)

    Fanghänel, Th.; Somers, J.

    2013-01-01

    The Sustainable Nuclear Initiative: • demonstrate long-term sustainability of nuclear energy; • demonstration reactors of Gen IV: •more efficient use of resources; • closed fuel cycle; • reduced proliferation risks; • enhanced safety features. • Systems pursued in Europe: • Sodium-cooled fast reactor SFR; • Lead-cooled fast reactor LFR; • Gas-cooled fast reactor GFR. Sustainable Nuclear Energy Technology Platform SNE-TP promotes research, development and demonstration of the nuclear fission technologies necessary to achieve the SET-Plan goals

  9. 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.)

  10. Research reactors

    International Nuclear Information System (INIS)

    Merchie, Francois

    2015-10-01

    This article proposes an overview of research reactors, i.e. nuclear reactors of less than 100 MW. Generally, these reactors are used as neutron generators for basic research in matter sciences and for technological research as a support to power reactors. The author proposes an overview of the general design of research reactors in terms of core size, of number of fissions, of neutron flow, of neutron space distribution. He outlines that this design is a compromise between a compact enough core, a sufficient experiment volume, and high enough power densities without affecting neutron performance or its experimental use. The author evokes the safety framework (same regulations as for power reactors, more constraining measures after Fukushima, international bodies). He presents the main characteristics and operation of the two families which represent almost all research reactors; firstly, heavy water reactors (photos, drawings and figures illustrate different examples); and secondly light water moderated and cooled reactors with a distinction between open core pool reactors like Melusine and Triton, pool reactors with containment, experimental fast breeder reactors (Rapsodie, the Russian BOR 60, the Chinese CEFR). The author describes the main uses of research reactors: basic research, applied and technological research, safety tests, production of radio-isotopes for medicine and industry, analysis of elements present under the form of traces at very low concentrations, non destructive testing, doping of silicon mono-crystalline ingots. The author then discusses the relationship between research reactors and non proliferation, and finally evokes perspectives (decrease of the number of research reactors in the world, the Jules Horowitz project)

  11. Reactor physics and reactor computations

    International Nuclear Information System (INIS)

    Ronen, Y.; Elias, E.

    1994-01-01

    Mathematical methods and computer calculations for nuclear and thermonuclear reactor kinetics, reactor physics, neutron transport theory, core lattice parameters, waste treatment by transmutation, breeding, nuclear and thermonuclear fuels are the main interests of the conference

  12. US Nuclear Regulatory Commission region IV

    International Nuclear Information System (INIS)

    Vanderburch, C.

    1996-01-01

    The NRC has established a policy to provide for the timely through and systematic inspection of significant operational events at nuclear power plants. This includes the use of an Augmented Inspection Team to determine the causes, conditions, and circumstances relevant to an event and to communicate its findings and conclusions to NRC management. In accordance with NRC Inspection Manual Chapter 0325. The Region IV Regional Administrator dispatched an Augmented Inspection Team to the Wolf Creek Nuclear Generating Station to review the circumstances surrounding a manual reactor trip on January 30, 1996, with the failure of five control rods to fully insert into the core, a failure of the turbine-driven auxiliary feedwater pump, and the subsequent loss of one train of the essential service water system

  13. Titanium(IV), zirconium, hafnium and thorium

    International Nuclear Information System (INIS)

    Brown, Paul L.; Ekberg, Christian

    2016-01-01

    Titanium can exist in solution in a number of oxidation states. The titanium(IV) exists in acidic solutions as the oxo-cation, TiO 2+ , rather than Ti 4+ . Zirconium is used in the ceramics industry and in nuclear industry as a cladding material in reactors where its reactivity towards hydrolysis reactions and precipitation of oxides may result in degradation of the cladding. In nature, hafnium is found together with zirconium and as a consequence of the contraction in ionic radii that occurs due to the 4f -electron shell, the ionic radius of hafnium is almost identical to that of zirconium. All isotopes of thorium are radioactive and, as a consequence of it being fertile, thorium is important in the nuclear fuel cycle. The polymeric hydrolysis species that have been reported for thorium are somewhat different to those identified for zirconium and hafnium, although thorium does form the Th 4 (OH) 8 8+ species.

  14. Iris reactor development

    International Nuclear Information System (INIS)

    Paramonov, D.V.; Carelli, M.D.; Miller, K.; Lombardi, C.V.; Ricotti, M.E.; Todreas, N.E.; Greenspan, E.; Yamamoto, K.; Nagano, A.; Ninokata, H.; Robertson, J.; Oriolo, F.

    2001-01-01

    The development progress of the IRIS (International Reactor Innovative and Secure) nuclear power system is presented. IRIS is currently being developed by an international consortium of industry, laboratory, university and utility establishments, led by Westinghouse. It is aimed at achieving the four major objectives of the Generation IV nuclear systems, i.e., proliferation resistance, enhanced safety, economic competitiveness and reduced waste. The project first year activities, which are summarized here, were focused on core neutronics, in-vessel configuration, steam generator and containment design, safety approach and economic performance. Details of these studies are provided in parallel papers in these proceedings. (author)

  15. Glycogen Storage Disease Type IV

    DEFF Research Database (Denmark)

    Bendroth-Asmussen, Lisa; Aksglaede, Lise; Gernow, Anne B

    2016-01-01

    molecular genetic analyses confirmed glycogen storage disease Type IV with the finding of compound heterozygosity for 2 mutations (c.691+2T>C and c.1570C>T, p.R524X) in the GBE1 gene. We conclude that glycogen storage disease Type IV can cause early miscarriage and that diagnosis can initially be made...

  16. About the structure and stability of complex carbonates of thorium (IV), cerium (IV), zirconium (IV), hafnium (IV)

    International Nuclear Information System (INIS)

    Dervin, Jacqueline

    1972-01-01

    This research thesis addressed the study of complex carbonates of cations of metals belonging to the IV A column, i.e. thorium (IV), zirconium (IV), hafnium (IV), and also cerium (IV) and uranium (VI), and more particularly focused on ionic compounds formed in solution, and also on the influence of concentration and nature of cations on stability and nature of the formed solid. The author first presents methods used in this study, discusses their precision and scope of validity. She reports the study of the formation of different complex ions which have been highlighted in solution, and the determination of their formation constants. She reports the preparation and study of the stability domain of solid complexes. The next part reports the use of thermogravimetric analysis, IR spectrometry, and crystallography for the structural study of these compounds

  17. Strategy of implementation of generation IV reactors in Slovak fleet

    International Nuclear Information System (INIS)

    Liska, P.

    2010-01-01

    In this presentation author deals with electric power systems of Europe as well as with electric power system of the Slovak Republic. Anticipated shut downs of Slovak NPPs and start up of new NPPs are reviewed.

  18. Fast reactor safety and related physics. Volume IV. Phenomenology

    Energy Technology Data Exchange (ETDEWEB)

    1976-01-01

    Separate abstracts are included for 58 papers concerning single-phase flow and sodium boiling; sodium boiling and subassembly flow blockages; transient-overpower and loss-of-flow experiments; fuel and cladding behavior and relocation; fuel and cladding freezing; molten-fuel-coolant interaction; aerosols and fission product release, and post-accident heat removal. Thirteen papers have been perivously abstracted and included in ERA.

  19. Liquid Metal Fast Breeder Reactor program. Volume IV. Environmental statement

    International Nuclear Information System (INIS)

    1974-12-01

    A broad overview is presented of the many implications of LMFBR program implementation, up to and encompassing a fully developed LMFBR power plant economy, including the secondary impacts, the unavoidable adverse environmental impacts, cumulative environmental impacts, and cost-benefit analyses, and alternative energy strategies. Under the heading of secondary impacts, the national implications of the availability of electricity from LMFBRs, and the specific economic impacts of the LMFBR program are examined. The currently feasible alternatives and potential future alternatives for mitigating adverse environmental impacts of the LMFBR fuel cycle are described. The problems of safeguarding special nuclear material from potential diversion to unauthorized purposes are analyzed. The cumulative environmental effects of LMFBR operation to the Year 2020, the decommissioning of LMFBRs and fuel cycle facilities upon the completion of their useful life, the irreversible and irretrievable commitments of resources that will accompany implementation of an LMFBR economy, and an analysis of the costs and benefits of implementing the LMFBR Program are included. (U.S.)

  20. AREVA's nuclear reactors portfolio

    International Nuclear Information System (INIS)

    Marincic, A.

    2009-01-01

    A reasonable assumption for the estimated new build market for the next 25 years is over 340 GWe net. The number of prospect countries is growing almost each day. To address this new build market, AREVA is developing a comprehensive portfolio of reactors intended to meet a wide range of power requirements and of technology choices. The EPR reactor is the flagship of the fleet. Intended for large power requirements, the four first EPRs are being built in Finland, France and China. Other countries and customers are in view, citing just two examples: the Usa where the U.S. EPR has been selected as the technology of choice by several U.S utilities; and the United Kingdom where the Generic Design Acceptance process of the EPR design submitted by AREVA and EDF is well under way, and where there is a strong will to have a plant on line in 2017. For medium power ranges, the AREVA portfolio includes a boiling water reactor and a pressurized water reactor which both offer all of the advantages of an advanced plant design, with excellent safety performance and competitive power generation cost: -) KERENA (1250+ MWe), developed in collaboration with several European utilities, and in particular with Eon; -) ATMEA 1 (1100+ MWe), a 3-loop evolutionary PWR which is being developed by AREVA and Mitsubishi. AREVA is also preparing the future and is deeply involved into Gen IV concepts. It has developed the ANTARES modular HTR reactor (pre-conceptual design completed) and is building upon its vast Sodium Fast Reactor experience to take part into the development of the next prototype. (author)

  1. Research reactors

    International Nuclear Information System (INIS)

    Kowarski, L.

    1955-01-01

    It brings together the techniques data which are involved in the discussion about the utility for a research institute to acquire an atomic reactor for research purposes. This type of decision are often taken by non-specialist people who can need a brief presentation of a research reactor and its possibilities in term of research before asking advises to experts. In a first part, it draws up a list of the different research programs which can be studied by getting a research reactor. First of all is the reactor behaviour and kinetics studies (reproducibility factor, exploration of neutron density, effect of reactor structure, effect of material irradiation...). Physical studies includes study of the behaviour of the control system, studies of neutron resonance phenomena and study of the fission process for example. Chemical studies involves the study of manipulation and control of hot material, characterisation of nuclear species produced in the reactor and chemical effects of irradiation on chemical properties and reactions. Biology and medicine research involves studies of irradiation on man and animals, genetics research, food or medical tools sterilization and neutron beams effect on tumour for example. A large number of other subjects can be studied in a reactor research as reactor construction material research, fabrication of radioactive sources for radiographic techniques or applied research as in agriculture or electronic. The second part discussed the technological considerations when choosing the reactor type. The technological factors, which are considered for its choice, are the power of the reactor, the nature of the fuel which is used, the type of moderator (water, heavy water, graphite or BeO) and the reflector, the type of coolants, the protection shield and the control systems. In the third part, it described the characteristics (place of installation, type of combustible and comments) and performance (power, neutron flux ) of already existing

  2. ARIES-IV Nested Shell Blanket Design

    International Nuclear Information System (INIS)

    Wong, C.P.C.; Redler, K.; Reis, E.E.; Will, R.; Cheng, E.; Hasan, C.M.; Sharafat, S.

    1993-11-01

    The ARIES-IV Nested Shell Blanket (NSB) Design is an alternate blanket concept of the ARIES-IV low activation helium-cooled reactor design. The reference design has the coolant routed in the poloidal direction and the inlet and outlet plena are located at the top and bottom of the torus. The NSB design has the high velocity coolant routed in the toroidal direction and the plena are located behind the blanket. This is of significance since the selected structural material is SiC-composite. The NSB is designed to have key high performance components with characteristic dimensions of no larger than 2 m. These components can be brazed to form the blanket module. For the diverter design, we eliminated the use of W as the divertor coating material by relying on the successful development of the gaseous divertor concept. The neutronics and thermal-hydraulic performance of both blanket concepts are similar. The selected blanket and divertor configurations can also meet all the projected structural, neutronics and thermal-hydraulics design limits and requirements. With the selected blanket and divertor materials, the design has a level of safety assurance rate of I (LSA-1), which indicates an inherently safe design

  3. Neutronic calculation of reactor cells

    International Nuclear Information System (INIS)

    Jaliff, J.O.

    1981-01-01

    Multigroup calculations of cylindrical pin cells were programmed, in Fortran IV, upon the basis of collision probabilities in each energy group. A rational approximation to the fuel-to-fuel collision probability in resonance groups was used. Together with the intermediate resonance theory, cross sections corrected for heterogeneity and absorber interactions were found. For the optimization of the program, the cell of a BWR reactor was taken as reference. Data for such a cell and the reactor's operating conditions are presented. PINCEL is a fast and flexible program, with checked results, around a 69-group library. (M.E.L.) [es

  4. Reactor container

    International Nuclear Information System (INIS)

    Naruse, Yoshihiro.

    1990-01-01

    The thickness of steel shell plates in a reactor container embedded in sand cussions is monitored to recognize the corrosion of the steel shell plates. That is, the reactor pressure vessel is contained in a reactor container shell and the sand cussions are disposed on the lower outside of the reactor container shell to elastically support the shell. A pit is disposed at a position opposing to the sand cussions for measuring the thickness of the reactor container shell plates. The pit is usually closed by a closing member. In the reactor container thus constituted, the closing member can be removed upon periodical inspection to measure the thickness of the shell plates. Accordingly, the corrosion of the steel shell plates can be recognized by the change of the plate thickness. (I.S.)

  5. Hybrid reactors

    International Nuclear Information System (INIS)

    Moir, R.W.

    1980-01-01

    The rationale for hybrid fusion-fission reactors is the production of fissile fuel for fission reactors. A new class of reactor, the fission-suppressed hybrid promises unusually good safety features as well as the ability to support 25 light-water reactors of the same nuclear power rating, or even more high-conversion-ratio reactors such as the heavy-water type. One 4000-MW nuclear hybrid can produce 7200 kg of 233 U per year. To obtain good economics, injector efficiency times plasma gain (eta/sub i/Q) should be greater than 2, the wall load should be greater than 1 MW.m -2 , and the hybrid should cost less than 6 times the cost of a light-water reactor. Introduction rates for the fission-suppressed hybrid are usually rapid

  6. Nuclear reactor

    International Nuclear Information System (INIS)

    Garabedian, G.

    1988-01-01

    A liquid reactor is described comprising: (a) a reactor vessel having a core; (b) one or more satellite tanks; (c) pump means in the satellite tank; (d) heat exchanger means in the satellite tank; (e) an upper liquid metal conduit extending between the reactor vessel and the satellite tank; (f) a lower liquid metal duct extending between the reactor vessel and satellite tanks the upper liquid metal conduit and the lower liquid metal duct being arranged to permit free circulation of liquid metal between the reactor vessel core and the satellite tank by convective flow of liquid metal; (g) a separate sealed common containment vessel around the reactor vessel, conduits and satellite tanks; (h) the satellite tank having space for a volume of liquid metal that is sufficient to dampen temperature transients resulting from abnormal operating conditions

  7. Advanced Reactors Around the World

    International Nuclear Information System (INIS)

    Majumdar, Debu

    2003-01-01

    At the end of 2002, 441 nuclear power plants were operating around the globe and providing 17% of the world's electricity. Although the rate of population growth has slowed, recent United Nations data suggest that two billion more people will be added to the world by 2050. A special report commissioned by the Intergovernmental Panel on Climate Change estimated that electricity demand would grow almost eight-fold from 2000 to 2050 in a high economic grown scenario and more than double in a low-growth scenario. There is also a global aspiration to keep the environment pristine. Because of these reasons, it is expected that a large number of new nuclear reactors may be operating by 2050. Realization of this has created an impetus for the development of a new generation of reactors in several countries. The goal is to make nuclear power cost-competitive with other resources and to enhance safety to a level that no evacuation outside a plant site would be necessary. It should also generate less waste, prevent materials diversion for weapons production, and be sustainable. This article discusses the status of next-generation reactors under development around the world. Specifically highlighted are efforts related to the Generation IV International Forum (GIF) and its six reactor concepts for research and development: Very High Temperature Reactor (VHTR); Gas-Cooled Fast Reactor (GFR); Supercritical Water-Cooled Reactor (SCWR); Sodium-Cooled Fast Reactor (SFR); Lead-Cooled Fast Reactor (LFR); and Molten Salt Reactor (MSR). Also highlighted are nuclear activities specific to Russia and India

  8. Nuclear reactor

    International Nuclear Information System (INIS)

    Batheja, P.; Huber, R.; Rau, P.

    1985-01-01

    Particularly for nuclear reactors of small output, the reactor pressure vessel contains at least two heat exchangers, which have coolant flowing through them in a circuit through the reactor core. The circuit of at least one heat exchanger is controlled by a slide valve, so that even for low drive forces, particularly in natural circulation, the required even loading of the heat exchanger is possible. (orig./HP) [de

  9. IRIS Responsiveness to Generation IV Road-map Goals

    International Nuclear Information System (INIS)

    Carelli, M.D.; Paramonov, D.V.; Petrovic, B.

    2002-01-01

    The DOE Generation IV road-map process is in its second and final year. Almost one hundred concepts submitted from all over the world have been reviewed against the Generation IV goals of resources sustainability; safety and reliability; and, economics. Advanced LWRs are taken as the reference point. IRIS (International Reactor Innovative and Secure), a 100-335 MWe integral light water reactor being developed by a vast international consortium led by Westinghouse, is one on the concepts being considered in the road-map and is perhaps the most visible representative of the concept set known as Integral Primary System Reactors (IPSR). This paper presents how IRIS satisfies the prescribed goals. The first goal of resource sustainability includes criteria like utilization of fuel resources, amount and toxicity of waste produced, environmental impact, proliferation and sabotage resistance. As a thermal reactor IRIS does not have the same fuel utilization as fast reactors. However, it has a significant flexibility in fuel cycles as it is designed to utilize either UO 2 or MOX with straight burn cycles of 4 to 10 years, depending on the fissile content. High discharge burnup and Pu recycling result in good fuel utilization and lower waste; IRIS has also attractive proliferation resistance characteristics, due to the reduced accessibility of the fuel. The safety and reliability goal include reliability, workers' exposure, robust safety features, models with well characterized uncertainty, source term and mechanisms of energy release, robust mitigation of accidents. IRIS is significantly better than advanced LWRs because of its safety by design which eliminates a variety of accidents such as LOCAs, its containment vessel coupled design which maintains the core safely covered during the accident sequences, its design simplification features such as no (or reduced) soluble boron, internal shielding and four-year refueling/maintenance interval which significantly reduce

  10. Strategies of development of reactor types

    International Nuclear Information System (INIS)

    Bacher, P.

    2004-01-01

    The development of nuclear energy in the coming decades will depend on the goals followed, on the available technologies and on the strategies implemented in the world in agreement with public acceptation. This article is limited to the technical aspects of the strategies of development of reactor types: 1 - objectives; 2 - common constraints to all reactor types: safety and terrorism risks, wastes, non-proliferation, economics; 3 - different reactor types: general considerations, proven technologies (PWR, BWR, Candu), non-proven technologies but having an important experience, technologies at the design stage; 4 - energy systems and 'Generation IV forum': systems based on thermal neutron reactors and low enrichment, systems for the valorization of 238 U, systems for Pu burning, systems allowing the destruction of minor actinides, thorium-based systems, the Gen IV international forum; 5 - conclusion. (J.S.)

  11. Heterogeneous reactors

    International Nuclear Information System (INIS)

    Moura Neto, C. de; Nair, R.P.K.

    1979-08-01

    The microscopic study of a cell is meant for the determination of the infinite multiplication factor of the cell, which is given by the four factor formula: K(infinite) = n(epsilon)pf. The analysis of an homogeneous reactor is similar to that of an heterogeneous reactor, but each factor of the four factor formula can not be calculated by the formulas developed in the case of an homogeneous reactor. A great number of methods was developed for the calculation of heterogeneous reactors and some of them are discussed. (Author) [pt

  12. Direct Bandgap Group IV Materials

    Science.gov (United States)

    2016-01-21

    AFRL-AFOSR-JP-TR-2017-0049 Direct Bandgap group IV Materials Hung Hsiang Cheng NATIONAL TAIWAN UNIVERSITY Final Report 01/21/2016 DISTRIBUTION A...NAME(S) AND ADDRESS(ES) NATIONAL TAIWAN UNIVERSITY 1 ROOSEVELT RD. SEC. 4 TAIPEI CITY, 10617 TW 8. PERFORMING ORGANIZATION REPORT NUMBER 9. SPONSORING...14. ABSTRACT Direct bandgap group IV materials have been long sought for in both academia and industry for the implementation of photonic devices

  13. Issues of high-burnup fuel for advanced nuclear reactors

    International Nuclear Information System (INIS)

    Belac, J.; Milisdoerfer, L.

    2004-12-01

    A brief description is given of nuclear fuels for Generation III+ and IV reactors, and the major steps needed for a successful implementation of new fuels in prospective types of newly designed power reactors are outlined. The following reactor types are discussed: gas cooled fast reactors, heavy metal (lead) cooled fast reactors, molten salt cooled reactors, sodium cooled fast reactors, supercritical water cooled reactors, and very high temperature reactors. The following are regarded as priority areas for future investigations: (i) spent fuel radiotoxicity; (ii) proliferation volatility; (iii) neutron physics characteristics and inherent safety element assessment; technical and economic analysis of the manufacture of advanced fuels; technical and economic analysis of the fuel cycle back end, possibilities of spent nuclear fuel reprocessing, storage and disposal. In parallel, work should be done on the validation and verification of analytical tools using existing and/or newly acquired experimental data. (P.A.)

  14. Development of Electromagnetic Analysis Model for IV-CEAPI

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jinseok; Jang, Yongtae; Lee, Myounggoo; Cho, Yeonho; Kim, Hyunmin [KEPCO Engineering and Construction, Inc., Daejeon (Korea, Republic of); Hong, Hoonbin; Baek, Minho [Woojin Inc., Osan (Korea, Republic of)

    2016-05-15

    There are many different types of position indicators such as reed switch type, ultrasonic type, solenoid type, etc. Through an analysis of strengths and weakness of those types, solenoid type was selected for an IV-CEAPI. Although solenoid type CEAPIs have been used world-wide, the IV-CEAPI is to be very different from the conventional designs due to its harsh operating environment. The concept of the IV-CEAPI is simple as shown in Figure 1. The coil is made of mineral insulated wire to be able to operate inside reactor vessel. The CEA is connected to the shaft which is made of ferromagnetic material. As the CEA position varies, the inductance variation is detected by the inductance meter located outside the vessel. Unlike the conventional ones, the IV-CEAPI used only one coil to eliminate coil connection point and electric components inside vessel. A finite element model was developed to calculate inductance of the solenoid type IV-CEAPI. The model considers eddy current effect to calculate frequency dependent inductance value. Analyses were performed to produce an inductance curve to the shaft position.

  15. Development of Electromagnetic Analysis Model for IV-CEAPI

    International Nuclear Information System (INIS)

    Park, Jinseok; Jang, Yongtae; Lee, Myounggoo; Cho, Yeonho; Kim, Hyunmin; Hong, Hoonbin; Baek, Minho

    2016-01-01

    There are many different types of position indicators such as reed switch type, ultrasonic type, solenoid type, etc. Through an analysis of strengths and weakness of those types, solenoid type was selected for an IV-CEAPI. Although solenoid type CEAPIs have been used world-wide, the IV-CEAPI is to be very different from the conventional designs due to its harsh operating environment. The concept of the IV-CEAPI is simple as shown in Figure 1. The coil is made of mineral insulated wire to be able to operate inside reactor vessel. The CEA is connected to the shaft which is made of ferromagnetic material. As the CEA position varies, the inductance variation is detected by the inductance meter located outside the vessel. Unlike the conventional ones, the IV-CEAPI used only one coil to eliminate coil connection point and electric components inside vessel. A finite element model was developed to calculate inductance of the solenoid type IV-CEAPI. The model considers eddy current effect to calculate frequency dependent inductance value. Analyses were performed to produce an inductance curve to the shaft position

  16. Technical Survey and Feasibility Review for Development of IV-CEAPI

    International Nuclear Information System (INIS)

    Jang, Yongtae; Park, Jinseok; Lee, Myounggoo; Cho, Yeonho; Kim, Hyunmin

    2016-01-01

    The purpose of this paper is to establish the development direction of the IV-CEAPI(Control element assembly position indicator). The paper presents the technologies of the existing CEAPI and other linear displacement sensors. The paper also presents feasibility review of those technologies for the IV-CEAPI considering its environmental conditions as shown in Table 1. an instrument to monitor vertical position of the control element assembly (CEA) in nuclear reactors. The CEAPI is installed in each control element drive mechanism (CEDM). The conventional CEDMs are installed outside the reactor vessel (RV) with nozzles penetrating the RV head. To select the type of the IV-CEAPI, technical surveys on linear displacement sensors were performed. Feasibility of those sensors was reviewed considering the environment conditions, experience, reliability and simplicity. The result is summarized in Table 2 which implies that the solenoid type is considered to be the best suitable types for the IV-CEAPI

  17. DEVELOPMENT OF RISK-BASED AND TECHNOLOGY-INDEPENDENT SAFETY CRITERIA FOR GENERATION IV SYSTEMS

    Energy Technology Data Exchange (ETDEWEB)

    William E. Kastenberg; Edward Blandford; Lance Kim

    2009-03-31

    This project has developed quantitative safety goals for Generation IV (Gen IV) nuclear energy systems. These safety goals are risk based and technology independent. The foundations for a new approach to risk analysis has been developed, along with a new operational definition of risk. This project has furthered the current state-of-the-art by developing quantitative safety goals for both Gen IV reactors and for the overall Gen IV nuclear fuel cycle. The risk analysis approach developed will quantify performance measures, characterize uncertainty, and address a more comprehensive view of safety as it relates to the overall system. Appropriate safety criteria are necessary to manage risk in a prudent and cost-effective manner. This study is also important for government agencies responsible for managing, reviewing, and for approving advanced reactor systems because they are charged with assuring the health and safety of the public.

  18. DEVELOPMENT OF RISK-BASED AND TECHNOLOGY-INDEPENDENT SAFETY CRITERIA FOR GENERATION IV SYSTEMS

    International Nuclear Information System (INIS)

    Kastenberg, William E.; Blandford, Edward; Kim, Lance

    2009-01-01

    This project has developed quantitative safety goals for Generation IV (Gen IV) nuclear energy systems. These safety goals are risk based and technology independent. The foundations for a new approach to risk analysis has been developed, along with a new operational definition of risk. This project has furthered the current state-of-the-art by developing quantitative safety goals for both Gen IV reactors and for the overall Gen IV nuclear fuel cycle. The risk analysis approach developed will quantify performance measures, characterize uncertainty, and address a more comprehensive view of safety as it relates to the overall system. Appropriate safety criteria are necessary to manage risk in a prudent and cost-effective manner. This study is also important for government agencies responsible for managing, reviewing, and for approving advanced reactor systems because they are charged with assuring the health and safety of the public

  19. Slurry reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kuerten, H; Zehner, P [BASF A.G., Ludwigshafen am Rhein (Germany, F.R.)

    1979-08-01

    Slurry reactors are designed on the basis of empirical data and model investigations. It is as yet not possible to calculate the flow behavior of such reactors. The swarm of gas bubbles and cluster formations of solid particles and their interaction in industrial reactors are not known. These effects control to a large extent the gas hold-up, the gas-liquid interface and, similarly as in bubble columns, the back-mixing of liquids and solids. These hydrodynamic problems are illustrated in slurry reactors which constructionally may be bubble columns, stirred tanks or jet loop reactors. The expected effects are predicted by means of tests with model systems modified to represent the conditions in industrial hydrogenation reactors. In his book 'Mass Transfer in Heterogeneous Catalysis' (1970) Satterfield complained of the lack of knowledge about the design of slurry reactors and hence of the impossible task of the engineer who has to design a plant according to accepted rules. There have been no fundamental changes since then. This paper presents the problems facing the engineer in designing slurry reactors, and shows new development trends.

  20. 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

  1. Nuclear reactor

    International Nuclear Information System (INIS)

    Mysels, K.J.; Shenoy, A.S.

    1976-01-01

    A nuclear reactor is described in which the core consists of a number of fuel regions through each of which regulated coolant flows. The coolant from neighbouring fuel regions is combined in a manner which results in an averaging of the coolant temperature at the outlet of the core. By this method the presence of hot streaks in the reactor is reduced. (UK)

  2. Reactor container

    International Nuclear Information System (INIS)

    Kato, Masami; Nishio, Masahide.

    1987-01-01

    Purpose: To prevent the rupture of the dry well even when the melted reactor core drops into a reactor pedestal cavity. Constitution: In a reactor container in which a dry well disposed above the reactor pedestal cavity for containing the reactor pressure vessel and a torus type suppression chamber for containing pressure suppression water are connected with each other, the pedestal cavity and the suppression chamber are disposed such that the flow level of the pedestal cavity is lower than the level of the pressure suppression water. Further, a pressure suppression water introduction pipeway for introducing the pressure suppression water into the reactor pedestal cavity is disposed by way of an ON-OFF valve. In case if the melted reactor core should fall into the pedestal cavity, the ON-OFF valve for the pressure suppression water introduction pipeway is opened to introduce the pressure suppression water in the suppression chamber into the pedestal cavity to cool the melted reactor core. (Ikeda, J.)

  3. RA Reactor

    International Nuclear Information System (INIS)

    1989-01-01

    This chapter includes the following: General description of the RA reactor, organization of work, responsibilities of leadership and operators team, regulations concerning operation and behaviour in the reactor building, regulations for performing experiments, regulations and instructions for inserting samples into experimental channels [sr

  4. Reactor physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, H.

    1998-01-01

    Progress in research on reactor physics in 1997 at the Belgian Nuclear Research Centre SCK/CEN is described. Activities in the following four domains are discussed: core physics, ex-core neutron transport, experiments in Materials Testing Reactors, international benchmarks

  5. A combined XAFS, ESI TOF-MS and LIBD study on the formation of polynuclear Zr(IV), Th(IV) and Pu(IV) species

    Science.gov (United States)

    Rothe, J.; Walther, C.; Brendebach, B.; Büchner, S.; Fuss, M.; Denecke, M. A.; Geckeis, H.

    2009-11-01

    The long term radiotoxicity of spent nuclear fuel disposed of in deep underground repositories after discharge from nuclear power reactors is determined by actinide elements, mainly plutonium. Water intrusion into the repository might cause container corrosion and leaching of the waste matrices, leading to the release of Pu and other actinides into the geological environment. Performance assessment for a future nuclear waste repository requires detailed knowledge on actinide aqueous chemistry in the aquifer surrounding the disposal site. Tetravalent actinides exhibit a strong tendency towards hydrolysis and subsequent polymerization and/or colloid formation. These species provide a potential pathway for migration of actinides away from the repository. Therefore, it is of fundamental interest to study their generation and properties in-situ. To this end, X-ray Absorption Fine Structure Spectroscopy (XAFS) at the INE-Beamline for actinide research at ANKA, Electrospray Mass-Spectrometry (ESI TOF-MS) and Laser Induced Breakdown Detection (LIBD) are combined at FZK-INE in a comprehensive attempt to characterize Zr(IV) (An(IV) analogue), Th(IV) and Pu(IV) polymerization and colloid formation.

  6. Reactor core

    International Nuclear Information System (INIS)

    Azekura, Kazuo; Kurihara, Kunitoshi.

    1992-01-01

    In a BWR type reactor, a great number of pipes (spectral shift pipes) are disposed in the reactor core. Moderators having a small moderating cross section (heavy water) are circulated in the spectral shift pipes to suppress the excess reactivity while increasing the conversion ratio at an initial stage of the operation cycle. After the intermediate stage of the operation cycle in which the reactor core reactivity is lowered, reactivity is increased by circulating moderators having a great moderating cross section (light water) to extend the taken up burnup degree. Further, neutron absorbers such as boron are mixed to the moderator in the spectral shift pipe to control the concentration thereof. With such a constitution, control rods and driving mechanisms are no more necessary, to simplify the structure of the reactor core. This can increase the fuel conversion ratio and control great excess reactivity. Accordingly, a nuclear reactor core of high conversion and high burnup degree can be attained. (I.N.)

  7. Reactor container

    International Nuclear Information System (INIS)

    Fukazawa, Masanori.

    1991-01-01

    A system for controlling combustible gases, it has been constituted at present such that the combustible gases are controlled by exhausting them to the wet well of a reactor container. In this system, however, there has been a problem, in a reactor container having plenums in addition to the wet well and the dry well, that the combustible gases in such plenums can not be controlled. In view of the above, in the present invention, suction ports or exhaust ports of the combustible gas control system are disposed to the wet well, the dry well and the plenums to control the combustible gases in the reactor container. Since this can control the combustible gases in the entire reactor container, the integrity of the reactor container can be ensured. (T.M.)

  8. Reactor container

    International Nuclear Information System (INIS)

    Kojima, Yoshihiro; Hosomi, Kenji; Otonari, Jun-ichiro.

    1997-01-01

    In the present invention, a catalyst for oxidizing hydrogen to be disposed in a reactor container upon rupture of pipelines of a reactor primary coolant system is prevented from deposition of water droplets formed from a reactor container spray to suppress elevation of hydrogen concentration in the reactor container. Namely, a catalytic combustion gas concentration control system comprises a catalyst for oxidizing hydrogen and a support thereof. In addition, there is also disposed a water droplet deposition-preventing means for preventing deposition of water droplets in a reactor pressure vessel on the catalyst. Then, the effect of the catalyst upon catalytic oxidation reaction of hydrogen can be kept high. The local elevation of hydrogen concentration can be prevented even upon occurrence of such a phenomenon that various kinds of mobile forces in the container such as dry well cooling system are lost. (I.S.)

  9. Nuclear reactor

    International Nuclear Information System (INIS)

    Tilliette, Z.

    1975-01-01

    A description is given of a nuclear reactor and especially a high-temperature reactor in which provision is made within a pressure vessel for a main cavity containing the reactor core and a series of vertical cylindrical pods arranged in spaced relation around the main cavity and each adapted to communicate with the cavity through two collector ducts or headers for the primary fluid which flows downwards through the reactor core. Each pod contains two superposed steam-generator and circulator sets disposed in substantially symmetrical relation on each side of the hot primary-fluid header which conveys the primary fluid from the reactor cavity to the pod, the circulators of both sets being mounted respectively at the bottom and top ends of the pod

  10. Computer simulation of the NASA water vapor electrolysis reactor

    Science.gov (United States)

    Bloom, A. M.

    1974-01-01

    The water vapor electrolysis (WVE) reactor is a spacecraft waste reclamation system for extended-mission manned spacecraft. The WVE reactor's raw material is water, its product oxygen. A computer simulation of the WVE operational processes provided the data required for an optimal design of the WVE unit. The simulation process was implemented with the aid of a FORTRAN IV routine.

  11. Reactor building

    International Nuclear Information System (INIS)

    Maruyama, Toru; Murata, Ritsuko.

    1996-01-01

    In the present invention, a spent fuel storage pool of a BWR type reactor is formed at an upper portion and enlarged in the size to effectively utilize the space of the building. Namely, a reactor chamber enhouses reactor facilities including a reactor pressure vessel and a reactor container, and further, a spent fuel storage pool is formed thereabove. A second spent fuel storage pool is formed above the auxiliary reactor chamber at the periphery of the reactor chamber. The spent fuel storage pool and the second spent fuel storage pool are disposed in adjacent with each other. A wall between both of them is formed vertically movable. With such a constitution, the storage amount for spent fuels is increased thereby enabling to store the entire spent fuels generated during operation period of the plant. Further, since requirement of the storage for the spent fuels is increased stepwisely during periodical exchange operation, it can be used for other usage during the period when the enlarged portion is not used. (I.S.)

  12. Reactor container

    International Nuclear Information System (INIS)

    Shibata, Satoru; Kawashima, Hiroaki

    1984-01-01

    Purpose: To optimize the temperature distribution of the reactor container so as to moderate the thermal stress distribution on the reactor wall of LMFBR type reactor. Constitution: A good heat conductor (made of Al or Cu) is appended on the outer side of the reactor container wall from below the liquid level to the lower face of a deck plate. Further, heat insulators are disposed to the outside of the good heat conductor. Furthermore, a gas-cooling duct is circumferentially disposed at the contact portion between the good heat conductor and the deck plate around the reactor container. This enables to flow the cold heat from the liquid metal rapidly through the good heat conductor to the cooling duct and allows to maintain the temperature distribution on the reactor wall substantially linear even with the abrupt temperature change in the liquid metal. Further, by appending the good heat conductor covered with inactive metals not only on the outer side but also on the inside of the reactor wall to introduce the heat near the liquid level to the upper portion and escape the same to the cooling layer below the roof slab, the effect can be improved further. (Ikeda, J.)

  13. Engineering design of the Aries-IV gaseous divertor

    International Nuclear Information System (INIS)

    Hasan, M.Z.; Najmabadi, F.; Sharafat, S.

    1994-01-01

    ARIES-IV is a conceptual, D-T burning, steady-state tokamak fusion reactor producing 1000 MWe net. It operates in the second plasma stability regime. The structural material is SiC composite and the primary coolant is helium at 10MPa base pressure. ARIES-IV uses double-null divertors for particle control. Total thermal power recovered from the divertors is 425MW, which is 16% of the total reactor thermal power. Among the desirable goals of divertor design were to avoid the use of tungsten and to use the same structural material and primary coolant as in the blanket design. In order to reduce peak heat flux, the innovative gaseous divertor has been used in ARIES-IV. A gaseous divertor reduces peak heat flux by increasing the surface area and by distributing particle and radiation energy more uniformly. Another benefit of gaseous divertor is the reduction of plasma temperature in the divertor chamber, so that material erosion due to sputtering, can be diminished. This makes the use of low-Z material possible in a gaseous divertor

  14. Nuclear reactor

    International Nuclear Information System (INIS)

    Rau, P.

    1980-01-01

    The reactor core of nuclear reactors usually is composed of individual elongated fuel elements that may be vertically arranged and through which coolant flows in axial direction, preferably from bottom to top. With their lower end the fuel elements gear in an opening of a lower support grid forming part of the core structure. According to the invention a locking is provided there, part of which is a control element that is movable along the fuel element axis. The corresponding locking element is engaged behind a lateral projection in the opening of the support grid. The invention is particularly suitable for breeder or converter reactors. (orig.) [de

  15. Free-format RPG IV

    CERN Document Server

    Martin, Jim

    2013-01-01

    This how-to guide offers a concise and thorough introduction to the increased productivity, better readability, and easier program maintenance that comes with the free-format style of programming in RPG IV. Although free-format information is available in IBM manuals, it is not separated from everything else, thereby requiring hours of tedious research to track down the information needed. This book provides everything one needs to know to write RPG IV in the free-format style, and author Jim Martin not only teaches rules and syntax but also explains how this new style of coding has the pot

  16. Nuclear fission today and tomorrow: from renaissance to technological breakthrough (Generation IV)

    International Nuclear Information System (INIS)

    Van Goethem, G.

    2010-01-01

    This paper describes briefly the major scientific and technological challenges related to the very innovative nuclear fission reactor systems to be deployed at the horizon 2040 (called Generation IV). The paper focuses on the benefits of the Generation IV systems, according to criteria or technology goals established at the international level (Generation IV International Forum (GIF)). This goals are drastic improvements on four areas: sustainable development, industrial competitiveness, safety and reliability and proliferation resistance. The focus is on the design objectives and associated research issues that have been agreed upon internationally to meet these four ambitious goals. (author)

  17. Massive computation methodology for reactor operation (MACRO)

    International Nuclear Information System (INIS)

    Gustavsson, Cecilia; Pomp, Stephan; Sjoestrand, Henrik; Wallin, Gustav; Oesterlund, Michael; Koning, Arjan; Rochman, Dimitri; Bejmer, Klaes-Hakan; Henriksson, Hans

    2010-01-01

    Today, nuclear data libraries do not handle uncertainties from nuclear data in a consistent manner and the reactor codes do not request uncertainties in nuclear data input. Thus, the output from these codes have unknown uncertainties. The plan is to use a method proposed by Koning and Rochman to investigate the propagation of nuclear data uncertainties into reactor physics codes and macroscopic parameters. A project (acronym MACRO) has started at Uppsala University in collaboration with A. Koning and with financial support from Vattenfall AB and the Swedish Research Council within the GENIUS (Generation IV research in universities of Sweden) project. In the proposed method the uncertainties in nuclear model parameters will be derived from theoretical considerations and comparisons of nuclear model results with experimental cross-section data. Given the probability distribution in the model parameters a large set of random, complete ENDF-formatted nuclear data libraries will be created using the TALYS code. The generated nuclear data libraries will then be used in neutron transport codes to obtain macroscopic reactor parameters. For this, models of reactor systems with proper geometry and elements will be used. This will be done for all data libraries and the variation of the final results will be regarded as a systematic uncertainty in the investigated reactor parameter. The understanding of these systematic uncertainties is especially important for the design and intercomparison of new reactor concepts, i.e., Generation IV, and optimization applications for current generation reactors is envisaged. (authors)

  18. Massive computation methodology for reactor operation (MACRO)

    Energy Technology Data Exchange (ETDEWEB)

    Gustavsson, Cecilia; Pomp, Stephan; Sjoestrand, Henrik; Wallin, Gustav; Oesterlund, Michael [Division of applied nuclear physics, Department of physics and astronomy, Uppsala University, Laegerhyddsvaegen 1, 751 20 Uppsala (Sweden); Koning, Arjan; Rochman, Dimitri [Nuclear Research and consultancy Group (NRG) Westerduinweg 3, Petten (Netherlands); Bejmer, Klaes-Hakan [Vattenfall Nuclear Fuel AB, Jaemtlandsgatan 99, Vaellingby (Sweden); Henriksson, Hans [Vattenfall Research and Development AB, Jaemtlandsgatan 99, Vaellingby (Sweden)

    2010-07-01

    Today, nuclear data libraries do not handle uncertainties from nuclear data in a consistent manner and the reactor codes do not request uncertainties in nuclear data input. Thus, the output from these codes have unknown uncertainties. The plan is to use a method proposed by Koning and Rochman to investigate the propagation of nuclear data uncertainties into reactor physics codes and macroscopic parameters. A project (acronym MACRO) has started at Uppsala University in collaboration with A. Koning and with financial support from Vattenfall AB and the Swedish Research Council within the GENIUS (Generation IV research in universities of Sweden) project. In the proposed method the uncertainties in nuclear model parameters will be derived from theoretical considerations and comparisons of nuclear model results with experimental cross-section data. Given the probability distribution in the model parameters a large set of random, complete ENDF-formatted nuclear data libraries will be created using the TALYS code. The generated nuclear data libraries will then be used in neutron transport codes to obtain macroscopic reactor parameters. For this, models of reactor systems with proper geometry and elements will be used. This will be done for all data libraries and the variation of the final results will be regarded as a systematic uncertainty in the investigated reactor parameter. The understanding of these systematic uncertainties is especially important for the design and intercomparison of new reactor concepts, i.e., Generation IV, and optimization applications for current generation reactors is envisaged. (authors)

  19. Pressurized water reactor fuel rod design methodology

    International Nuclear Information System (INIS)

    Silva, A.T.; Esteves, A.M.

    1988-08-01

    The fuel performance program FRAPCON-1 and the structural finite element program SAP-IV are applied in a pressurized water reactor fuel rod design methodology. The applied calculation procedure allows to dimension the fuel rod components and characterize its internal pressure. (author) [pt

  20. C.E.C. - cod for calculus of the evolution fuel for thermal reactors

    International Nuclear Information System (INIS)

    Biciolla, L.; Marcu, G.; Mociornita, G.

    1975-01-01

    The study of ''burnup'' into thermal reactor involves two main aspects: the economic one and another regarding the reactor operation, its stability and control. In the CEC-code written in FORTRAN IV language was analysed the change of the isotopic composition of nuclear fuel from thermal reactor during its operation

  1. Nuclear reactors

    International Nuclear Information System (INIS)

    Prescott, R.F.

    1976-01-01

    A nuclear reactor containment vessel faced internally with a metal liner is provided with thermal insulation for the liner, comprising one or more layers of compressible material such as ceramic fiber, such as would be conventional in an advanced gas-cooled reactor and also a superposed layer of ceramic bricks or tiles in combination with retention means therefor, the retention means (comprising studs projecting from the liner, and bolts or nuts in threaded engagement with the studs) being themselves insulated from the vessel interior so that the coolant temperatures achieved in a High-Temperature Reactor or a Fast Reactor can be tolerated with the vessel. The layer(s) of compressible material is held under a degree of compression either by the ceramic bricks or tiles themselves or by cover plates held on the studs, in which case the bricks or tiles are preferably bedded on a yielding layer (for example of carbon fibers) rather than directly on the cover plates

  2. Nuclear reactor

    International Nuclear Information System (INIS)

    Miyashita, Akio.

    1981-01-01

    Purpose: To facilitate and accelerate a leakage test of valves of a main steam pipe by adding a leakage test partition valve thereto. Constitution: A leakage testing partition valve is provided between a pressure vessel for a nuclear reactor and the most upstream side valve of a plurality of valves to be tested for leakage, a testing branch pipe is communicated with the downstream side of the partition valve, and the testing water for preventing leakage is introduced thereto through the branch pipe. Since main steam pipe can be simply isolated by closing the partition valve in the leakage test, the leakage test can be conducted without raising or lowering the water level in the pressure vessel, and since interference with other work in the reactor can be eliminated, the leakage test can be readily conducted parallel with other work in the reactor in a short time. Clean water can be used without using reactor water as the test water. (Yoshihara, H.)

  3. Reactor container

    International Nuclear Information System (INIS)

    Abe, Yoshihito; Sano, Tamotsu; Ueda, Sabuo; Tanaka, Kazuhisa.

    1987-01-01

    Purpose: To improve the liquid surface disturbance in LMFBR type reactors. Constitution: A horizontal flow suppressing mechanism mainly comprising vertical members is suspended near the free liquid surface of coolants in the upper plenum. The horizontal flow of coolants near the free liquid surface is reduced by the suppressing mechanism to effectively reduce the surface disturbance. The reduction in the liquid surface disturbance further prevails to the entire surface region with no particular vertical variations to the free liquid surface to remarkably improve the preventive performance for the liquid surface disturbance. Accordingly, it is also possible to attain the advantageous effects such as prevention for the thermal fatigue in reactor vessel walls, reactor upper mechanisms, etc. and prevention of burning damage to the reactor core due to the reduction of envolved Ar gas. (Kamimura, M.)

  4. REACTOR SHIELD

    Science.gov (United States)

    Wigner, E.P.; Ohlinger, L.E.; Young, G.J.; Weinberg, A.M.

    1959-02-17

    Radiation shield construction is described for a nuclear reactor. The shield is comprised of a plurality of steel plates arranged in parallel spaced relationship within a peripheral shell. Reactor coolant inlet tubes extend at right angles through the plates and baffles are arranged between the plates at right angles thereto and extend between the tubes to create a series of zigzag channels between the plates for the circulation of coolant fluid through the shield. The shield may be divided into two main sections; an inner section adjacent the reactor container and an outer section spaced therefrom. Coolant through the first section may be circulated at a faster rate than coolant circulated through the outer section since the area closest to the reactor container is at a higher temperature and is more radioactive. The two sections may have separate cooling systems to prevent the coolant in the outer section from mixing with the more contaminated coolant in the inner section.

  5. NUCLEAR REACTOR

    Science.gov (United States)

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

    1958-01-21

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

  6. Breeder reactors

    International Nuclear Information System (INIS)

    Gollion, H.

    1977-01-01

    The reasons for the development of fast reactors are briefly reviewed (a propitious neutron balance oriented towards a maximum uranium burnup) and its special requirements (cooling, fissile material density and reprocessing) discussed. The three stages in the French program of fast reactor development are outlined with Rapsodie at Cadarache, Phenix at Marcoule, and Super Phenix at Creys-Malville. The more specific features of the program of research and development are emphasized: kinetics and the core, the fuel and the components [fr

  7. Nuclear reactor

    International Nuclear Information System (INIS)

    Schulze, I.; Gutscher, E.

    1980-01-01

    The core contains a critical mass of UN or U 2 N 3 in the form of a noncritical solution with melted Sn being kept below a N atmosphere. The lining of the reactor core consists of graphite. If fission progresses part of the melted metal solution is removed and cleaned from fission products. The reactor temperatures lie in the range of 300 to 2000 0 C. (Examples and tables). (RW) [de

  8. 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.)

  9. Reactor containment

    International Nuclear Information System (INIS)

    Kawabe, Ryuhei; Yamaki, Rika.

    1990-01-01

    A water vessel is disposed and the gas phase portion of the water vessel is connected to a reactor container by a pipeline having a valve disposed at the midway thereof. A pipe in communication with external air is extended upwardly from the liquid phase portion to a considerable height so as to resist against the back pressure by a waterhead in the pipeline. Accordingly, when the pressure in the container is reduced to a negative level, air passes through the pipeline and uprises through the liquid phase portion in the water vessel in the form of bubbles and then flows into the reactor container. When the pressure inside of the reactor goes higher, since the liquid surface in the water vessel is forced down, water is pushed up into the pipeline. Since the waterhead pressure of a column of water in the pipeline and the pressure of the reactor container are well-balanced, gases in the reactor container are not leaked to the outside. Further, in a case if a great positive pressure is formed in the reactor container, the inner pressure overcomes the waterhead of the column of water, so that the gases containing radioactive aerosol uprise in the pipeline. Since water and the gases flow being in contact with each other, this can provide the effect of removing aerosol. (T.M.)

  10. Fast reactors

    International Nuclear Information System (INIS)

    Vasile, A.

    2001-01-01

    Fast reactors have capacities to spare uranium natural resources by their breeding property and to propose solutions to the management of radioactive wastes by limiting the inventory of heavy nuclei. This article highlights the role that fast reactors could play for reducing the radiotoxicity of wastes. The conversion of 238 U into 239 Pu by neutron capture is more efficient in fast reactors than in light water reactors. In fast reactors multi-recycling of U + Pu leads to fissioning up to 95% of the initial fuel ( 238 U + 235 U). 2 strategies have been studied to burn actinides: - the multi-recycling of heavy nuclei is made inside the fuel element (homogeneous option); - the unique recycling is made in special irradiation targets placed inside the core or at its surroundings (heterogeneous option). Simulations have shown that, for the same amount of energy produced (400 TWhe), the mass of transuranium elements (Pu + Np + Am + Cm) sent to waste disposal is 60,9 Kg in the homogeneous option and 204.4 Kg in the heterogeneous option. Experimental programs are carried out in Phenix and BOR60 reactors in order to study the feasibility of such strategies. (A.C.)

  11. 11. IV avati Draakoni galeriis...

    Index Scriptorium Estoniae

    2005-01-01

    Tanel Saare (sünd. 1979) näitus "Gott und huhn episode IV: seed shower". Eksponeeritakse väljavõtteid aktsioonidest aastatel 2000-2004 Turus, Nürnbergis, Berliinis, Lohusalus ja Soulis. Osa aktsioone toimus koos rühmitusega Non Grata

  12. Review on Korea Participation of Generation IV International Forum (GIF)

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jewhan; Jeong, Ji-Young; Hahn, Dohee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    Generation IV International Forum (GIF) originates from US proposal of an initiative in 2000. The vision was to leapfrog LWR technology and collaborate with international partners to share R and D on advanced nuclear systems. Nine countries and EU joined the initiative and Gen IV concept was defined via technology goals and legal framework. Two years study with more than 100 experts worldwide has evaluated nearly 100 reactor designs and down selected six most promising concepts. In 2005, the first signatures on Framework Agreement were collected and the first research projects were defined in 2006. Korea is one of the founding members of GIF and actively participating in various areas. In 2013, TD was assigned to Korean expert and Korea is endeavoring to enhance the benefit of participation since this turning point. In this paper, pros and cons of engaging with GIF were briefly introduced and items to maximize the benefit were suggested.

  13. Review on Korea Participation of Generation IV International Forum (GIF)

    International Nuclear Information System (INIS)

    Lee, Jewhan; Jeong, Ji-Young; Hahn, Dohee

    2015-01-01

    Generation IV International Forum (GIF) originates from US proposal of an initiative in 2000. The vision was to leapfrog LWR technology and collaborate with international partners to share R and D on advanced nuclear systems. Nine countries and EU joined the initiative and Gen IV concept was defined via technology goals and legal framework. Two years study with more than 100 experts worldwide has evaluated nearly 100 reactor designs and down selected six most promising concepts. In 2005, the first signatures on Framework Agreement were collected and the first research projects were defined in 2006. Korea is one of the founding members of GIF and actively participating in various areas. In 2013, TD was assigned to Korean expert and Korea is endeavoring to enhance the benefit of participation since this turning point. In this paper, pros and cons of engaging with GIF were briefly introduced and items to maximize the benefit were suggested

  14. Annual report 1990 annex IV

    International Nuclear Information System (INIS)

    Smith, B.; Foskolos, K.

    1991-01-01

    The Nuclear Energy Research Department consists of three Laboratories: Reactor Physics and Systems Engineering (LRS), Thermal Hydraulics (LTH), and Materials and Nuclear Processes (LWV). There are additionally two Research Programmes: 'Waste Disposal' (PES) and LWR-Safety (PLS). The Department's activities focus on three main areas: safety and safety-related operational problems of the Swiss nuclear power plants (NPPs), contributions to the issues involved in radioactive waste disposal, future reactors. In addition, scientific-technical services are provided in the nuclear and non-nuclear sectors and for the operation of the relevant site facilities. (author) figs., tabs., refs

  15. Research reactors - an overview

    International Nuclear Information System (INIS)

    West, C.D.

    1997-01-01

    A broad overview of different types of research and type reactors is provided in this paper. Reactor designs and operating conditions are briefly described for four reactors. The reactor types described include swimming pool reactors, the High Flux Isotope Reactor, the Mark I TRIGA reactor, and the Advanced Neutron Source reactor. Emphasis in the descriptions is placed on safety-related features of the reactors. 7 refs., 7 figs., 2 tabs

  16. Generation IV international forum 2002 - remarks

    International Nuclear Information System (INIS)

    Abraham, S.

    2002-01-01

    Analyses and forecasts underscore the important role of nuclear power in energy supply in the 21st century. Important aspects in this respect are the conservation of fossil resources, the protection of the world's climate, and the continuity of supply. Present 1st and 2nd generation nuclear power plants ensure an economical and technically mature electricity supply. Advanced systems offering, e.g., higher efficiency of fuel utilization, simplified systems technology, and advanced safety characteristics, can make available additional benefits in using nuclear power. Upon an initiative of the U.S. Department of Energy (DOE), ten countries combine their efforts in developing such reactor concepts in the Generation IV International Forum (GIF). Argentina, Brazil, Canada, France, Japan, South Africa, South Korea, Switzerland, the United Kingdom, and the United States pursue the common objective in GIF to identify suitable nuclear power systems and promote their development up to the envisaged readiness for construction in 2030. Besides technical and economic questions of nuclear power generation, also other aspects must be considered with a view to the future use of nuclear power. The particularly relevant issues, such as the management of radioactive waste, the intensification of research and development, and international cooperation, have been taken up by the Bush administration at an early point in time and have been, or will be, incorporated in practical solutions, as in the case of the Yucca Mountain repository project. (orig.)

  17. The concepts of liquid metal of IV generation

    International Nuclear Information System (INIS)

    Carbonnier, J. L.

    2005-01-01

    The concepts of liquid metals, due to their large spectrum, show important possibility of sustainable development: two concepts of liquid metal (Sodium and Lead) were engaged in the frame of the IV generation. The reactors with sodium benefit from considerable background of experience and of important work on projects to aim at the price diminution and the increase of safety (EFR, JSFR). The commitment of Japan as a leader of this concept and the support by France allow to contemplate an industrial deployment from 2015. The lead reactors offer some advantages in the domain of safety but otherwise require a highly important research and development binded to the control of the corrosion, the perspective of deployment of this concept are more hypothetical

  18. Gen IV International Forum - GIF, 2010 Annual Report

    International Nuclear Information System (INIS)

    Anon.

    2011-01-01

    The Generation IV International Forum (GIF), created in 2000 to foster international collaboration at a detailed level of actual R and D, is a cooperative international endeavor, organized to develop the research necessary to test the feasibility and performance capabilities of fourth generation nuclear systems, with the goal of making such systems deployable in large numbers around 2030. Since its beginning, GIF members stated the following goals for the fourth generation of nuclear power plants when compared to previous generations: a) improve sustainability (including effective fuel utilization and minimization of waste); b) improve economics (competitiveness with respect to other energy sources); c) improve safety and reliability (e.g. no need for offsite emergency response); and d) improve proliferation resistance and physical protection. After an in-depth analysis of the different available concepts, whatever their level of development, the Forum selected six concepts as the most promising, and decided to focus R and D on these systems: - the very-high-temperature reactor (VHTR); - the sodium-cooled fast reactor (SFR); - the supercritical-water-cooled reactor (SCWR); - the gas-cooled fast reactor (GFR); - the lead-cooled fast reactor (LFR); - the molten salt reactor (MSR). Active members of the GIF are Canada, Euratom, France, Japan, People's Republic of China, Republic of Korea, Republic of South Africa, Russian Federation, Switzerland and the United States. Altogether, they represent around 90% of the world installed nuclear capacity for producing electricity, and all key technology holders. The forum is led by the policy group, where all members are represented, and currently chaired by Japan since December 2009, assisted by vice-chairs from France and United States. The year 2010 has seen some important achievements and decisions regarding these six systems. For example, two sodium-cooled fast reactors (re)started this year: Monju in Japan restarted after

  19. AMZ, multigroup constant library for EXPANDA code, generated by NJOY code from ENDF/B-IV

    International Nuclear Information System (INIS)

    Chalhoub, E.S.; Moraes, Marisa de

    1985-01-01

    It is described a library of multigroup constants with 70 energy groups and 37 isotopes to fast reactor calculation. The cross sections, scattering matrices and self-shielding factors were generated by NJOY code and RGENDF interface program, from ENDF/B-IV'S evaluated data. The library is edited in adequated format to be used by EXPANDA code. (M.C.K.) [pt

  20. Uruguay; 2011 Article IV Consultation

    OpenAIRE

    International Monetary Fund

    2011-01-01

    This 2011 Article IV Consultation highlights that the growth momentum in Uruguay has continued into 2011 but a slowdown is under way, led by weaker exports and slower public investment. Uruguay’s economic and financial vulnerabilities are modest, and the government has reduced debt vulnerabilities significantly and built important financial buffers. Executive Directors have commended authorities’ skillful macroeconomic management that has underpinned Uruguay’s excellent economic performance, ...

  1. Austria; 2013 Article IV Consultation

    OpenAIRE

    International Monetary Fund

    2013-01-01

    This paper presents details of Austria’s 2013 Article IV Consultation. Austria has been growing economically but is facing challenges in the financial sector. Full implementation of medium-term fiscal adjustment plans require specifying several measures and plans that need gradual strengthening to take expected further bank restructuring cost into account. It suggests that strong early bank intervention and resolution tools, a better designed deposit insurance system, and a bank-financed reso...

  2. Nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Prescott, R F; George, B V; Baglin, C J

    1978-05-10

    Reference is made to thermal insulation on the inner surfaces of containment vessels of fluid cooled nuclear reactors and particularly in situations where the thermal insulation must also serve a structural function and transmit substantial load forces to the surface which it covers. An arrangement is described that meets this requirement and also provides for core support means that favourably influences the flow of hot coolant from the lower end of the core into a plenum space in the hearth of the reactor. The arrangement comprises a course of thermally insulating bricks arranged as a mosaic covering a wall of the reactor and a course of thermally insulating tiles arranged as a mosaic covering the course of bricks. Full constructional details are given.

  3. Nuclear reactors

    International Nuclear Information System (INIS)

    Prescott, R.F.; George, B.V.; Baglin, C.J.

    1978-01-01

    Reference is made to thermal insulation on the inner surfaces of containment vessels of fluid cooled nuclear reactors and particularly in situations where the thermal insulation must also serve a structural function and transmit substantial load forces to the surface which it covers. An arrangement is described that meets this requirement and also provides for core support means that favourably influences the flow of hot coolant from the lower end of the core into a plenum space in the hearth of the reactor. The arrangement comprises a course of thermally insulating bricks arranged as a mosaic covering a wall of the reactor and a course of thermally insulating tiles arranged as a mosaic covering the course of bricks. Full constructional details are given. (UK)

  4. Bioconversion reactor

    Science.gov (United States)

    McCarty, Perry L.; Bachmann, Andre

    1992-01-01

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

  5. A recent source modification for noble gases at the Los Alamos on-line mass analysis facility

    International Nuclear Information System (INIS)

    Balestrini, S.J.; Forman, L.

    1976-01-01

    The Los Alamos on-line mass analysis experiment at the Godiva-IV burst reactor facility has been modified to determine independent fission yields of noble gases. The gases are released from a stearate target and ionization by electron bombardment. The distance traveled by the gases from the target to the ionization chamber is 20 cm. The efficiency of the electron bombardment source is lower than that of the surface ionization source that was employed to measure the yields of Rb and Cs. But this effect is compensated by the larger quantity of target metal that is possible when using a stearate target. (Auth.)

  6. Development of high temperature gas cooled reactor in China

    Energy Technology Data Exchange (ETDEWEB)

    Guo, Wentao [Paul Scherrer Institute, Villigen (Switzerland). Dept. of Nuclear Energy and Safety; Schorer, Michael [Swiss Nuclear Forum, Olten (Switzerland)

    2018-02-15

    High temperature gas cooled reactor (HTGR) is one of the six Generation IV reactor types put forward by Generation IV International Forum (GIF) in 2002. This type of reactor has high outlet temperature. It uses Helium as coolant and graphite as moderator. Pebble fuel and ceramic reactor core are adopted. Inherit safety, good economy, high generating efficiency are the advantages of HTGR. According to the comprehensive evaluation from the international nuclear community, HTGR has already been given the priority to the research and development for commercial use. A demonstration project of the High Temperature Reactor-Pebble-�bed Modules (HTR-PM) in Shidao Bay nuclear power plant in China is under construction. In this paper, the development history of HTGR in China and the current situation of HTR-PM will be introduced. The experiences from China may be taken as a reference by the international nuclear community.

  7. Nuclear reactor

    International Nuclear Information System (INIS)

    Scholz, M.

    1976-01-01

    An improvement of the accessibility of that part of a nuclear reactor serving for biological shield is proposed. It is intended to provide within the biological shield, distributed around the circumference of the reactor pressure vessel, several shielding chambers filled with shielding material, which are isolated gastight from the outside by means of glass panes with a given bursting strength. It is advantageous that, on the one hand, inspection and maintenance will be possible without great effort and, on the other, a large relief cross section will be at desposal if required. (UWI) [de

  8. NEUTRONIC REACTOR

    Science.gov (United States)

    Wigner, E.P.; Weinberg, A.W.; Young, G.J.

    1958-04-15

    A nuclear reactor which uses uranium in the form of elongated tubes as fuel elements and liquid as a coolant is described. Elongated tubular uranium bodies are vertically disposed in an efficient neutron slowing agent, such as graphite, for example, to form a lattice structure which is disposed between upper and lower coolant tanks. Fluid coolant tubes extend through the uranium bodies and communicate with the upper and lower tanks and serve to convey the coolant through the uranium body. The reactor is also provided with means for circulating the cooling fluid through the coolant tanks and coolant tubes, suitable neutron and gnmma ray shields, and control means.

  9. The status of proliferation resistance evaluation methodology development in GEN IV international forum

    International Nuclear Information System (INIS)

    Inoue, Naoko; Kawakubo, Yoko; Seya, Michio; Suzuki, Mitsutoshi; Kuno, Yusuke; Senzaki, Masao

    2010-01-01

    The Generation IV Nuclear Energy Systems International Forum (GIF) Proliferation Resistance and Physical Protection Working Group (PR and PP WG) was established in December 2002 in order to develop the PR and PP evaluation methodology for GEN IV nuclear energy systems. The methodology has been studied and established by international consensus. The PR and PP WG activities include development of the measures and metrics; establishment of the framework of PR and PP evaluation, the demonstration study using Example Sodium Fast Reactor (ESFR), which included the development of three evaluation approaches; the Case Study using ESFR and four kinds of threat scenarios; the joint study with GIF System Steering Committees (SSCs) of the six reactor design concepts; and the harmonization study with the IAEA's International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO). This paper reviews the status of GIF PR and PP studies and identifies the challenges and directions for applying the methodology to evaluate future nuclear energy systems in Japan. (author)

  10. Generation IV concepts - Presentation at ACRS workshop 'Regulatory challenges for future nuclear power plants'

    International Nuclear Information System (INIS)

    Versluis, Rob M.

    2001-01-01

    The concept of the Near-Term Deployment Working Group was to define a technical approach for Generation IV system with enough detail to allow evaluation against the goal, bur broad enough to allow for optional features and trade. The following concepts were taken into account: water coolant (water or heavy water), gas coolant, liquid metal coolants. Concepts were grouped according to concept sets of technology base share and design approach. Water coolant concepts were grouped as follows: PWR loop reactors, integral primary system PWRs, Integral BWRs, pressure tube reactors, high conversion cores, supercritical water reactors, advanced fuel cycle concepts. Gas coolant concepts were grouped as follows: pebble bed modular reactors; prismatic modular reactors, very high temperature reactors, fast spectrum reactors, others (fluidized bed, moving ignition zone concept). Liquid metal concepts were grouped in four major categories: Medium-to-large oxide-fueled systems; Medium-sized metal-fueled systems; Medium-sized Pb/Pb-Bi systems; Small-sized Pb/Pb-Bi systems. The three supporting technology areas were examined: Fuels (oxide, metal, nitride); Coolants (Na, Pb/Pb-Bi); Fuel Cycle (advanced aqueous, pyroprocess). Non-classical concepts were also grouped as follows: Eutectic metallic fuel; Molten salt fuel; Gas core reactor; Molten salt cooled/solid fuel; Organic cooled reactor; Solid conduction/heat pipe; Fission product direct energy conversion. The Technical working Groups are analyzing the candidate concepts for performance potential relative to the goals; and technology gaps

  11. Neutronic reactor

    International Nuclear Information System (INIS)

    Wende, C.W.J.

    1976-01-01

    The method of operating a water-cooled neutronic reactor having a graphite moderator is described which comprises flowing a gaseous mixture of carbon dioxide and helium, in which the helium comprises 40--60 volume percent of the mixture, in contact with the graphite moderator. 2 claims, 4 figures

  12. Neutronic reactor

    International Nuclear Information System (INIS)

    Wende, C.W.J.

    1976-01-01

    A safety rod for a nuclear reactor has an inner end portion having a gamma absorption coefficient and neutron capture cross section approximately equal to those of the adjacent shield, a central portion containing materials of high neutron capture cross section and an outer end portion having a gamma absorption coefficient at least equal to that of the adjacent shield

  13. Reactor facility

    International Nuclear Information System (INIS)

    Suzuki, Hiroaki; Murase, Michio; Yokomizo, Osamu.

    1997-01-01

    The present invention provides a BWR type reactor facility capable of suppressing the amount of steams generated by the mutual effect of a failed reactor core and coolants upon occurrence of an imaginal accident, and not requiring spacial countermeasures for enhancing the pressure resistance of the container vessel. Namely, a means for supplying cooling water at a temperature not lower by 30degC than the saturated temperature corresponding to the inner pressure of the containing vessel upon occurrence of an accident is disposed to a lower dry well below the pressure vessel. As a result, upon occurrence of such an accident that the reactor core should be melted and flown downward of the pressure vessel, when cooling water at a temperature not lower than the saturated temperature, for example, cooling water at 100degC or higher is supplied to the lower dry well, abrupt generation of steams by the mutual effect of the failed reactor core and cooling water is scarcely caused compared with a case of supplying cooling water at a temperature lower than the saturation temperature by 30degC or more. Accordingly, the amount of steams to be generated can be suppressed, and special countermeasure is no more necessary for enhancing the pressure resistance of the container vessel is no more necessary. (I.S.)

  14. Nuclear reactor

    International Nuclear Information System (INIS)

    Gilroy, J.E.

    1980-01-01

    An improved cover structure for liquid metal cooled fast breeder type reactors is described which it is claimed reduces the temperature differential across the intermediate grid plate of the core cover structure and thereby reduces its subjection to thermal stresses. (UK)

  15. Reactor licensing

    International Nuclear Information System (INIS)

    Harvie, J.D.

    2002-01-01

    This presentation discusses reactor licensing and includes the legislative basis for licensing, other relevant legislation , the purpose of the Nuclear Safety and Control Act, important regulations, regulatory document, policies, and standards. It also discusses the role of the CNSC, its mandate and safety philosophy

  16. Nuclear reactor

    International Nuclear Information System (INIS)

    Hattori, Sadao; Sekine, Katsuhisa.

    1987-01-01

    Purpose: To decrease the thickness of a reactor container and reduce the height and the height and plate thickness of a roof slab without using mechanical vibration stoppers. Constitution: Earthquake proofness is improved by filling fluids such as liquid metal between a reactor container and a secondary container and connecting the outer surface of the reactor container with the inner surface of the secondary container by means of bellows. That is, for the horizontal seismic vibrations, horizontal loads can be supported by the secondary container without providing mechanical vibration stoppers to the reactor container and the wall thickness can be reduced thereby enabling to simplify thermal insulation structure for the reduction of thermal stresses. Further, for the vertical seismic vibrations, verical loads can be transmitted to the secondary container thereby enabling to reduce the wall thickness in the same manner as for the horizontal load. By the effect of transferring the point of action of the container load applied to the roof slab to the outer circumferential portion, the intended purpose can be attained and, in addition, the radiation dose rate at the upper surface of the roof slab can be decreased. (Kamimura, M.)

  17. Reactor system

    International Nuclear Information System (INIS)

    Miyano, Hiroshi; Narabayashi, Naoshi.

    1990-01-01

    The represent invention concerns a reactor system with improved water injection means to a pressure vessel of a BWR type reactor. A steam pump is connected to a heat removing system pipeline, a high pressure water injection system pipeline and a low pressure water injection system pipeline for injecting water into the pressure vessel. A pump actuation pipeline is disposed being branched from a main steam pump or a steam relieaf pipeline system, through which steams are supplied to actuate the steam pump and supply cooling water into the pressure vessel thereby cooling the reactor core. The steam pump converts the heat energy into the kinetic energy and elevates the pressure of water to a level higher than the pressure of the steams supplied by way of a pressure-elevating diffuser. Cooling water can be supplied to the pressure vessel by the pressure elevation. This can surely inject cooling water into the pressure vessel upon loss of coolant accident or in a case if reactor scram is necessary, without using an additional power source. (I.N.)

  18. Reactor core

    International Nuclear Information System (INIS)

    Matsuura, Tetsuaki; Nomura, Teiji; Tokunaga, Kensuke; Okuda, Shin-ichi

    1990-01-01

    Fuel assemblies in the portions where the gradient of fast neutron fluxes between two opposing faces of a channel box is great are kept loaded at the outermost peripheral position of the reactor core also in the second operation cycle in the order to prevent interference between a control rod and the channel box due to bending deformation of the channel box. Further, the fuel assemblies in the second row from the outer most periphery in the first operation cycle are also kept loaded at the second row in the second operation cycle. Since the gradient of the fast neutrons in the reactor core is especially great at the outer circumference of the reactor core, the channel box at the outer circumference is bent such that the surface facing to the center of the reactor core is convexed and the channel box in the second row is also bent to the identical direction, the insertion of the control rod is not interfered. Further, if the positions for the fuels at the outermost periphery and the fuels in the second row are not altered in the second operation cycle, the gaps are not reduced to prevent the interference between the control rod and the channel box. (N.H.)

  19. Analysis and evaluation of the Dual Fluid Reactor concept

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Xiang

    2017-06-27

    The Dual Fluid Reactor is a molten salt fast reactor developed by IFK in Berlin based on the Gen-IV Molten-Salt Reactor concept and the Liquid-Metal Cooled Reactor. The design aims to combine these two concepts to improve these two concepts. The Dissertation focuses on the concept and performs diverse calculations and estimations on the subjects of neutron physics, depletion and thermal-hydraulic behaviors to validate the new features of the concept. Based on the results it is concluded that this concept is feasible to its desired purpose and with great potential.

  20. New about research reactors

    International Nuclear Information System (INIS)

    Egorenkov, P.M.

    2001-01-01

    The multi-purpose research reactor MAPLE (Canada) and concept of new reactor MAPLE-CNF as will substitute the known Canadian research reactor NRU are described. New reactor will be used as contributor for investigations into materials, neutron beams and further developments for the CANDU type reactor. The Budapest research reactor (BRR) and its application after the last reconstruction are considered also [ru

  1. Crystalline cerium(IV) phosphates

    International Nuclear Information System (INIS)

    Herman, R.G.; Clearfield, A.

    1976-01-01

    The ion exchange behaviour of seven crystalline cerium(IV) phosphates towards some of the alkali metal cations is described. Only two of the compounds (A and C) possess ion exchange properties in acidic solutions. Four others show some ion exchange characteristics in basic media with some of the alkali cations. Compound G does not behave as an ion exchanger in solutions of pH + , but show very little Na + uptake. Compound E undergoes ion exchange with Na + and Cs + , but not with Li+. Both Li + and Na + are sorbed by compounds A and C. The results are indicative of structures which show steric exclusion phenomena. (author)

  2. PREPARATION OF OXOPORPHINATOMANGANESE (IV) COMPLEX

    Energy Technology Data Exchange (ETDEWEB)

    Willner, I.; Otvos, J.; Calvin, M.

    1980-07-01

    Oxo-manganese-tetraphenylporphyrin (O=Mn{sup IV}-TPP) has been prepared by an oxygen-transfer reaction from iodosylbenzene to MnIITPP and characterized by its i.r. and field desorption mass spectra, which are identical to those of the product obtained by direct oxidation of Mn{sup III}(TPP) in an aqueous medium; it transfers oxygen to triphenylphosphine to produce triphenylphosphine oxide, and it is suggested that similar intermediates are important in oxygen activation by cytochrome P-450 as well as in the photosynthetic evolution of oxygen.

  3. Reactor core of nuclear reactor

    International Nuclear Information System (INIS)

    Sasagawa, Masaru; Masuda, Hiroyuki; Mogi, Toshihiko; Kanazawa, Nobuhiro.

    1994-01-01

    In a reactor core, a fuel inventory at an outer peripheral region is made smaller than that at a central region. Fuel assemblies comprising a small number of large-diameter fuel rods are used at the central region and fuel assemblies comprising a great number of smalldiameter fuel rods are used at the outer peripheral region. Since a burning degradation rate of the fuels at the outer peripheral region can be increased, the burning degradation rate at the infinite multiplication factor of fuels at the outer region can substantially be made identical with that of the fuels in the inner region. As a result, the power distribution in the direction of the reactor core can be flattened throughout the entire period of the burning cycle. Further, it is also possible to make the degradation rate of fuels at the outer region substantially identical with that of fuels at the inner side. A power peak formed at the outer circumferential portion of the reactor core of advanced burning can be lowered to improve the fuel integrity, and also improve the reactor safety and operation efficiency. (N.H.)

  4. Reactor core and passive safety systems descriptions of a next generation pressure tube reactor - mechanical aspects

    Energy Technology Data Exchange (ETDEWEB)

    Yetisir, M.; Gaudet, M.; Rhodes, D.; Hamilton, H.; Pencer, J. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

    2014-07-01

    Canada has been developing a channel-type supercritical water-cooled nuclear reactor concept, often called the Canadian SCWR. The objective of this reactor concept is to meet the technology goals of the Generation IV International Forum (GIF) for the next generation nuclear reactor development, which include enhanced safety features (inherent safe operation and deploying passive safety features), improved resource utilization, sustainable fuel cycle, and greater proliferation resistance than Generation III nuclear reactors. The Canadian SCWR core concept consists of a high-pressure inlet plenum, a separate low-pressure heavy water moderator contained in a calandria vessel, and 336 pressure tubes surrounded by the moderator. The reactor uses supercritical water as a coolant, and a direct steam power cycle to generate electricity. The reactor concept incorporates advanced safety features such as passive core cooling, long-term decay heat rejection to the environment and fuel melt prevention via passive moderator cooling. These features significantly reduce core damage frequency relative to existing nuclear reactors. This paper presents a description of the design concepts for the Canadian SCWR core, reactor building layout and the plant layout. Passive safety concepts are also described that address containment and core cooling following a loss-of coolant accident, as well as long term reactor heat removal at station blackout conditions. (author)

  5. Test Review: Advanced Clinical Solutions for WAIS-IV and WMS-IV

    Science.gov (United States)

    Chu, Yiting; Lai, Mark H. C.; Xu, Yining; Zhou, Yuanyuan

    2012-01-01

    The authors review the "Advanced Clinical Solutions for WAIS-IV and WMS-IV". The "Advanced Clinical Solutions (ACS) for the Wechsler Adult Intelligence Scale-Fourth Edition" (WAIS-IV; Wechsler, 2008) and the "Wechsler Memory Scale-Fourth Edition" (WMS-IV; Wechsler, 2009) was published by Pearson in 2009. It is a…

  6. Progress in the development of the neutron flux monitoring system of the French GEN-IV SFR: simulations and experimental validations [ANIMMA--2015-IO-98

    Energy Technology Data Exchange (ETDEWEB)

    Jammes, C.; Filliatre, P.; De Izarra, G. [CEA, DEN, DER, Instrumentation, Sensors and Dosimetry Laboratory, Cadarache, F-13108 Saint-Paul-lez-Durance, (France); Elter, Zs.; Pazsit, I. [Chalmers University of Technology, Department of Applied Physics, Division of Nuclear Engineering, SE-412 96 Goteborg, (Sweden); Verma, V.; Hellesen, C.; Jacobsson, S. [Division of Applied Nuclear Physics, Uppsala University, Box 516, SE-75120 Uppsala, (Sweden); Hamrita, H.; Bakkali, M. [CEA, DRT, LIST, Sensors and Electronic Architecture Laboratory, Saclay, F-91191 Gif Sur Yvette, (France); Chapoutier, N.; Scholer, A-C.; Verrier, D. [AREVA NP, 10 rue Juliette Recamier F-69456 Lyon, (France); Cantonnet, B.; Nappe, J-C. [PHONIS France S.A.S, Nuclear Instrumentation, Avenue Roger Roncier, B.P. 520, F-19106 Brive Cedex, (France); Molinie, P.; Dessante, P.; Hanna, R.; Kirkpatrick, M.; Odic, E. [Supelec, Department of Power and Energy System, F-91192 Gif Sur Yvette, (France); Jadot, F. [CEA, DEN, DER, ASTRID Project Group, Cadarache, F-13108 Saint-Paul-lez-Durance, (France)

    2015-07-01

    The neutron flux monitoring system of the French GEN-IV sodium-cooled fast reactor will rely on high temperature fission chambers installed in the reactor vessel and capable of operating over a wide-range neutron flux. The definition of such a system is presented and the technological solutions are justified with the use of simulation and experimental results. (authors)

  7. Nuclear reactor

    International Nuclear Information System (INIS)

    Gibbons, J.F.; McLaughlin, D.J.

    1978-01-01

    In the pressure vessel of the water-cooled nuclear reactor there is provided an internal flange on which the one- or two-part core barrel is hanging by means of an external flange. A cylinder is extending from the reactor vessel closure downwards to a seat on the core cupport structure and serves as compression element for the transmission of the clamping load from the closure head to the core barrel (upper guide structure). With the core barrel, subject to tensile stress, between the vessel internal flange and its seat on one hand and the compression of the cylinder resp. hold-down element between the closure head and the seat on the other a very strong, elastic sprung structure is obtained. (DG) [de

  8. Nuclear reactor

    International Nuclear Information System (INIS)

    Sasaki, Tomozo.

    1987-01-01

    Purpose: To improve the nuclear reactor availability by enabling to continuously exchange fuels in the natural-slightly enriched uranium region during operation. Constitution: A control rod is withdrawn to the midway of a highly enriched uranium region by means of control rod drives and the highly enriched uranium region is burnt to maintain the nuclear reactor always at a critical state. At the same time, fresh uranium-slightly enriched uranium is continuously supplied gravitationally from a fresh fuel reservoir through fuel reservoir to each of fuel pipes in the natural-slightly enriched uranium region. Then, spent fuels reduced with the reactivity by the burn up are successively taken out from the bottom of each of the fuel pipes through an exit duct and a solenoid valve to the inside of a spent fuel reservoir and the burn up in the natural-slightly enriched uranium region is conducted continuously. (Kawakami, Y.)

  9. Nuclear reactor

    International Nuclear Information System (INIS)

    Sakurai, Mikio; Yamauchi, Koki.

    1983-01-01

    Purpose: To improve the channel stability and the reactor core stability in a spontaneous circulation state of coolants. Constitution: A reactor core stabilizing device comprising a differential pressure automatic ON-OFF valve is disposed between each of a plurality of jet pumps arranged on a pump deck. The stabilizing device comprises a piston exerted with a pressure on the lower side of the pump deck by way of a pipeway and a valve for flowing coolants through the bypass opening disposed to the pump deck by the opening and closure of the valve ON-OFF. In a case where the jet pumps are stopped, since the differential pressure between the upper and the lower sides of the pump deck is removed, the valve lowers gravitationally into an opened state, whereby the coolants flow through the bypass opening to increase the spontaneous circulation amount thereby improve the stability. (Yoshino, Y.)

  10. Nuclear reactor

    International Nuclear Information System (INIS)

    Aleite, W.; Bock, H.W.; Struensee, S.

    1976-01-01

    The invention concerns the use of burnable poisons in a nuclear reactor, especially in PWRs, in order to improve the controllability of the reactor. An unsymmetrical arrangement in the lattice is provided, if necessary also by insertion of special rods for these additions. It is proposed to arrange the burnable poisons in fuel elements taken over from a previous burn-up cycle and to distribute them, going out from the side facing the control rods, over not more than 20% of the lenth of the fuel elements. It seems sufficient, for the burnable poisons to bind an initial reactivity of only 0.1% and to become ineffective after normal operation of 3 to 4 months. (ORU) [de

  11. Reactor container

    International Nuclear Information System (INIS)

    Ichiki, Tadaharu; Saba, Kazuhisa.

    1979-01-01

    Purpose: To improve the earthquake resistance as well as reduce the size of a container for a nuclear reactor with no adverse effects on the decrease of impact shock to the container and shortening of construction step. Constitution: Reinforcing profile steel materials are welded longitudinally and transversely to the inner surface of a container, and inner steel plates are secured to the above profile steel materials while keeping a gap between the materials and the container. Reactor shielding wall planted to the base concrete of the container is mounted to the pressure vessel, and main steam pipeways secured by the transverse beams and led to the outside of container is connected. This can improve the rigidity earthquake strength and the safetiness against the increase in the inside pressure upon failures of the container. (Yoshino, Y.)

  12. Reactor container

    International Nuclear Information System (INIS)

    Oyamada, Osamu; Furukawa, Hideyasu; Uozumi, Hiroto.

    1979-01-01

    Purpose: To lower the position of an intermediate slab within a reactor container and fitting a heat insulating material to the inner wall of said intermediate slab, whereby a space for a control rod exchanging device and thermal stresses of the inner peripheral wall are lowered. Constitution: In the pedestal at the lower part of a reactor pressure vessel there is formed an intermediate slab at a position lower than diaphragm floor slab of the outer periphery of the pedestal thereby to secure a space for providing automatic exchanging device of a control rod driving device. Futhermore, a heat insulating material is fitted to the inner peripheral wall at the upper side of the intermediate slab part, and the temperature gradient in the wall thickness direction at the time of a piping rupture trouble is made gentle, and thermal stresses at the inner peripheral wall are lowered. (Sekiya, K.)

  13. Neutronic reactor

    International Nuclear Information System (INIS)

    Lewis, W.R.

    1978-01-01

    Disclosed is a graphite-moderated, water-cooled nuclear reactor including a plurality of rectangular graphite blocks stacked in abutting relationship in layers, alternate layers having axes which are normal to one another, alternate rows of blocks in alternate layers being provided with a channel extending through the blocks, said channeled blocks being provided with concave sides and having smaller vertical dimensions than adjacent blocks in the same layer, there being nuclear fuel in the channels

  14. Nuclear reactors

    International Nuclear Information System (INIS)

    Humphreys, P.; Davidson, D.F.; Thatcher, G.

    1980-01-01

    The cooling system of a liquid metal cooled fast breeder nuclear reactor of the pool kind is described. It has an intermediate heat exchange module comprising a tube-in-shell heat exchanger and an electromagnetic flow coupler in the base region of the module. Primary coolant is flowed through the heat exchanger being driven by electromagnetic interaction with secondary liquid metal coolant flow effected by a mechanical pump. (author)

  15. Nuclear reactor

    International Nuclear Information System (INIS)

    Jungmann, A.

    1975-01-01

    Between a PWR's reactor pressure vessel made of steel and the biological shield made of concrete there is a gap. This gap is filled up with a heat insulation facting the reactor pressure vessel, for example with insulating concrete segments jacketed with sheet steel and with an additional layer. This layer serves for smooth absorption of compressive forces originating in radial direction from the reactor pressure vessel. It consists of cylinder-segment shaped bricks made of on situ concrete, for instance. The bricks have cooling agent ports in one or several rows which run parallel to the wall of the pressure vessel and in alignment with superposed bricks. Between the layer of bricks and the biological shield or rather the heat insulation, there are joints which are filled, however, with injected mortar. That guarantees a smooth series of connected components resistant tom compression. Besides, a slip foil can be set between the heat insulation and the joining joint filled with mortar for the reduction of the friction at thermal expansions. (TK) [de

  16. Reactor building

    International Nuclear Information System (INIS)

    Ebata, Sakae.

    1990-01-01

    At least one valve rack is disposed in a reactor building, on which pipeways to a main closure valve, valves and bypasses of turbines are placed and contained. The valve rack is fixed to the main body of the building or to a base mat. Since the reactor building is designed as class A earthquake-proofness and for maintaining the S 1 function, the valve rack can be fixed to the building main body or to the base mat. With such a constitution, the portions for maintaining the S 1 function are concentrated to the reactor building. As a result, the dispersion of structures of earthquake-proof portion corresponding to the reference earthquake vibration S 1 can be prevented. Accordingly, the conditions for the earthquake-proof design of the turbine building and the turbine/electric generator supporting rack are defined as only the class B earthquake-proof design conditions. In view of the above, the amount of building materials can be saved and the time for construction can be shortened. (I.S.)

  17. Nuclear reactors

    International Nuclear Information System (INIS)

    Yoshioka, Michiko.

    1985-01-01

    Purpose: To obtain an optimum structural arrangement of IRM having a satisfactory responsibility to the inoperable state of a nuclear reactor and capable of detecting the reactor power in an averaged manner. Constitution: As the structural arrangement of IRM, from 6 to 16 even number of IRM are bisected into equial number so as to belong two trip systems respectively, in which all of the detectors are arranged at an equal pitch along a circumference of a circle with a radius rl having the center at the position of the central control rod in one trip system, while one detector is disposed near the central control rod and other detectors are arranged substantially at an equal pitch along the circumference of a circle with a radius r2 having the center at the position for the central control rod in another trip system. Furthermore, the radius r1 and r2 are set such that r1 = 0.3 R, r2 = 0.5 R in the case where there are 6 IRM and r1 = 0.4 R and R2 = 0.8 R where there are eight IRM where R represents the radius of the reactor core. (Kawakami, Y.)

  18. MLR reactor

    International Nuclear Information System (INIS)

    Ryazantsev, E.P.; Egorenkov, P.M.; Nasonov, V.A.; Smimov, A.M.; Taliev, A.V.; Gromov, B.F.; Kousin, V.V.; Lantsov, M.N.; Radchenko, V.P.; Sharapov, V.N.

    1998-01-01

    The Material Testing Loop Reactor (MLR) development was commenced in 1991 with the aim of updating and widening Russia's experimental base to validate the selected directions of further progress of the nuclear power industry in Russia and to enhance its reliability and safety. The MLR reactor is the pool-type one. As coolant it applies light water and as side reflector beryllium. The direction of water circulation in the core is upward. The core comprises 30 FA arranged as hexagonal lattice with the 90-95 mm pitch. The central materials channel and six loop channels are sited in the core. The reflector includes up to 11 loop channels. The reactor power is 100 MW. The average power density of the core is 0.4 MW/I (maximal value 1.0 MW/l). The maximum neutron flux density is 7.10 14 n/cm 2 s in the core (E>0.1 MeV), and 5.10 14 n/cm 2 s in the reflector (E<0.625 eV). In 1995 due to the lack of funding the MLR designing was suspended. (author)

  19. Nuclear reactor

    International Nuclear Information System (INIS)

    Shirakawa, Toshihisa.

    1979-01-01

    Purpose: To prevent cladding tube injuries due to thermal expansion of each of the pellets by successively extracting each of the control rods loaded in the reactor core from those having less number of notches, as well as facilitate the handling work for the control rods. Constitution: A recycle flow control device is provided to a circulation pump for forcibly circulating coolants in the reactor container and an operational device is provided for receiving each of the signals concerning number of notches for each of the control rods and flow control depending on the xenon poisoning effect obtained from the signals derived from the in-core instrument system connected to the reactor core. The operational device is connected with a control rod drive for moving each of the control rods up and down and a recycle flow control device. The operational device is set with a pattern for the aimed control rod power and the sequence of extraction. Upon extraction of the control rods, they are extracted successively from those having less notch numbers. (Moriyama, K.)

  20. Reactor container

    International Nuclear Information System (INIS)

    Hidaka, Masataka; Hatamiya, Shigeo; Kawasaki, Terufumi; Fukui, Toru; Suzuki, Hiroaki; Kataoka, Yoshiyuki; Kawabe, Ryuhei; Murase, Michio; Naito, Masanori.

    1990-01-01

    In order to suppress the pressure elevation in a reactor container due to high temperature and high pressure steams jetted out upon pipeway rupture accidents in the reactor container, the steams are introduced to a pressure suppression chamber for condensating them in stored coolants. However, the ability for suppressing the pressure elevation and steam coagulation are deteriorated due to the presence of inactive incondensible gases. Then, there are disposed a vent channel for introducing the steams in a dry well to a pressure suppression chamber in the reactor pressure vessel, a closed space disposed at the position lower than a usual liquid level, a first channel having an inlet in the pressure suppression chamber and an exit in the closed space and a second means connected by way of a backflow checking means for preventing the flow directing to the closed space. The first paths are present by plurality, a portion of which constitutes a syphon. The incondensible gases and the steams are discharged to the dry well at high pressure by using the difference of the water head for a long cooling time after the pipeway rupture accident. Then, safety can be improved without using dynamic equipments as driving source. (N.H.)

  1. Reactor core in FBR type reactor

    International Nuclear Information System (INIS)

    Masumi, Ryoji; Kawashima, Katsuyuki; Kurihara, Kunitoshi.

    1989-01-01

    In a reactor core in FBR type reactors, a portion of homogenous fuels constituting the homogenous reactor core is replaced with multi-region fuels in which the enrichment degree of fissile materials is lower nearer to the axial center. This enables to condition the composition such that a reactor core having neutron flux distribution either of a homogenous reactor core or a heterogenous reactor core has substantially identical reactivity. Accordingly, in the transfer from the homogenous reactor core to the axially heterogenous reactor core, the average reactivity in the reactor core is substantially equal in each of the cycles. Further, by replacing a portion of the homogenous fuels with a multi-region fuels, thereby increasing the heat generation near the axial center, it is possiable to reduce the linear power output in the regions above and below thereof and, in addition, to improve the thermal margin in the reactor core. (T.M.)

  2. Biofilm Community Dynamics in Bench-Scale Annular Reactors Simulating Arrestment of Chloraminated Drinking Water Nitrification

    Science.gov (United States)

    Annular reactors (ARs) were used to study biofilm community succession and provide an ecological insight during nitrification arrestment through simultaneously increasing monochloramine (NH2Cl) and chlorine to nitrogen mass ratios, resulting in four operational periods (I to IV)....

  3. Electro-regeneration of Ce(IV) in real spent Cr-etching solutions

    International Nuclear Information System (INIS)

    Chen, Te-San; Huang, Kuo-Lin

    2013-01-01

    Highlights: • An electrochemical process is used to regenerate Ce(IV) in real (hazardous) spent TFT-LCD Cr-etching solutions. • The Ce(IV) yield on tested anodes was in order BDD > Pt > DSA. • A Neosepta CMX separator was better than Nafion ones to be used in the process. • The activation energy on Pt was 10.7 kJ/mol. • The obtained parameters are useful to design reactors for 100% Ce(IV) regeneration in real spent Cr-etching solutions. -- Abstract: This paper presents the electro-regeneration of Ce(IV) in real (hazardous) spent thin-film transistor liquid-crystal display (TFT-LCD) Cr-etching solutions. In addition to Ce(III) > Ce(IV) in diffusivity, a quasi-reversible behavior of Ce(III)/Ce(IV) was observed at both boron-doped diamond (BDD) and Pt disk electrodes. The Ce(IV) yield on Pt increased with increasing current density, and the best current efficiency (CE) was obtained at 2 A/2.25 cm 2 . The performance in terms of Ce(IV) yield and CE of tested anodes was in order BDD > Pt > dimensional stable anode (DSA). At 2 A/2.25 cm 2 on Pt and 40 °C for 90 min, the Ce(IV) yield, CE and apparent rate constant (k) for Ce(III) oxidation were 81.4%, 21.8% and 3.17 × 10 −4 s −1 , respectively. With the increase of temperature, the Ce(IV) yield, CE, and k increased (activation energy = 10.7 kJ/mol), but the specific electricity consumption decreased. The Neosepta CMX membrane was more suitable than Nafion-117 and Nafion-212 to be used as the separator of the Ce(IV) regeneration process. The obtained parameters are useful to design divided batch reactors for the Ce(IV) electro-regeneration in real spent Cr-etching solutions

  4. Trends on development for reactor engineering

    International Nuclear Information System (INIS)

    Schulenberg, T.; Starflinger, J.

    2004-01-01

    There are basically two ways of improving the performance of a nuclear power plant. First, the conversion rate of thermal into mechanical energy can be increased, similar to conventional steam power plant technology, whose experience can be integrated. Secondly, the burnup rate can be optimized, i.e. the conversion of fuel into thermal energy. The worldwide resarch project GENERATION IV follows both strategies. This contribution presents two typical examples: A LWR reactor in which the efficiency was raised to more than 44 percent by supercritical steam states, and a lead-cooled fast breeder reactor in which burnup was increased to more than 150 GWd/t. (orig.)

  5. Reevaluation of an individual's radiation exposure at NTS in 1963-64

    International Nuclear Information System (INIS)

    Myers, D.S.

    1983-01-01

    The FRAN prompt burst reactor began operation at NTS on November 1, 1962 and continued in use until April 1965. From January 2, 1963 to August 12, 1964, an individual periodically performed maintenance and troubleshooting functions on various components of the FRAN reactor system. In June, 1980, the individual requested a review of the radiation dose that he received from his involvement with the FRAN reactor. An evaluation of the individual's radiation dose associated with the FRAN reactor operation was performed. This report details the reevaluation of the individual's estimated radiation dose from the FRAN reactor assembly, as derived from computer calculations, GODIVA-IV measurements, personnel dosimetry results, and a reconstruction of work scenarios

  6. Environmental sensitivity studies for Gen-IV roadmap DUPIC scenario

    International Nuclear Information System (INIS)

    Jeong, Chang Joon

    2004-03-01

    The environmental effect of the DUPIC (Direct Use of Spent PWR Fuel in CANDU Reactors) fuel cycle, which is considered as one of the partial recycle scenario in Gen-IV roadmap, has been analyzed by using the dynamic analysis method. Through the parametric calculations for the DUPIC fuel cycle deployment time and the fraction of the DUPIC reactors, the environmental effects of the fuel cycle for important parameters such as the amount of spent fuel and the combined amounts of plutonium and minor actinides were estimated and compared to those of the once-through LWR fuel cycle. The results of the sensitivity calculations showed that an early deployment of the DUPIC fuel cycle with a high DUPIC reactor fraction can reduce the accumulation of spent fuel by up to 40%. More important is the associated reduction in the combined amount of plutonium and minor actinides, which may reduce the key repository parameter (long term decay heat). Therefore it is expected that favorable environmental effects will be the outcome of the implementation of the DUPIC fuel cycle

  7. Transmutation of americium in critical reactors

    International Nuclear Information System (INIS)

    Wallenius, J.

    2005-01-01

    Already in 1974, a Los Alamos report suggested that the recycling of higher actinides would be detrimental for the safety of critical reactors. Later investigations confirmed this understanding, and stringent limits on the fraction of minor actinides allowed to be present in the fuel of fast neutron reactors were established. In recent years, and in particular in connection with the generation IV initiative, it has been advocated that recycling of americium in critical reactors is not only feasible, but also a recommendable approach. In the present contribution, it is shown, to the contrary, that introduction of americium into reactors with uranium based fuels deteriorates the safety margin of these reactors to a degree that will not allow consumption of the americium sources present in any economically competitive nuclear fuel cycle. Further, it is shown that uranium and thorium free cores with plutonium based fuels may be designed, that features excellent safety characteristics, as long as americium is not present in the feed. Hence, a closed fuel cycle is suggested, that consists of commercial power production in light water reactors, plutonium burning in uranium and thorium free fast neutron critical reactors, and higher actinide consumption in accelerator driven systems with inert matrix fuel. It is argued that such a fuel cycle (being a refinement of the Double Strata fuel cycle proposed by JAERI and further developed by M. Salvatores) provides a minimum cost penalty for implementing P and T under realistic boundary conditions. (author)

  8. Some oxozirconium(IV) compounds

    Energy Technology Data Exchange (ETDEWEB)

    Paul, R C; Gupta, S K; Parmar, S S; Vasisht, S K [Punjab Univ., Chandigarh (India). Dept. of Chemistry

    1976-01-01

    Some new oxozirconium(IV) complexes, ZrO(An)/sub 2/, ZrO(Gly)/sub 2/, ZrO(HSal)/sub 2/, ZrO(HPth)/sub 2/, ZrO(Pic)/sub 2/(HPic)/sub 2/, and ZrO(Quin)/sub 2/(HQuin)/sub 2/ have been isolated from the reactions of ZrO(CH/sub 3/COO)/sub 2/CH/sub 3/COOH with anthranilic acid (HAn), glycine (HGly), salicylic acid (H/sub 2/Sal), phthalic acid (H/sub 2/Pth), picolinic acid (HPic), and 8-quinolinol (HQuin) respectively. Their important infrared bands and wherever possible molar conductance and molecular weight have been reported.

  9. Reactor Simulations for Safeguards with the MCNP Utility for Reactor Evolution Code

    International Nuclear Information System (INIS)

    Shiba, T.; Fallot, M.

    2015-01-01

    To tackle nuclear material proliferation, we conducted several proliferation scenarios using the MURE (MCNP Utility for Reactor Evolution) code. The MURE code, developed by CNRS laboratories, is a precision, open-source code written in C++ that automates the preparation and computation of successive MCNP (Monte Carlo N-Particle) calculations and solves the Bateman equations in between, for burnup or thermal-hydraulics purposes. In addition, MURE has been completed recently with a module for the CHaracterization of Radioactive Sources, called CHARS, which computes the emitted gamma, beta and alpha rays associated to any fuel composition. Reactor simulations could allow knowing how plutonium or other material generation evolves inside reactors in terms of time and amount. The MURE code is appropriate for this purpose and can also provide knowledge on associated particle emissions. Using MURE, we have both developed a cell simulation of a typical CANDU reactor and a detailed model of light water PWR core, which could be used to analyze the composition of fuel assemblies as a function of time or burnup. MURE is also able to provide, thanks to its extension MURE-CHARTS, the emitted gamma rays from fuel assemblies unloaded from the core at any burnup. Diversion cases of Generation IV reactors have been also developed; a design of Very High Temperature Reactor (a Pebble Bed Reactor (PBR), loaded with UOx, PuOx and ThUOx fuels), and a Na-cooled Fast Breeder Reactor (FBR) (with depleted Uranium or Minor Actinides in the blanket). The loading of Protected Plutonium Production (P3) in the FBR was simulated. The simulations of various reactor designs taking into account reactor physics constraints may bring valuable information to inspectors. At this symposium, we propose to show the results of these reactor simulations as examples of the potentiality of reactor simulations for safeguards. (author)

  10. Molten salt reactors: reactor cores

    International Nuclear Information System (INIS)

    1983-01-01

    In this critical analysis of the MSBR I project are examined the problems concerning the reactor core. Advantages of breeding depend essentially upon solutions to technological problems like continuous reprocessing or graphite behavior under neutron irradiation. Graphite deformation, moderator unloading, control rods and core instrumentation require more studies. Neutronics of the core, influence of core geometry and salt composition, fuel evolution, and thermohydraulics are reviewed [fr

  11. Increased SRP reactor power

    International Nuclear Information System (INIS)

    MacAfee, I.M.

    1983-01-01

    Major changes in the current reactor hydraulic systems could be made to achieve a total of about 1500 MW increase of reactor power for P, K, and C reactors. The changes would be to install new, larger heat exchangers in the reactor buildings to increase heat transfer area about 24%, to increase H 2 O flow about 30% per reactor, to increase D 2 O flow 15 to 18% per reactor, and increase reactor blanket gas pressure from 5 psig to 10 psig. The increased reactor power is possible because of reduced inlet temperature of reactor coolant, increased heat removal capacity, and increased operating pressure (larger margin from boiling). The 23% reactor power increase, after adjustment for increased off-line time for reactor reloading, will provide a 15% increase of production from P, K, and C reactors. Restart of L Reactor would increase SRP production 33%

  12. Nuclear research reactors

    International Nuclear Information System (INIS)

    1985-01-01

    It's presented data about nuclear research reactors in the world, retrieved from the Sien (Nuclear and Energetic Information System) data bank. The information are organized in table forms as follows: research reactors by countries; research reactors by type; research reactors by fuel and research reactors by purpose. (E.G.) [pt

  13. Nuclear reactor physics course for reactor operators

    International Nuclear Information System (INIS)

    Baeten, P.

    2006-01-01

    The education and training of nuclear reactor operators is important to guarantee the safe operation of present and future nuclear reactors. Therefore, a course on basic 'Nuclear reactor physics' in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The aim of the basic course on 'Nuclear Reactor Physics for reactor operators' is to provide the reactor operators with a basic understanding of the main concepts relevant to nuclear reactors. Seen the education level of the participants, mathematical derivations are simplified and reduced to a minimum, but not completely eliminated

  14. Nuclear reactor

    International Nuclear Information System (INIS)

    Jolly, R.

    1979-01-01

    The support grid for the fuel rods of a liquid metal cooled fast breeder reactor has a regular hexagonal contour and contains a large number of unit cells arranged honeycomb fashion. The totality of these cells make up a hexagonal shape. The grid contains a number of strips of material, and there is a window in each of three sidewalls staggered by one sidewall. The other sidewalls have embossed protrusions, thus generating a guide lining or guide bead. The windows reduce the rigidity of the areas in the middle between the ends of the cells. (DG) [de

  15. Nuclear reactor

    International Nuclear Information System (INIS)

    Anthony, A.J.; Gruber, E.A.

    1979-01-01

    A nuclear reactor with control rods in channels between fuel assemblies wherein the fuel assemblies incorporate guide rods which protrude outwardly into the control rod channels to prevent the control rods from engaging the fuel elements. The guide rods also extend back into the fuel assembly such that they are relatively rigid members. The guide rods are tied to the fuel assembly end or support plates and serve as structural members which are supported independently of the fuel element. Fuel element spacing and support means may be attached to the guide rods. 9 claims

  16. Genetics Home Reference: glycogen storage disease type IV

    Science.gov (United States)

    ... Home Health Conditions Glycogen storage disease type IV Glycogen storage disease type IV Printable PDF Open All ... Javascript to view the expand/collapse boxes. Description Glycogen storage disease type IV (GSD IV) is an ...

  17. A cerium(IV)-carbon multiple bond

    Energy Technology Data Exchange (ETDEWEB)

    Gregson, Matthew; Lu, Erli; McMaster, Jonathan; Lewis, William; Blake, Alexander J.; Liddle, Stephen T. [Nottingham Univ. (United Kingdom). School of Chemistry

    2013-12-02

    Straightforward access to a cerium(IV)-carbene complex was provided by one-electron oxidation of an anionic ''ate'' cerium(III)-carbene precursor, thereby avoiding decomposition reactions that plague oxidations of neutral cerium(III) compounds. The cerium(IV)-carbene complex is the first lanthanide(IV)-element multiple bond and involves a twofold bonding interaction of two electron pairs between cerium and carbon. [German] Auf direktem Wege zu einem Cer(IV)-Carbenkomplex gelangt man durch die Einelektronenoxidation einer anionischen Carben-Cerat(III)-Vorstufe. So werden Zersetzungsprozesse vermieden, die die Oxidation neutraler Cer(III)-Verbindungen erschweren. Der Cer(IV)-Carbenkomplex enthaelt die erste Lanthanoid(IV)-Element-Mehrfachbindung; dabei binden Cer und Kohlenstoff ueber zwei Elektronenpaare.

  18. iväkoti Riemula

    OpenAIRE

    Alanko, Reetta; Ihanamäki, Katja

    2012-01-01

    Opinnäytetyössä kuvataan yleisesti päivähoidon kehitystä Suomessa sekä päivähoitoa yrittäjän näkökulmasta, tuoden esille sen tämän päivän haasteet ja mahdollisuudet. Työssä on pohdittu yhteistyön merkitystä kunnan kanssa ja sitä, miten kunta voi osaltaan joko rajoittaa tai edesauttaa yksityisen päivähoitoyrityksen toimintaa. Opinnäytetyössä kerrotaan teoriassa Päiväkoti Riemula nimisen, erityispäivähoitopalveluita tarjoavan yrityksen perustamiseen liittyvistä suunnitelmista. Suunnitelluss...

  19. Molt salts reactors capacity for wastes incineration and energy production

    International Nuclear Information System (INIS)

    David, S.; Nuttin, A.

    2005-01-01

    The molten salt reactors present many advantages in the framework of the IV generation systems development for the energy production and/or the wastes incineration. After a recall of the main studies realized on the molten salt reactors, this document presents the new concepts and the identified research axis: the MSRE project and experience, the incinerators concepts, the thorium cycle. (A.L.B.)

  20. Nuclear reactors

    International Nuclear Information System (INIS)

    Prescott, R.F.; George, B.V.; Baglin, C.J.

    1979-01-01

    In a nuclear reactor (e.g. one having coolant down-flow through a core to a hearth below) thermal insulation (e.g. of a floor of the hearth) comprises a layer of bricks and a layer of tiles thereon, with smaller clearances between the tiles than between the bricks but with the bricks being of reduced cross-section immediately adjacent the tiles so as to be surrounded by interconnected passages, of relatively large dimensions, constituting a continuous chamber extending behind the layer of tiles. By this arrangement, lateral coolant flow in the inter-brick clearances is much reduced. The reactor core is preferably formed of hexagonal columns, supported on diamond-shaped plates each supported on a pillar resting on one of the hearth-floor tiles. Each plate has an internal duct, four upper channels connecting the duct with coolant ducts in four core columns supported by the plate, and lower channels connecting the duct to a downwardly-open recess common to three plates, grouped to form a hexagon, at their mutually-adjacent corners. This provides mixing, and temperature-averaging, of coolant from twelve columns

  1. Reactor container

    International Nuclear Information System (INIS)

    Oikawa, Hirohide; Otonari, Jun-ichiro; Tozaki, Yuka.

    1993-01-01

    Partition walls are disposed between a reactor pressure vessel and a suppression chamber to separate a dry well to an upper portion and a lower portion. A communication pipe is disposed to the partition walls. One end of the communication pipe is opened in an upper portion of the dry well at a position higher than a hole disposed to a bent tube of the suppression chamber. When coolants overflow from a depressurization valve by an erroneous operation of an emergency reactor core cooling device, the coolants accumulate in the upper portion of the dry well. When the pipeline is ruptured at the upper portion of the pressure vessel, only the inside of the pressure vessel and the upper portion of the dry well are submerged in water. In this case, the water level of the coolants does not elevate to the opening of the commuication pipe but they flow into the suppression chamber from the hole disposed to the bent tube. Since the coolants do not flow out to the lower portion of the dry well, important equipments such as control rod drives disposed at the lower portion of the dry wall can be prevented from submerging in water. (I.N.)

  2. Reactor monitor

    International Nuclear Information System (INIS)

    Takada, Tamotsu.

    1992-01-01

    The device of the present invention monitors a reactor so that each of the operations for the relocation of fuel assemblies and the withdrawal and the insertion of control rods upon exchange of fuel assemblies and control rods in the reactor. That is, when an operator conducts relocating operation by way of a fuel assembly operation section, the device of the present invention judges whether the operation indication is adequate or not, based on the information of control rod arrangement in a control rod memory section. When the operation indication is wrong, a stop signal is sent to a fuel assembly relocating device. Further, when the operator conducts control rod operation by way of a control rod operation section, the device of the present invention judges in the control rod withdrawal judging section, as to whether the operation indication given by the operator is adequate or not by comparing it with fuel assembly arrangement information. When the operation indication is wrong, a stop signal is sent to control rod drives. With such procedures, increase of nuclear heating upon occurrence of erroneous operation can be prevented. (I.S.)

  3. Nuclear reactors

    International Nuclear Information System (INIS)

    Matheson, J.E.

    1983-01-01

    A nuclear reactor has an upper and a lower grid plate. Protrusions project from the upper grid plate. Fuel assemblies having end fittings fit between the grid plates. An arrangement is provided for accepting axial forces generated during the operation of the nuclear reactor by the flow of the cooling medium and thermal expansion and irradiation-induced growth of the fuel assembly, which comprises rods. Each fuel assembly rests on the lower grid plate and its upper end is elastically supported against the upper grid plate by the above-mentioned arrangement. The arrangement comprises four (for example) torsion springs each having a torsion tube and a torsion bar nested within the torsion tube and connected at one end thereto. The other end of the torsion bar is connected to an associated one of four lever arms. The torsion tube is rigidly connected to the other end fitting and the springs are disposed such that the lever arms are biassed against the protrusions. (author)

  4. ENDF/B-IV fission-product files: summary of major nuclide data

    International Nuclear Information System (INIS)

    England, T.R.; Schenter, R.E.

    1975-09-01

    The major fission-product parameters [sigma/sub th/, RI, tau/sub 1/2/, E-bar/sub β/, E-bar/sub γ/, E-bar/sub α/, decay and (n,γ) branching, Q, and AWR] abstracted from ENDF/B-IV files for 824 nuclides are summarized. These data are most often requested by users concerned with reactor design, reactor safety, dose, and other sundry studies. The few known file errors are corrected to date. Tabular data are listed by increasing mass number

  5. Reactor core fuel management

    International Nuclear Information System (INIS)

    Silvennoinen, P.

    1976-01-01

    The subject is covered in chapters, entitled: concepts of reactor physics; neutron diffusion; core heat transfer; reactivity; reactor operation; variables of core management; computer code modules; alternative reactor concepts; methods of optimization; general system aspects. (U.K.)

  6. Hybrid adsorptive membrane reactor

    Science.gov (United States)

    Tsotsis, Theodore T [Huntington Beach, CA; Sahimi, Muhammad [Altadena, CA; Fayyaz-Najafi, Babak [Richmond, CA; Harale, Aadesh [Los Angeles, CA; Park, Byoung-Gi [Yeosu, KR; Liu, Paul K. T. [Lafayette Hill, PA

    2011-03-01

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

  7. Reactor outage schedule (tentative)

    Energy Technology Data Exchange (ETDEWEB)

    Walton, R.P.

    1969-11-01

    This single page document is the November 1, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the production reactor.

  8. Reactor outage schedule (tentative)

    Energy Technology Data Exchange (ETDEWEB)

    Walton, R.P.

    1969-10-01

    This single page document is the October 1, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the Production Reactor.

  9. Reactor outage schedule (tentative)

    Energy Technology Data Exchange (ETDEWEB)

    Walton, R.P.

    1969-10-15

    This single page document is the October 15, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the Production Reactor.

  10. Reactor outage schedule (tentative)

    Energy Technology Data Exchange (ETDEWEB)

    Walton, R.P.

    1969-09-15

    This single page document is the September 15, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the Production Reactor.

  11. Reactor outage schedule (tentative)

    Energy Technology Data Exchange (ETDEWEB)

    Walton, R.P.

    1969-12-15

    This single page document is the December 16, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the Production reactor.

  12. Reactor outage schedule (tentative)

    Energy Technology Data Exchange (ETDEWEB)

    Walton, R.P.

    1969-12-01

    This single page document is the December 1, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the Production reactor.

  13. Reactor theory and power reactors. 1. Calculational methods for reactors. 2. Reactor kinetics

    International Nuclear Information System (INIS)

    Henry, A.F.

    1980-01-01

    Various methods for calculation of neutron flux in power reactors are discussed. Some mathematical models used to describe transients in nuclear reactors and techniques for the reactor kinetics' relevant equations solution are also presented

  14. Experimental facilities for gas-cooled reactor safety studies. Task group on Advanced Reactor Experimental Facilities (TAREF)

    International Nuclear Information System (INIS)

    2009-01-01

    In 2007, the NEA Committee on the Safety of Nuclear Installations (CSNI) completed a study on Nuclear Safety Research in OECD Countries: Support Facilities for Existing and Advanced Reactors (SFEAR) which focused on facilities suitable for current and advanced water reactor systems. In a subsequent collective opinion on the subject, the CSNI recommended to conduct a similar exercise for Generation IV reactor designs, aiming to develop a strategy for ' better preparing the CSNI to play a role in the planned extension of safety research beyond the needs set by current operating reactors'. In that context, the CSNI established the Task Group on Advanced Reactor Experimental Facilities (TAREF) in 2008 with the objective of providing an overview of facilities suitable for performing safety research relevant to gas-cooled reactors and sodium fast reactors. This report addresses gas-cooled reactors; a similar report covering sodium fast reactors is under preparation. The findings of the TAREF are expected to trigger internationally funded CSNI projects on relevant safety issues at the key facilities identified. Such CSNI-sponsored projects constitute a means for efficiently obtaining the necessary data through internationally co-ordinated research. This report provides an overview of experimental facilities that can be used to carry out nuclear safety research for gas-cooled reactors and identifies priorities for organizing international co-operative programmes at selected facilities. The information has been collected and analysed by a Task Group on Advanced Reactor Experimental Facilities (TAREF) as part of an ongoing initiative of the NEA Committee on the Safety of Nuclear Installations (CSNI) which aims to define and to implement a strategy for the efficient utilisation of facilities and resources for Generation IV reactor systems. (author)

  15. Discussion on safety analysis approach for sodium fast reactors

    International Nuclear Information System (INIS)

    Hong, Soon Joon; Choo, Yeon Joon; Suh, Nam Duk; Shin, Ahn Dong; Bae, Moo Hoon

    2012-01-01

    Utilization of nuclear energy is increasingly necessary not only because of the increasing energy consumption but also because of the controls on greenhouse emissions against global warming. To keep step with such demands, advanced reactors are now world widely under development with the aims of highly economical advances, and enhanced safety. Recently, further elaborating is encouraged on the research and development program for Generation IV (GEN IV) reactors, and in collaboration with other interested countries through the Generation IV International Forum (GIF). Sodium cooled Fast Reactor (SFR) is a strong contender amongst the GEN IV reactor concepts. Korea also takes part in that program and plans to construct demonstration reactor of SFR. SFR is under the development for a candidate of small modular reactors, for example, PRISM (Power Reactor Innovative Small Module). Understanding of safety analysis approach has also advanced by the demand of increasing comprehensive safety requirement. Reviewing the past development of the licensing and safety basis in the advanced reactors, such approaches seemed primarily not so satisfactory because the reference framework of licensing and safety analysis approach in the advanced reactors was always the one in water reactors. And, the framework is very plant specific one and thereby the advanced reactors and their frameworks don't look like a well assorted couple. Recently as a result of considerable advances in probabilistic safety assessment (PSA), risk informed approaches are increasingly applied together with some of the deterministic approaches like as the ones in water reactors. Technology neutral framework (TNF) can be said to be the utmost works of such risk informed approaches, even though an intensive assessment of the applicability has not been sufficiently accomplished. This study discusses the viable safety analysis approaches for the urgent application to the construction of pool type SFR. As discussed in

  16. The CEA research reactors

    International Nuclear Information System (INIS)

    Schwartz, J.P.

    1993-01-01

    Two main research reactors, specifically designed, PEGASE reactor and Laue-Langevin high flux reactor, are presented. The PEGASE reactor was designed at the end of the 50s for the study of the gas cooled reactor fuel element behaviour under irradiation; the HFR reactor, was designed in the late 60s to serve as a high yield and high level neutron source. Historical backgrounds, core and fuel characteristics and design, flux characteristics, etc., are presented. 5 figs

  17. 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.

  18. Extended analysis of Cu IV

    International Nuclear Information System (INIS)

    Meinders, E.; Uijlings, P.

    1980-01-01

    Wavelength data and classifications of 974 Cu IV lines in the region 750-1275 Angstroem are presented. Most of the lines have been classified as transitions from the previously unknown high even configurations 3d 7 5s and 3d 7 4d to 3d 7 4p. The configuration 3d 7 4d is seriously perturbed by 3d 6 4s 2 . The analysis resulted in the identification of 27 levels of 3d 7 5s and 113 levels of (3d 7 4d + 3d 6 4s 2 ) which are reported. The earlier published levels of 3d 7 4s and 3d 7 4p have to be shifted downward as a consequence of improved wavelength data. Radial paramter values, resulting from least-squares fits, are compared to Hartree-Fock values. The eigenvectors obtained in the parametric fitting are used to calculate transition probabilities in intermediate coupling. The relation between the observed intensities of the transitions 3d 7 4d-3d 7 4p and 3d 7 Ss-3d 7 4p is compared to the relation between theoretical values of the transition integrals obtained from Hartree-Fock calculations. A spectroscopic value for the ionization potentials is calculated from the 3d 7 ns configurations. (orig.)

  19. Studies of binary cerium(IV)-praseodymium(IV) and cerium(IV)-terbium(IV) oxides as pigments for ceramic applications

    International Nuclear Information System (INIS)

    Furtado, L.M.L.

    1991-01-01

    It was investigated a series of pigments of general composition Ce 1-x Pr x O 2 , and Ce x Tb y O 2 , exhibiting radish and brown colors, respectively, and high temperature stability. The pigments were obtained by dissolving appropriate amounts of the pure lanthanide oxides in acids and precipitating the rare earths as mixed oxalates, which were isolated and calcined under air, at 1000 0 C. X-Ray powder diffractograms were consistent with a cubic structure for the pigments. Magnetic susceptibility measurements, using Gouy method, indicated the presence of Pr(IV) ions in the Ce 1-x Pr x O 2 pigments and of Terbium predominantly as Tb(III) ions in the Ce-tb mixed oxides. A new method, based on suspension of solid samples in PVA-STB gels (STB = sodium tetradecaborate), was employed for the measurements of the electronic spectra of the pigments. The thermal behaviour the pigments was investigated by the calcination of the oxalates in the temperature range of 500 to 1200 O C, from 10 to 60 minutes. (author)

  20. To question of NPP power reactor choice for Kazakhstan

    International Nuclear Information System (INIS)

    Batyrbekov, G.A.; Makhanov, Y.M.; Reznikova, R.A.; Sidorenco, A.V.

    2004-01-01

    Full text: The requirements to NPP power reactors that will be under construction in Kazakhstan are proved and given in the report. A comparative analysis of the most advanced projects of power reactors with light and heavy water under pressure of large, medium and low power is carried out. Different reactors have been considered as follows: 1. Reactors with high-power (700 MW(el) and up) such as EPR, French - German reactor; CANDU-9, Canadian heavy-water reactor; System 80+, developed by ABB Combustion Engineering company, USA; KNGR, Korean reactor of the next generation; APWR, Japanese advanced reactor; WWER-1000 (V-392) - development of Atomenergoproect /Gydropress, Russian Federation; EP 1000, European passive reactor. 2. Reactors with medium power (300 MW (el) - 700 MW (el): AP-600, passive PWR of the Westinghouse company; CANDU-6, Canadian heavy-water reactor; AC-600, Chinese passive PWR; WWER-640, Russian passive reactor; MS-600 Japanese reactor of Mitsubishi Company; KSNP-600, South Korean reactor. 3. Reactors with low power (a few MW(el)- 300 MW(el)): IRIS, reactor of IV generation, developed by the International Corporation of 13 organizations from 7 countries, SMART, South Korean integrated reactor; CAREM, Argentina integrated reactor; MRX, Japanese integrated reactor; 'UNITERM', Russian NPP with integrated reactor, development of NIKIET; AHEC-80, Russian NPP, developed by OKBM. A comparison of the projects of the above-mentioned power reactors was carried out with respect to 15 criteria of nuclear, radiating, ecological safety and economic competitiveness, developed especially for this case. Data on a condition and prospects of power production and power consumption, stations and networks in Kazakhstan necessary for the choice of projects of NPP reactors for Kazakhstan are given. According to the data a balance of power production and power consumption as a whole in the country was received at the level of 59 milliard kw/h. However, strong dis balance

  1. To question of NPP power reactor choice for Kazakhstan

    International Nuclear Information System (INIS)

    Batyrbekov, G.A.; Makhanov, Y.M.; Reznikova, R.A.; Sidorenco, A.V.

    2004-01-01

    The requirements to NPP power reactors that will be under construction in Kazakhstan are proved and given in the report. A comparative analysis of the most advanced projects of power reactors with light and heavy water under pressure of large, medium and low power is carried out. Different reactors have been considered as follows: 1. Reactors with high-power (700 MW(el) and up) such as EPR, French - German reactor; CANDU-9, Canadian heavy-water reactor; System 80+, developed by ABB Combustion Engineering company, USA; KNGR, Korean reactor of the next generation; APWR, Japanese advanced reactor; WWER-1000 (V-392) - development of Atomenergoproect /Gydropress, Russian Federation; EP 1000, European passive reactor. 2. Reactors with medium power (300 MW (el) - 700 MW (el): AP-600, passive PWR of the Westinghouse company; CANDU-6, Canadian heavy-water reactor; AC-600, Chinese passive PWR; WWER-640, Russian passive reactor; MS-600 Japanese reactor of Mitsubishi Company; KSNP-600, South Korean reactor. 3. Reactors with low power (a few MW(el)- 300 MW(el)): IRIS, reactor of IV generation, developed by the International Corporation of 13 organizations from 7 countries, SMART, South Korean integrated reactor; CAREM, Argentina integrated reactor; MRX, Japanese integrated reactor; 'UNITERM', Russian NPP with integrated reactor, development of NIKIET; AHEC-80, Russian NPP, developed by OKBM. A comparison of the projects of the above-mentioned power reactors was carried out with respect to 15 criteria of nuclear, radiating, ecological safety and economic competitiveness, developed especially for this case. Data on a condition and prospects of power production and power consumption, stations and networks in Kazakhstan necessary for the choice of projects of NPP reactors for Kazakhstan are given. According to the data a balance of power production and power consumption as a whole in the country was received at the level of 59 milliard kw/h. However, strong dis balance in the

  2. Nuclear reactors safety issues

    International Nuclear Information System (INIS)

    Barre, Francois; Seiler, Nathalie

    2008-01-01

    fuels as well as the applied methodologies. The IRSN proceeds in a relevant and independent assessment of the submitted safety reports. To achieve this goal and maintain over time an independent and relevant assessment capability, the IRSN relies on the excellence of its experts and on state of art techniques and knowledge. The IRSN contributes by its work in key area to cutting edge research and development in order to drive nuclear industry towards making the best use of scientific and technological progress for improving safety, environmental protection and health. To maintain at all times the state of the art knowledge and the operational expertise necessary to deal efficiently with major nuclear accident consequences, the IRSN carries out, on the one hand, its own research and development programs to gain accurate knowledge on still unknown phenomena for safety analysis. On the other hand, the IRSN works out its own scientific calculation methodologies involving industrial calculation chain. Concerning more particularly the 'two-phase flows' thematic, The ISRN must correctly simulate the primary fluid behavior in the reactor in normal operation as well as in accidental situations, to estimate if, in such situations, the core reactor state is fully safe and any safety risk is undergone The research and development programs launched at the ISRN on two-phase flows gather work on advanced thermohydraulic configurations encounter in various reactor states (normal operation or accidental situations), in particular: (i)The estimation of the margin to the critical heat flux in normal operation (DNBR), (ii) The pressurized thermal shock, which is due to mechanical important constraints in the reactor vessel resulting from the injection of a cold fluid in case of emergency cooling (PTS), (iii) The reactivity insertion accident (RIA), (iv) The loss of coolant accident (LOCA), (vi) The accidents in spent-fuel pools and (vii) The severe accident, which could lead to core

  3. Nuclear reactor types

    International Nuclear Information System (INIS)

    Jones, P.M.S.

    1987-01-01

    The characteristics of different reactor types designed to exploit controlled fission reactions are explained. Reactors vary from low power research devices to high power devices especially designed to produce heat, either for direct use or to produce steam to drive turbines to generate electricity or propel ships. A general outline of basic reactors (thermal and fast) is given and then the different designs considered. The first are gas cooled, including the Magnox reactors (a list of UK Magnox stations and reactor performance is given), advanced gas cooled reactors (a list of UK AGRs is given) and the high temperature reactor. Light water cooled reactors (pressurized water [PWR] and boiling water [BWR] reactors) are considered next. Heavy water reactors are explained and listed. The pressurized heavy water reactors (including CANDU type reactors), boiling light water, steam generating heavy water reactors and gas cooled heavy water reactors all come into this category. Fast reactors (liquid metal fast breeder reactors and gas cooled fast reactors) and then water-cooled graphite-moderated reactors (RBMK) (the type at Chernobyl-4) are discussed. (U.K.)

  4. Safeguards Licensing Aspects of a Future Gen IV Test Facility - a Case Study

    International Nuclear Information System (INIS)

    Lindell, M. Aberg; Grape, S.; Hakansson, A.; Svaerd, S. Jacobsson

    2010-01-01

    The scope of this study covers safeguards licensing aspects of a possible future Gen IV demonstration facility. As a basis for the investigation, the facility was assumed to be located in Sweden, comprising a lead-cooled fast reactor and a reprocessing plant with fuel fabrication. The aim has been to identify safeguards requirements that may be set by the IAEA and the Swedish Radiation Safety Authority, and also to suggest how the safeguards system could be implemented in practice. The changed usage and handling of nuclear fuel, as compared to that of today, has been examined in order to determine how today's safeguards measures can be modified and extended to meet the needs of the demonstration facility. This work is part of GENIUS, the Swedish Gen IV research and development programme, which emphasizes lead-cooled fast reactors. (author)

  5. Fast reactor programme

    International Nuclear Information System (INIS)

    Plakman, J.C.

    1981-06-01

    The accuracy requirements and the status of the evaluated fission-product cross sections for fast reactors are reviewed; the work on calculating the sensitivity of the sodium void effect to fission-product cross sections is described; some results of the intercomparison of adjusted data sets for capture cross sections of fission-products (RCN-2A and CARNAVAL-IV) are discussed; the applicability of the maximum-likelihood method for the analysis of resolved resonance parameters for a large class of fission-product nuclides is demonstrated; the neutron cross sections for corrosion product 64 Ni are evaluated. Some results of post-irradiation examination of a loss-of-cooling experiment are given; the progress in testing the equipment and instrumentation for transient-overpower experiments is reported. The proceedings in the thermochemical investigations on uranium compounds with some fission-products are described. The creep behaviour of a heat of DIN 1.4948 parent metal is investigated with respect to the changes in strain with different test temperatures. Sodium smoke aerosols have been produced and analysed with respect to their aerodynamic behaviour and morphology. The two-phase local boiling experiments have been analysed to find criteria for the occurrence of different boiling regimes with the objection to deduce general dryout correlations

  6. Nuclear cycle of thorium and molt salts reactors. PE 5.8

    International Nuclear Information System (INIS)

    Doubre, H.

    2004-01-01

    In the framework of the nuclear industry development, many scenario are studied from the standard reactors using enriched uranium to the IV generation reactors. The study of new systems for the future of the nuclear needs to develop new simulation tools. The research programs of the IPN of Orsay are presented. (A.L.B.)

  7. Reactor safety

    International Nuclear Information System (INIS)

    Meneley, D.A.

    The people of Ontario have begun to receive the benefits of a low cost, assured supply of electrical energy from CANDU nuclear stations. This indigenous energy source also has excellent safety characteristics. Safety has been one of the central themes of the CANDU development program from its very beginning. A great deal of work has been done to establish that public risks are small. However, safety design criteria are now undergoing extensive review, with a real prospect of more stringent requirements being applied in the future. Considering the newness of the technology it is not surprising that a consensus does not yet exist; this makes it imperative to discuss the issues. It is time to examine the policies and practice of reactor safety management in Canada to decide whether or not further restrictions are justified in the light of current knowledge

  8. Nuclear reactor

    International Nuclear Information System (INIS)

    Schabert, H.P.; Weber, R.; Bauer, A.

    1975-01-01

    The refuelling of a PWR power reactor of about 1,200 MWe is performed by a transport pipe in the containment leading from an external to an internal fuel pit. A wagon to transport the fuel elements can go from a vertical loading position to an also vertical deloading position in the inner fuel pit via guide rollers. The necessary horizontal movement is effected by means of a cable line through the transport pipe which is inclined at least 10 0 . Gravity thus helps in the movement to the deloading position. The cable line with winch is fastened outside the containment. Swivelling devices tip the wagon from the horizontal to the vertical position or vice versa. Loading and deloading are done laterally. (TK/LH) [de

  9. Nuclear reactor

    International Nuclear Information System (INIS)

    Schweiger, F.; Glahe, E.

    1976-01-01

    In a nuclear reactor of the kind which is charged with spherical reaction elements and in which control rods are arranged to be thrust directly into the charge, each control rod has at least one screw thread on its external surface so that as the rod is thrust into the charge it is caused to rotate and thus make penetration easier. The length of each control rod may have two distinct portions, a latter portion which carries a screw thread and a lead-in portion which is shorter than the latter portion and which may carry a thread of greater pitch than that on the latter portion or may have a number of axially extending ribs instead of a thread

  10. Reactor container

    International Nuclear Information System (INIS)

    Furukawa, Hideyasu; Oyamada, Osamu; Uozumi, Hiroto.

    1976-01-01

    Purpose: To provide a container for a reactor provided with a pressure suppressing chamber pool which can prevent bubble vibrating load, particularly negative pressure generated at the time of starting to release exhaust from a main steam escape-safety valve from being transmitted to a lower liner plate of the container. Constitution: This arrangement is characterized in that a safety valve exhaust pool for main steam escape, in which a pressure suppressing chamber pool is separated and intercepted from pool water in the pressure suppressing chamber pool, a safety valve exhaust pipe is open into said safety valve exhaust pool, and an isolator member, which isolates the bottom liner plate in the pressure suppressing chamber pool from the pool water, is disposed on the bottom of the safety valve exhaust pool. (Nakamura, S.)

  11. Fluid element in SAP IV

    International Nuclear Information System (INIS)

    Yilmaz, C.; Akkas, N.

    1979-01-01

    In previous studies a fluid element is incorporated in the widely used general purpose finite element program SAPIV. This type of problem is of interest in the design of nuclear components involving geometric complexities and nonlinearities. The elasticity matrix of a general-purpose finite element program is modified in such a way that it becomes possible to idealize fluid as a structural finite element with zero shear modulus and a given bulk modules. Using the modified version of SAPIV, several solid-fluid interactions problems are solved. The numerical solutions are compared with the available analytical solutions. They are shown to be in reasonable aggrement. It is also shown that by solving an exterior-fluid interaction problem, the pressure wave propagation in the acoustic medium can be solved with the same approach. In this study, two of the problem not studied in the previous work will be presented. These problems are namely the effects of the link elements used at solid-fluid interfaces and of the concentrated loads on the response of the fluid medium. Truss elements are used as the link elements. After these investigations, it is decided that general purpose finite element programs with slight modifications can be used in the safety analysis of nuclear reactor plants. By this procedure it is possible to handle two-dimensional plane strain and tridimensional axisymmetric problems of this type. (orig.)

  12. Nuclear reactors

    International Nuclear Information System (INIS)

    Prescott, R.F.

    1976-01-01

    In the system described the fuel elements are arranged vertically in groups and are supported in such a manner as to tend to tilt them towards the center of the respective group, the fuel elements being urged laterally into abutment with one another. The elements have interlocking bearing pads, whereby lateral movement of adjacent elements is resisted; this improves the stability of the reactor core during refuelling operations. Fuel elements may comprise clusters of parallel fuel pins enclosed in a wrapper of hexagonal cross section, with bearing pads in the form of spline-like ribs located on each side of the wrapper and extending parallel to the longitudinal axis of the fuel element, being interlockable with ribs on pads of adjacent fuel elements. The arrangement is applicable to a reactor core in which fuel elements and control rod guide tubes are arranged in modules each of which comprises a cluster of at least three fuel elements, one of which is rigidly supported whilst the others are resiliently tilted towards the center of the cluster so as to lean on the rigidly supported element. It is also applicable to modules comprising a cluster of six fuel elements, each resiliently tilted towards a central void to form a circular arch. The modules may include additional fuel elements located outside the clusters and also resiliently tilted towards the central voids, the latter being used to accommodate control rod guide tubes. The need for separate structural members to act as leaning posts is thus avoided. Such structural members are liable to irradiation embrittlement, that could lead to core failure. (U.K.)

  13. Developing new nuclear curricula for GEN IV needs

    International Nuclear Information System (INIS)

    Ghitescu, P.; Pavel, G.L.

    2014-01-01

    States who wish to start and develop a nuclear program must take into consideration a strong proven strategy for developing a sustainable program. A complete nuclear research program must include: a good national strategy and support on the topic; strong research laboratories supported by good personnel; education component to provide sustainable and qualified workforce; national/international interest from stakeholders and governments and a well informed society. New demonstrators are foreseen for the next period to be built in Europe and skilled supporting personnel is strongly needed. Current situation in nuclear higher education with perspective will be analysed. EURATOM strongly supports development of multidisciplinary co-operational projects in order to built such novel initiatives. An example of such program supported by European Commission, ARCADIA, will be given. The project is based on the cooperation of a large number of participants all over Europe and the main purpose is to develop a road-map for Gen IV reactor. (authors)

  14. Zirconium (IV) complexes with some polymethylenediimines | Na ...

    African Journals Online (AJOL)

    The syntheses of zirconium (IV) complexes have been carried out by the reaction of oxozirconium (IV) chloride with the appropriate diimines (Schiff bases). The complexes were isolated as yellow solids which are stable to heat. The complexes were found to be insoluble in most solvents. The infrared spectra, elemental ...

  15. Astragaloside IV liposomes ameliorates adriamycin-induced ...

    African Journals Online (AJOL)

    Methods: The rats were given a single tail intravenous injection of adriamycin (6 mg/kg) within 1 week, and then divided into four groups including normal, model, benazepril and astragaloside IV liposomes group. They were all orally administered dosage of benazepril and astragaloside IV liposomes once daily for 8 weeks.

  16. Innovative reactor core: potentialities and design

    International Nuclear Information System (INIS)

    Artioli, C.; Petrovich, Carlo; Grasso, Giacomo

    2010-01-01

    Gen IV nuclear reactors are considered a very attractive answer for the demand of energy. Because public acceptance they have to fulfil very clearly the requirement of sustainable development. In this sense a reactor concept, having by itself a rather no significant interaction with the environment both on the front and back end ('adiabatic concept'), is vital. This goal in mind, a new way of designing such a core has to be assumed. The starting point must be the 'zero impact'. Therefore the core will be designed having as basic constraints: a) fed with only natural or depleted Uranium, and b) discharges only fission products. Meantime its potentiality as a net burner of Minor Actinide has to be carefully estimated. This activity, referred to the ELSY reactor, shows how to design such an 'adiabatic' core and states its reasonable capability of burning MA legacy in the order of 25-50 kg/GW e y. (authors)

  17. Nuclear reactor neutron shielding

    Science.gov (United States)

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

    2017-09-12

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

  18. Fuel research for subcritical and critical GEN-IV systems cooled by heavy liquid metal

    International Nuclear Information System (INIS)

    Sobolev, V.; Verwerft, M.

    2009-01-01

    The participation of the Belgian Nuclear Research Centre SCK-CEN in the worldwide GEN-IV research can be considered as an opportunity. Today's GEN-IV research at SCK-CEN is mainly driven by the interests of the project MYRRHA (Multipurpose hYbrid Research Reactor for High-tech Applications). The main goal of this project is to build at SCK-CEN in Mol a new generation fast spectrum, subcritical, research and materials testing reactor MYRRHA driven by a high-energy proton accelerator. This GEN-IV MTR is cooled by heavy liquid metal (Pb-Bi) and will be used for the ADS concept demonstration, testing and qualification of new fuels, transmutation targets and innovative materials. On the European scale, MYRRHA is integrated in the Euratom FP6 Integrated Project (IP) EUROTRANS (EUROpean research programme for TRANSmutation of high level nuclear waste in an accelerator driven system), as the small-scale experimental machine for transmutation demonstration called XT-ADS. Last but not least, this experimental facility will also demonstrate the technological feasibility of the LFR (Lead-cooled Fast Reactor) GEN-IV concept; in EU the LFR design studies are performed in the framework of the Euratom FP6 ELSY (European Lead-cooled SYstem) project, where SCK-CEN is a partner. Among the research needed to ensure a safe and reliable operation of the MYRRHA/XT ADS reactor, the development and qualification of fuel and cladding materials have been recognized as one of the main key issues to be addressed

  19. Chemistry and technology of Molten Salt Reactors - history and perspectives

    International Nuclear Information System (INIS)

    Uhlir, Jan

    2007-01-01

    Molten Salt Reactors represent one of promising future nuclear reactor concept included also in the Generation IV reactors family. This reactor type is distinguished by an extraordinarily close connection between the reactor physics and chemical technology, which is given by the specific features of the chemical form of fuel, representing by molten fluoride salt and circulating through the reactor core and also by the requirements of continuous 'on-line' reprocessing of the spent fuel. The history of Molten Salt Reactors reaches the period of fifties and sixties, when the first experimental Molten Salt Reactors were constructed and tested in ORNL (US). Several molten salt techniques dedicated to fresh molten salt fuel processing and spent fuel reprocessing were studied and developed in those days. Today, after nearly thirty years of discontinuance, a renewed interest in the Molten Salt Reactor technology is observed. Current experimental R and D activities in the area of Molten Salt Reactor technology are realized by a relatively small number of research institutions mainly in the EU, Russia and USA. The main effort is directed primarily to the development of separation processes suitable for the molten salt fuel processing and reprocessing technology. The techniques under development are molten salt/liquid metal extraction processes, electrochemical separation processes from the molten salt media, fused salt volatilization techniques and gas extraction from the molten salt medium

  20. Thermodynamic data for predicting concentrations of Th(IV), U(IV), Np(IV), and Pu(IV) in geologic environments

    Energy Technology Data Exchange (ETDEWEB)

    Rai, Dhanpat; Roa, Linfeng; Weger, H.T.; Felmy, A.R. [Battelle, Pacific Northwest National Laboratory (PNNL) (United States); Choppin, G.R. [Florida State University (United States); Yui, Mikazu [Waste Isolation Research Division, Tokai Works, Japan Nuclear Cycle Development Inst., Tokai, Ibaraki (Japan)

    1999-01-01

    This report provides thermodynamic data for predicting concentrations of Th(IV), U(IV), Np(IV), and Pu(IV) in geologic environments, and contributes to an integration of the JNC chemical thermodynamic database, JNC-TDB (previously PNC-TDB), for the performance analysis of geological isolation system for high-level radioactive wastes. Thermodynamic data for the formation of complexes or compounds with hydroxide, chloride, fluoride, carbonate, nitrate, sulfate and phosphate are discussed in this report. Where data for specific actinide(IV) species was lacking, the data were selected based on chemical analogy to other tetravalent actinides. In this study, the Pitzer ion-interaction model is used to extrapolate thermodynamic constants to zero ionic strength at 25degC. (author)

  1. Site Environmental Report for Calendar Year 2011. DOE Operations at The Boeing Company Santa Susana Field Laboratory, Area IV

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Ning [The Boeing Company, Canoga Park, CA (United States); Rutherford, Phil [The Boeing Company, Canoga Park, CA (United States); Dassler, David [The Boeing Company, Canoga Park, CA (United States)

    2012-09-01

    This Annual Site Environmental Report (ASER) for 2011 describes the environmental conditions related to work performed for the Department of Energy (DOE) at Area IV of Boeing’s Santa Susana Field Laboratory (SSFL). The Energy Technology Engineering Center (ETEC), a government-owned, company-operated test facility, was located in Area IV. The operations in Area IV included development, fabrication, operation and disassembly of nuclear reactors, reactor fuel, and other radioactive materials. Other activities in the area involved the operation of large-scale liquid metal facilities that were used for testing non-nuclear liquid metal fast breeder reactor components. All nuclear work was terminated in 1988, and all subsequent radiological work has been directed toward environmental restoration and decontamination and decommissioning (D&D) of the former nuclear facilities and their associated sites. Liquid metal research and development ended in 2002. Since May 2007, the D&D operations in Area IV have been suspended by the DOE, but the environmental monitoring and characterization programs have continued. Results of the radiological monitoring program for the calendar year 2011 continue to indicate that there are no significant releases of radioactive material from Area IV of SSFL. All potential exposure pathways are sampled and/or monitored, including air, soil, surface water, groundwater, direct radiation, transfer of property (land, structures, waste), and recycling.

  2. Site Environmental Report For Calendar Year 2012. DOE Operations at The Boeing Company Santa Susana Field Laboratory, Area IV

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Ning [The Boeing Company, Canoga Park, CA (United States); Rutherford, Phil [The Boeing Company, Canoga Park, CA (United States); Dassler, David [The Boeing Company, Canoga Park, CA (United States)

    2013-09-01

    This Annual Site Environmental Report (ASER) for 2012 describes the environmental conditions related to work performed for the Department of Energy (DOE) at Area IV of Boeing’s Santa Susana Field Laboratory (SSFL). The Energy Technology Engineering Center (ETEC), a government-owned, company-operated test facility, was located in Area IV. The operations in Area IV included development, fabrication, operation and disassembly of nuclear reactors, reactor fuel, and other radioactive materials. Other activities in the area involved the operation of large-scale liquid metal facilities that were used for testing non-nuclear liquid metal fast breeder reactor components. All nuclear work was terminated in 1988, and all subsequent radiological work has been directed toward environmental restoration and decontamination and decommissioning (D&D) of the former nuclear facilities and their associated sites. Liquid metal research and development ended in 2002. Since May 2007, the D&D operations in Area IV have been suspended by the DOE, but the environmental monitoring and characterization programs have continued. Results of the radiological monitoring program for the calendar year 2012 continue to indicate that there are no significant releases of radioactive material from Area IV of SSFL. All potential exposure pathways are sampled and/or monitored, including air, soil, surface water, groundwater, direct radiation, transfer of property (land, structures, waste), and recycling.

  3. Site Environmental Report for Calendar Year 2009. DOE Operations at The Boeing Company Santa Susana Field Laboratory, Area IV

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Ning [The Boeing Company, Canoga Park, CA (United States); Rutherford, Phil [The Boeing Company, Canoga Park, CA (United States); Amar, Ravnesh [The Boeing Company, Canoga Park, CA (United States)

    2010-09-01

    This Annual Site Environmental Report (ASER) for 2009 describes the environmental conditions related to work performed for the Department of Energy (DOE) at Area IV of Boeing’s Santa Susana Field Laboratory (SSFL). The Energy Technology Engineering Center (ETEC), a government-owned, company-operated test facility, was located in Area IV. The operations in Area IV included development, fabrication, and disassembly of nuclear reactors, reactor fuel, and other radioactive materials. Other activities in the area involved the operation of large-scale liquid metal facilities that were used for testing non-nuclear liquid metal fast breeder reactor components. All nuclear work was terminated in 1988, and all subsequent radiological work has been directed toward decontamination and decommissioning (D&D) of the former nuclear facilities and their associated sites. Liquid metal research and development ended in 2002. Since May 2007, the D&D operations in Area IV have been suspended by the DOE, but the environmental monitoring and characterization programs have continued. Results of the radiological monitoring program for the calendar year 2009 continue to indicate that there are no significant releases of radioactive material from Area IV of SSFL. All potential exposure pathways are sampled and/or monitored, including air, soil, surface water, groundwater, direct radiation, transfer of property (land, structures, waste), and recycling.

  4. Site Environmental Report for Calendar Year 2010. DOE Operations at The Boeing Company Santa Susana Field Laboratory, Area IV

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Ning [The Boeing Company, Canoga Park, CA (United States); Rutherford, Phil [The Boeing Company, Canoga Park, CA (United States); Amar, Ravnesh [The Boeing Company, Canoga Park, CA (United States)

    2011-09-01

    This Annual Site Environmental Report (ASER) for 2010 describes the environmental conditions related to work performed for the Department of Energy (DOE) at Area IV of Boeing’s Santa Susana Field Laboratory (SSFL). The Energy Technology Engineering Center (ETEC), a government-owned, company-operated test facility, was located in Area IV. The operations in Area IV included development, fabrication, and disassembly of nuclear reactors, reactor fuel, and other radioactive materials. Other activities in the area involved the operation of large-scale liquid metal facilities that were used for testing non-nuclear liquid metal fast breeder reactor components. All nuclear work was terminated in 1988, and all subsequent radiological work has been directed toward decontamination and decommissioning (D&D) of the former nuclear facilities and their associated sites. Liquid metal research and development ended in 2002. Since May 2007, the D&D operations in Area IV have been suspended by the DOE, but the environmental monitoring and characterization programs have continued. Results of the radiological monitoring program for the calendar year 2010 continue to indicate that there are no significant releases of radioactive material from Area IV of SSFL. All potential exposure pathways are sampled and/or monitored, including air, soil, surface water, groundwater, direct radiation, transfer of property (land, structures, waste), and recycling.

  5. Calculation of fundamental parameters for the dynamical study of TRIGA-3-Salazar reactor (Mixed reactor core)

    International Nuclear Information System (INIS)

    Viais J, J.

    1994-01-01

    Kinetic parameters for dynamic study of two different configurations, 8 and 9, both with standard fuel, 20% enrichment and Flip (Fuel Life Improvement Program with 70% enrichment) fuel, for TRIGA Mark-III reactor from Mexico Nuclear Center, are obtained. A calculation method using both WIMS-D4 and DTF-IV and DAC1 was established, to decide which of those two configurations has the best safety and operational conditions. Validation of this methodology is done by calculate those parameters for a reactor core with new standard fuel. Configuration 9 is recommended to be use. (Author)

  6. U.S. Status of Fast Reactor Research and Technology

    International Nuclear Information System (INIS)

    Hill, Robert

    2012-01-01

    Summary: • Fast reactor R&D is focused on key technologies innovations for performance improvement (cost reduction) and safety: 1. System Integration and Concept Development; 2. Safety Technology; 3. Advanced Materials; 4. Ultrasonic Viewing; 5. Advanced Energy Conversion (Supercritical CO 2 Brayton cycle); 6. Reactor Simulation; 7. Nuclear Data; 8. Advanced Fuels. • Fast reactors have flexible capability for actinide management: – A wide variety of fuel cycle options are being considered; • International R&D collaboration pursued in Generation-IV and multilateral arrangements

  7. The story of fission reactors: from Chicago Pile to advanced energy systems

    International Nuclear Information System (INIS)

    Kannan, Umasankari

    2017-01-01

    Nuclear reactors have been designed which cater to different applications from small research reactors of a few watts to power reactors of several Giga Watts. Based on the neutron energy, there are thermal, intermediate and fast reactors operating are being designed. On the fuel utilization front, there are designs ranging from reactors using natural uranium fuel to enriched uranium to more efficient thorium based reactors. Reactors have also been designed which are neutron eaters, minor actinide burners and breeders. There have been variety of coolant and moderating materials used for different applications from water, gas cooled, liquid sodium cooled to molten salt cooled reactors. Several new reactor designs have been developed using innovative concepts in high temperature reactors, nuclear power packs and compact reactors for special purposes. The design challenges are many from modest designs to complicated hybrid reactors. The GEN-IV forum of IAEA has selected a few of these reactor designs for commercial power production in the coming years based on several quantified indicators. The evolutionary and revolutionary design approaches have been made over the years catering to different need of energy generation. A glimpse of some of the reactors being currently developed and the design modifications done in existing reactors have been given in this paper

  8. Development of a Neutron Flux Monitoring System for Sodium-cooled Fast Reactors

    OpenAIRE

    Verma, Vasudha

    2017-01-01

    Safety and reliability are one of the key objectives for future Generation IV nuclear energy systems. The neutron flux monitoring system forms an integral part of the safety design of a nuclear reactor and must be able to detect any irregularities during all states of reactor operation. The work in this thesis mainly concerns the detection of in-core perturbations arising from unwanted movements of control rods with in-vessel neutron detectors in a sodium-cooled fast reactor. Feasibility stud...

  9. FBR type reactor

    International Nuclear Information System (INIS)

    Kimura, Kimitaka; Fukuie, Ken; Iijima, Tooru; Shimpo, Masakazu.

    1994-01-01

    In an FBR type reactor for exchanging fuels by pulling up reactor core upper mechanisms, a connection mechanism is disposed for connecting the top of the reactor core and the lower end of the reactor core upper mechanisms. In addition, a cylindrical body is disposed surrounding the reactor core upper mechanisms, and a support member is disposed to the cylindrical body for supporting an intermediate portion of the reactor core upper mechanisms. Then, the lower end of the reactor core upper mechanisms is connected to the top of the reactor core. Same displacements are caused to both of them upon occurrence of earthquakes and, as a result, it is possible to eliminate mutual horizontal displacement between a control rod guide hole of the reactor core upper mechanisms and a control rod insertion hole of the reactor core. In addition, since the intermediate portion of the reactor core upper mechanisms is supported by the support member disposed to the cylindrical body surrounding the reactor core upper mechanisms, deformation caused to the lower end of the reactor core upper mechanisms is reduced, so that the mutual horizontal displacement with respect to the control rod insertion hole of the reactor core can be reduced. As a result, performance of control rod insertion upon occurrence of the earthquakes is improved, so that reactor shutdown is conducted more reliably to improve reactor safety. (N.H.)

  10. The prototype fast reactor

    International Nuclear Information System (INIS)

    Broomfield, A.M.

    1985-01-01

    The paper concerns the Prototype Fast Reactor (PFR), which is a liquid metal cooled fast reactor power station, situated at Dounreay, Scotland. The principal design features of a Fast Reactor and the PFR are given, along with key points of operating history, and health and safety features. The role of the PFR in the development programme for commercial reactors is discussed. (U.K.)

  11. Department of reactor technology

    International Nuclear Information System (INIS)

    1980-01-01

    The activities of the Department of Reactor Technology at Risoe during 1979 are described. The work is presented in five chapters: Reactor Engineering, Reactor Physics and Dynamics, Heat Transfer and Hydraulics, The DR 1 Reactor, and Non-Nuclear Activities. A list of the staff and of publications is included. (author)

  12. NCSU Reactor Sharing Program

    International Nuclear Information System (INIS)

    Perez, P.B.

    1993-01-01

    The Nuclear Reactor Program at North Carolina State University provides the PULSTAR Research Reactor and associated facilities to eligible institutions with support, in part, from the Department of Energy Reactor Sharing Program. Participation in the NCSU Reactor Sharing Program continues to increase steadily with visitors ranging from advance high school physics and chemistry students to Ph.D. level research from neighboring universities

  13. Reactor safety method

    International Nuclear Information System (INIS)

    Vachon, L.J.

    1980-01-01

    This invention relates to safety means for preventing a gas cooled nuclear reactor from attaining criticality prior to start up in the event the reactor core is immersed in hydrogenous liquid. This is accomplished by coating the inside surface of the reactor coolant channels with a neutral absorbing material that will vaporize at the reactor's operating temperature

  14. Methodology of fuel rod design for pressurized light water reactors

    International Nuclear Information System (INIS)

    Teixeira e Silva, A.; Esteves, A.M.

    1988-01-01

    The fuel performance program FRAPCON-1 and the structural finite element program SAP-IV are applied in a pressurized water reactor fuel rod design methodology. The applied calculation procedure allows to dimension the fuel rod components and characterize its internal pressure. (author) [pt

  15. Computerised programming of the Dragon reactor fuel handling operations

    International Nuclear Information System (INIS)

    Butcher, P.

    1976-11-01

    Two suites of FORTRAN IV computer programs have been written to produce check lists for the operation of the two remote control fuel handling machines of the Dragon Reactor. This document describes the advantages of these programs over the previous manual system of writing check lists, and provides a detailed guide to the programs themselves. (author)

  16. Physics of nuclear reactors

    International Nuclear Information System (INIS)

    Baeten, Peter

    2006-01-01

    This course gives an introduction to Nuclear Reactor Physics. The first chapter explains the most important parameters and concepts in nuclear reactor physics such as fission, cross sections and the effective multiplication factor. Further on, in the second chapter, the flux distributions in a stationary reactor are derived from the diffusion equation. Reactor kinetics, reactor control and reactor dynamics (feedback effects) are described in the following three chapters. The course concludes with a short description of the different types of existing and future reactors. (author)

  17. Reactor core and initially loaded reactor core of nuclear reactor

    International Nuclear Information System (INIS)

    Koyama, Jun-ichi; Aoyama, Motoo.

    1989-01-01

    In BWR type reactors, improvement for the reactor shutdown margin is an important characteristic condition togehter with power distribution flattening . However, in the reactor core at high burnup degree, the reactor shutdown margin is different depending on the radial position of the reactor core. That is , the reactor shutdown margin is smaller in the outer peripheral region than in the central region of the reactor core. In view of the above, the reactor core is divided radially into a central region and as outer region. The amount of fissionable material of first fuel assemblies newly loaded in the outer region is made less than the amount of the fissionable material of second fuel assemblies newly loaded in the central region, to thereby improve the reactor shutdown margin in the outer region. Further, the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower portion of the first fuel assemblies is made smaller than the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower region of the second fuel assemblies, to thereby obtain a sufficient thermal margin in the central region. (K.M.)

  18. Adduct formation in Ce(IV) thenolytrifluoroacetonate

    International Nuclear Information System (INIS)

    Anufrieva, S.I.; Polyakova, G.V.; Snezhko, N.I.; Pechurova, N.I.; Martynenko, L.I.; Spitsyn, V.I.

    1982-01-01

    The literature contains no information on adduct formation in Ce(IV) β-diketonates with additional ligands. Since tetrakis-β-diketonates of Ce(IV) have four six-membered chelate rings, we can suppose that the introduction of an additional monodentate or bidentate ligand into the coordination sphere of Ce(IV) β-diketonates would lead to an increase in the coordination number (CN) of the Ce(IV) to nine or ten. The possibility of realization of such a high CN for Ce(IV) has not been proved; a study of adduct formation by Ce(IV) tetrakis-β-diketonates is thus of theoretical interest. Such an investigation might also be of practical interest, because the introduction of an additional ligand into the coordination sphere of a rare-earth β-diketonate usually increases the solubility of the β-diketonate in nonpolar solvents and increases the volatility of the compound; such a modification of the properties is important for various practical purposes. The aim of our work was to study the possibility of separating solid adducts of Ce(IV) tetrakis-thenoyltrifluoroacetonate with certain oxygen-containing and nitrogen-containing donor monodentate and bidentate ligands, and also to investigate their properties. As the β-diketone we used thenoyltrifluoroacetone (HTTFA), since in a parallel investigation it was found that Ce(TTFA) 4 has a high oxidation-reduction stability

  19. Nuclear reactors. Introduction

    International Nuclear Information System (INIS)

    Boiron, P.

    1997-01-01

    This paper is an introduction to the 'nuclear reactors' volume of the Engineers Techniques collection. It gives a general presentation of the different articles of the volume which deal with: the physical basis (neutron physics and ionizing radiations-matter interactions, neutron moderation and diffusion), the basic concepts and functioning of nuclear reactors (possible fuel-moderator-coolant-structure combinations, research and materials testing reactors, reactors theory and neutron characteristics, neutron calculations for reactor cores, thermo-hydraulics, fluid-structure interactions and thermomechanical behaviour of fuels in PWRs and fast breeder reactors, thermal and mechanical effects on reactors structure), the industrial reactors (light water, pressurized water, boiling water, graphite moderated, fast breeder, high temperature and heavy water reactors), and the technology of PWRs (conceiving and building rules, nuclear parks and safety, reactor components and site selection). (J.S.)

  20. Gas cooled fast reactor research in Europe

    International Nuclear Information System (INIS)

    Stainsby, Richard; Peers, Karen; Mitchell, Colin; Poette, Christian; Mikityuk, Konstantin; Somers, Joe

    2011-01-01

    Research on the gas-cooled fast reactor system is directed towards fulfilling the ambitious long term goals of Generation IV (Gen IV), i.e., to develop a safe, sustainable, reliable, proliferation-resistant and economic nuclear energy system. In common with other fast reactors, gas-cooled fast reactors (GFRs) have exceptional potential as sustainable energy sources, for both the utilisation of fissile material and minimisation of nuclear waste through transmutation of minor actinides. The primary goal of GFR research is to develop the system primarily to be a reliable and economic electricity generator, with good safety and sustainability characteristics. However, for the longer term, GFR retains the potential for hydrogen production and other process heat applications facilitated through a high core outlet temperature which, in this case, is not limited by the characteristics of the coolant. In this respect, GFR can inherit the non-electricity applications of the thermal HTRs in a sustainable manner in a future in which natural uranium becomes scarce. GFR research within Europe is performed directly by those states who have signed the 'System Arrangement' document within the Generation IV International Forum (the GIF), specifically France and Switzerland and Euratom. Importantly, Euratom provides a route by which researchers in other European states, and other non-European affiliates, can contribute to the work of the GIF, even when these states are not signatories to the GFR System Arrangement in their own right. This paper is written from the perspective of Euratom's involvement in research on the GFR system, starting with the 5th Framework Programme (FP5) GCFR project in 2000, through the FP6 project between 2005 and 2009 and looking ahead to the proposed activities within the current 7th Framework Programme (FP7). The evolution of the GFR concept from the 1960s onwards is discussed briefly, followed by the current perceived role, objectives and progress with

  1. Comparison between steady-state and dynamic I-V measurements from a single-cell thermionic fuel element

    International Nuclear Information System (INIS)

    Wernsman, Bernard

    1997-01-01

    A comparison between steady-state and dynamic I-V measurements from a single-cell thermionic fuel element (TFE) is made. The single-cell TFE used in this study is the prototype for the 40 kW e space nuclear power system that is similar to the 6 kW e TOPAZ-II. The steady-state I-V measurements influence the emitter temperature due to electron cooling. Therefore, to eliminate the steady-state I-V measurement influence on the TFE and provide a better understanding of the behavior of the thermionic energy converter and TFE characteristics, dynamic I-V measurements are made. The dynamic I-V measurements are made at various input power levels, cesium pressures, collector temperatures, and steady-state current levels. From these measurements, it is shown that the dynamic I-V's do not change the TFE characteristics at a given operating point. Also, the evaluation of the collector work function from the dynamic I-V measurements shows that the collector optimization is not due to a minimum in the collector work function but due to an emission optimization. Since the dynamic I-V measurements do not influence the TFE characteristics, it is believed that these measurements can be done at a system level to understand the influence of TFE placement in the reactor as a function of the core thermal distribution

  2. Development of methods for theoretical analysis of nuclear reactors (Phase II), I-V, Part IV, Fuel depletion

    International Nuclear Information System (INIS)

    Pop-Jordanov, J.

    1962-10-01

    This report includes the analysis of plutonium isotopes from U 238 depletion chain. Two theoretical approaches for solving the depletion of fuel are shown. One results in the system of differential equations that can be solved only by using electronic calculators and the second, Machinari-Goto method enables obtaining analytical equations for approximative values of particular nuclei. In addition, differential equations are given for different approximation levels in calculating Pu 239 , as well as relations between the released energy and irradiation [sr

  3. LOCA analysis of the IRIS reactor

    International Nuclear Information System (INIS)

    Bajs, T.; Grgic, D.; Cavlina, N.

    2003-01-01

    The IRIS reactor (International Reactor Innovative and Secure) is an integral, light water cooled, medium power reactor. IRIS has been selected as an International Near Term Deployable (INTD) reactor, within the Generation IV International Forum activities. The IRIS concept addresses the key-requirements defined by the US DOE for next generation reactors, i.e. enhanced reliability and safety, and improved economics. It features innovative, advanced engineering, but it is firmly based on the proven technology of pressurized water reactors (PWR). An innovative safety approach has been developed to mitigate the IRIS response to small-to-medium Loss of Coolant Accident (LOCA). This strategy is based on the interaction of IRIS compact containment with the reactor vessel to limit initial blowdown, and on depressurization through the use of a passive Emergency Heat Removal System (EHRS). A small Automatic Depressurization System (ADS) provides supplementary depressurization capability. A pressure suppression system is provided to limit the pressure peak following the initial blowdown to well below the containment design limit. The ultimate result is that during a small-to-medium LOCA, the core remains covered for an extended period of time, without credit for emergency water injection or external core makeup. The IRIS LOCA response is based on 'maintaining water inventory' rather than on the principle of safety injection. This novel safety approach poses significant issues for computational and analysis methods since the IRIS vessel and containment are strongly coupled, and the system response is based on the interaction between the two. The small break LOCA was calculated using RELAP5/mod3.3 and GOTHIC codes. Break of the largest line connected to the IRIS Reactor Pressure Vessel (RPV) was analyzed. The results of the calculations confirmed good performance of the IRIS system during LOCA. (author)

  4. Coupled thermal-hydraulic and neutronic simulations of Phenix control rod withdrawal tests with SIMMER-IV

    International Nuclear Information System (INIS)

    Kriventsev, Vladimir; Gabrielli, Fabrizio; Rineiski, Andrei

    2014-01-01

    The “end-of-life” tests performed in the Phenix reactor before its final shutdown in 2009, in particular the Control Rod (CR) withdrawal experiments provide an excellent opportunity for the validation and verification of the reactor physics computer codes and modeling approaches. SIMMER-IV, a modern three-dimensional reactor safety code, has been recently employed at Karlsruhe Institute of Technology (KIT) for simulating Phenix experiments in the framework of a benchmark exercise organized under the IAEA project. In this paper, we report and discuss main results obtained with SIMMER-IV at KIT. Particular attention is devoted to the coupling features of thermal-hydraulics and neutronics and their mutual influences. The reactor reactivity, power and neutron flux distributions calculated with SIMMER-IV are in good agreement both with experimental results and with calculations with advanced neutronics codes, such as ERANOS, while the CR reactivity worth is overestimated due to neglecting heterogeneity effects. Because of its multi-physics capabilities SIMMER also calculates the temperature distributions which are in a good agreement with the experimental test results. In this work we describe the improvements in SIMMER neutronics model by employing a correction that is based on the results of cell calculations performed with ERANOS. The study confirms that the 3D SIMMER-IV code can accurately predict major fast reactor neutronics and thermal hydraulic parameters, provided that a special treatment is employed for CR modeling. The results of calculations are analyzed in frames of SIMMER-IV validation and verification assessment. (author)

  5. Thermonuclear reactor

    International Nuclear Information System (INIS)

    Yasutomi, Yoshiyuki; Nakagawa, Moroo; Sawai, Yuichi; Chiba, Akio; Suzuki, Yasutaka.

    1997-01-01

    Silicon composited with reinforcing metals is used for a divertor cooling substrate having an effect as a cooling tube to provide a silicon base composite material having increased electric resistance and toughness. The blending ratio of reinforcing materials in the form of granules, whiskers or long fibers is controlled in order to control heat conductivity, electric resistivity and mechanical performances. The divertor cooling substrate comprising the silicon base composite material is integrated with a plasma facing material. The production method therefor includes ordinary metal matrix composite forming methods such as powder metallurgy, melting penetration method, high pressure solidification casting method, centrifugal casting method and vacuum casting method. Since the cooling plate is constituted with the light metal and highly electric resistant metal base composite material, sharing force due to eddy current can be reduced, and radiation exposure can be minimized. Accordingly, a cooling structure for a thermonuclear reactor effective for the improvement of environmental problems caused by waste disposal can be attained. (N.H.)

  6. Nuclear reactor

    International Nuclear Information System (INIS)

    Irion, L.; Tautz, J.; Ulrych, G.

    1976-01-01

    This additional patent complements the arrangement of non-return valves to prevent loss of cooling water on fracture of external tubes in the main coolant circuit (according to PS 24 24 427.7) by ensuring that the easily movable valves only operate in case of a fault, but do not flutter in operation, because the direction of flow is not the same at each location where they are installed. The remedy for this undesirable effect consists of allocating 1 non-return valve unit with 5 to 10 valves to each (of several) ducts for the cooling water intake. These units are installed in the annular space between the reactor vessel and the pressure vessel below the inlet of the ducts. Due to flow guidance surfaces in the same space, the incoming cooling water is deflected downwards and as the guiding surfaces are closed at the sides, must pass parallel to the valves of the non-return valve unit. On fracture of the external cooling water inlet pipe concerned, all valves of this unit close due to reversal of flow on the outlet side. (TK) [de

  7. Nuclear reactors

    International Nuclear Information System (INIS)

    Pearson, K.G.

    1977-01-01

    Reference is made to auxiliary means of cooling the nuclear fuel clusters used in light or heavy water cooled nuclear reactors. One method is to provide one or more spray cooling tubes. From holes in the side walls of those tubes coolant water may be sprayed laterally into the cluster against the rods. The flow of main coolant may thus be supplemented or even replaced by the auxiliary coolant. A difficulty, however, is that only those fuel rods close to a spray cooling tube can readily be reached by the auxiliary coolant. In the arrangement described, where the fuel rods are spaced apart by transverse grids, at least one of the interspaces between the grids is provided with an axially extending auxiliary coolant conduit having lateral holes through which an auxiliary coolant is sprayed into the cluster. A deflector is provided that extends from a transverse grid into a position in front of the holes and deflects auxiliary coolant on to parts of the fuel rods otherwise inaccessible to the auxiliary coolant. The construction of the deflector is described. (U.K.)

  8. Nuclear reactor coolant channels

    International Nuclear Information System (INIS)

    Macbeth, R.V.

    1978-01-01

    A nuclear reactor coolant channel is described that is suitable for sub-cooled reactors as in pressurised water reactors as well as for bulk boiling, as in boiling water reactors and steam generating nuclear reactors. The arrangement aims to improve heat transfer between the fuel elements and the coolant. Full constructional details are given. See also other similar patents by the author. (U.K.)

  9. Fast breeder reactors

    International Nuclear Information System (INIS)

    Heinzel, V.

    1975-01-01

    The author gives a survey of 'fast breeder reactors'. In detail the process of breeding, the reasons for the development of fast breeders, the possible breeder reactors, the design criteria, fuels, cladding, coolant, and safety aspects are reported on. Design data of some experimental reactors already in operation are summarized in stabular form. 300 MWe Prototype-Reactors SNR-300 and PFR are explained in detail and data of KWU helium-cooled fast breeder reactors are given. (HR) [de

  10. Evaluation of LLNL's Nuclear Accident Dosimeters at the CALIBAN Reactor September 2010

    International Nuclear Information System (INIS)

    Hickman, D.P.; Wysong, A.R.; Heinrichs, D.P.; Wong, C.T.; Merritt, M.J.; Topper, J.D.; Gressmann, F.A.; Madden, D.J.

    2011-01-01

    participants were limited in what they were allowed to do at the Caliban and Silene exercises and testing of various elements of the nuclear accident dosimetry programs cannot always be performed as guests at other sites, it has become evident that DOE needs its own capability to test nuclear accident dosimeters. Angular dependence determination and correction factors for NADs desperately need testing as well as more evaluation regarding the correct determination of gamma doses. It will be critical to properly design any testing facility so that the necessary experiments can be performed by DOE laboratories as well as guest laboratories. Alternate methods of dose assessment such as using various metals commonly found in pockets and clothing have yet to be evaluated. The DOE is planning to utilize the Godiva or Flattop reactor for testing nuclear accident dosimeters. LLNL has been assigned the primary operational authority for such testing. Proper testing of nuclear accident dosimeters will require highly specific characterization of the pulse fields. Just as important as the characterization of the pulsed fields will be the design of facilities used to process the NADs. Appropriate facilities will be needed to allow for early access to dosimeters to test and develop quick sorting techniques. These facilities will need appropriate laboratory preparation space and an area for measurements. Finally, such a facility will allow greater numbers of LLNL and DOE laboratory personnel to train on the processing and interpretation of nuclear accident dosimeters and results. Until this facility is fully operational for test purposes, DOE laboratories may need to continue periodic testing as guests of other reactor facilities such as Silene and Caliban.

  11. Market opportunities: U.S. - PADD IV

    International Nuclear Information System (INIS)

    Garner, R.P.

    1997-01-01

    The current supply and demand balance, the short and long term expectations and marketing opportunities for Canadian crude oil in PADD IV, the Rocky Mountain region in the US, were reviewed. It was suggested that market opportunities in PADD IV are derived from the following four factors: (1) crude oil declines within that area, (2) federal regulations, (3) competitive presence with markets, and (4) population growth. The overall conclusion was that Canadian producers and PADD IV refiners will be looking at an ever-growing relationship based on freight equalized world crude prices. 8 tabs., 5 figs

  12. Recent designs for advanced fusion reactor blankets

    International Nuclear Information System (INIS)

    Sze, D.K.

    1994-01-01

    A series of reactor design studies based on the Tokamak configuration have been carried out under the direction of Professor Robert Conn of UCLA. They are called ARIES-I through IV. The key mission of these studies is to evaluate the attractiveness of fusion assuming different degrees of advancement in either physics or engineering development. This paper discusses the directions and conclusions of the blanket and related engineering systems for those design studies. ARIES-1 investigated the use of SiC composite as the structural material to increase the blanket temperature and reduce the blanket activation. Li 2 ZrO 3 was used as the breeding material due to its high temperature stability and good tritium recovery characteristics. The ARIES-IV is a modification of ARIES-1. The plasma was in the second stability regime. Li 2 O was used as the breeding material to remove Zr. A gaseous divertor was used to replace the conventional divertor so that high Z divertor target is not required. The physics of ARIES-II was the same as ARIES-IV. The engineering design of the ARIES-II was based on a self-cooled lithium blanket with a V-alloy as the structural material. Even though it was assumed that the plasma was in the second stability regime, the plasma beta was still rather low (3.4%). The ARIES-III is an advanced fuel (D- 3 He) tokamak reactor. The reactor design assumed major advancement on the physics, with a plasma beta of 23.9%. A conventional structural material is acceptable due to the low neutron wall loading. From the radiation damage point of view, the first wall can last the life of the reactor, which is expected to be a major advantage from the engineering design and waste disposal point of view

  13. Generation IV nuclear energy systems: road map and concepts. 2. Generation II Measurement Systems for Generation IV Nuclear Power Plants

    International Nuclear Information System (INIS)

    Miller, Don W.

    2001-01-01

    need for substantial research. As we consider I and C systems in Generation IV reactors, we have the opportunity to take a much less 'timid' design philosophy than was taken in the design of I and C systems in the ALWRs. We need to make use of advanced technology to design an I and C system for the Generation IV multi-unit plant designs currently being considered. Such a design should accomplish the following: 1. provides for multi-unit control; 2. contributes to a plant design objective of a very low core damage frequency; 3. maximizes plant thermal efficiency (>50%); 4. maximizes plant capacity factor (>90%); 5. optimizes operability; 6. maximizes maintainability; 7. provides for on-line monitoring, calibration, and diagnostics; 8. provides optimum response to disturbances; 9. provides excellent load-following capability. When we consider the current situation in operating Generation I and II nuclear power plants and even Generation III ALWR design, we conclude that Generation IV reactors should employ at least Generation II measurement systems. Let us first consider data transmission, which is a form of communication, and ask the question: Do new communication-transferring methods by electrons flow in copper wires? The obvious answer is no. Virtually all new communication systems are using some electromagnetic method, such as light, microwaves, HF or VHF radio signals, and virtually no copper wires. When we envision Generation IV nuclear power plants, we should minimize the use of copper wires for data transmission. We should transmit data primarily by fiber optics and various wireless methods, some of which can penetrate thick barriers. Now let us consider sensors. If we use light for data transmission, then we should also use optical-based sensors. We should also take advantage of microprocessors, which provide opportunities to embed 'intelligence' in the sensor that can be used to increase accuracy, stability, and tolerance to external stressors (i.e., radiation

  14. Diorganotin(IV) Complexes with Methionine Methyl Ester. Equilibria ...

    African Journals Online (AJOL)

    IV) (DBT) and diphenyltin(IV) (DPT) was investigated at 25 °C and 0.1 mol dm–3 ionic strength in water for dimethyltin(IV) and in 50 % dioxane–water mixture for dibutyltin(IV) and diphenyltin(IV). Methionine methyl ester forms1:1 and 1:2 ...

  15. Reactor Physics Training

    International Nuclear Information System (INIS)

    Baeten, P.

    2007-01-01

    University courses in nuclear reactor physics at the universities consist of a theoretical description of the physics and technology of nuclear reactors. In order to demonstrate the basic concepts in reactor physics, training exercises in nuclear reactor installations are also desirable. Since the number of reactor facilities is however strongly decreasing in Europe, it becomes difficult to offer to students a means for demonstrating the basic concepts in reactor physics by performing training exercises in nuclear installations. Universities do not generally possess the capabilities for performing training exercises. Therefore, SCK-CEN offers universities the possibility to perform (on a commercial basis) training exercises at its infrastructure consisting of two research reactors (BR1 and VENUS). Besides the organisation of training exercises in the framework of university courses, SCK-CEN also organizes theoretical courses in reactor physics for the education and training of nuclear reactor operators. It is indeed a very important subject to guarantee the safe operation of present and future nuclear reactors. In this framework, an understanding of the fundamental principles of nuclear reactor physics is also necessary for reactor operators. Therefore, the organisation of a basic Nuclear reactor physics course at the level of reactor operators in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The objectives this activity are: (1) to provide training and education activities in reactor physics for university students and (2) to organise courses in nuclear reactor physics for reactor operators

  16. Periodontal Disease Part IV: Periodontal Infections

    OpenAIRE

    Turnbull, Robert S.

    1988-01-01

    In Part IV of this article, the author describes two periodontal infections, acute necrotizing ulcerative gingivitis (trench mouth) and periodontal abscess, both acute painful conditions for which patients may seek advice from their family physician rather than their dentist.

  17. Safety assessment for Generation IV nuclear systems

    International Nuclear Information System (INIS)

    Leahy, T.J.

    2012-01-01

    The Generation IV International Forum (GIF) Risk and Safety Working Group (RSWG) was created to develop an effective approach for the safety of Generation IV advanced nuclear energy systems. Recent RSWG work has focused on the definition of an integrated safety assessment methodology (ISAM) for evaluating the safety of Generation IV systems. ISAM is an integrated 'tool-kit' consisting of 5 analytical techniques that are available and matched to appropriate stages of Generation IV system concept development: 1) qualitative safety features review - QSR, 2) phenomena identification and ranking table - PIRT, 3) objective provision tree - OPT, 4) deterministic and phenomenological analyses - DPA, and 5) probabilistic safety analysis - PSA. The integrated methodology is intended to yield safety-related insights that help actively drive the evolving design throughout the technology development cycle, potentially resulting in enhanced safety, reduced costs, and shortened development time

  18. Determination of uranium (IV) by flow voltammetry

    International Nuclear Information System (INIS)

    Ding Anqing

    1987-01-01

    According to the quantitative reaction of U(IV) and Fe(III) in H 2 SO 4 as well as the relation between current and concentration of substance detected, U(IV) has been determined indirectly by measurement of the electrolysis current of residual Fe(III). The columniform electrode used is made of glass carbon particles. At the range of U(IV) from a few micrograms to 40 μg, the linear relation is excellent. The relative standard deviation is within ±4%. The interference of Fe(II), Ti(IV) and U(VI) is negligible but of Ti(III) is serious. This method has been successfully applied in the determination of actual samples (both out line and on line). Main advantages of this procedure are rapid, simple, small amount of sample (only at microgram level) and easy to realize automation, able to use for on line or process analysis

  19. IV&V Project Assessment Process Validation

    Science.gov (United States)

    Driskell, Stephen

    2012-01-01

    The Space Launch System (SLS) will launch NASA's Multi-Purpose Crew Vehicle (MPCV). This launch vehicle will provide American launch capability for human exploration and travelling beyond Earth orbit. SLS is designed to be flexible for crew or cargo missions. The first test flight is scheduled for December 2017. The SLS SRR/SDR provided insight into the project development life cycle. NASA IV&V ran the standard Risk Based Assessment and Portfolio Based Risk Assessment to identify analysis tasking for the SLS program. This presentation examines the SLS System Requirements Review/System Definition Review (SRR/SDR), IV&V findings for IV&V process validation correlation to/from the selected IV&V tasking and capabilities. It also provides a reusable IEEE 1012 scorecard for programmatic completeness across the software development life cycle.

  20. Genetics Home Reference: mucopolysaccharidosis type IV

    Science.gov (United States)

    ... enzymes, GAGs accumulate within cells, specifically inside the lysosomes . Lysosomes are compartments in the cell that break down ... that cause molecules to build up inside the lysosomes are called lysosomal storage disorders. In MPS IV, ...