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Sample records for general electric nuclear test reactor

  1. REACTOR AND SHIELD PHYSICS. Comprehensive Technical Report, General Electric Direct-Air-Cycle, Aircraft Nuclear Propulsion Program.

    Energy Technology Data Exchange (ETDEWEB)

    Edwards, W.E.; Simpson, J.D.

    1962-01-01

    This volume is one of twenty-one summarizing the Aircraft Nuclear Propulsion Program of the General Electric Company. This volume describes the experimental and theoretical work accomplished in the areas of reactor and shield physics.

  2. Reactors for nuclear electric propulsion

    Energy Technology Data Exchange (ETDEWEB)

    Buden, D.; Angelo, J.A. Jr.

    1981-01-01

    Propulsion is the key to space exploitation and power is the key to propulsion. This paper examines the role of nuclear fission reactors as the primary power source for high specific impulse electric propulsion systems for space missions of the 1980s and 1990s. Particular mission applications include transfer to and a reusable orbital transfer vehicle from low-Earth orbit to geosynchronous orbit, outer planet exploration and reconnaissance missions, and as a versatile space tug supporting lunar resource development. Nuclear electric propulsion is examined as an indispensable component in space activities of the next two decades.

  3. Dynamic Response Testing in an Electrically Heated Reactor Test Facility

    Science.gov (United States)

    Bragg-Sitton, Shannon M.; Morton, T. J.

    2006-01-01

    Non-nuclear testing can be a valuable tool in development of a space nuclear power or propulsion system. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Standard testing allows one to fully assess thermal, heat transfer, and stress related attributes of a given system, but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. The integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and full nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response characteristics, and assess potential design improvements at a relatively small fiscal investment. Initial system dynamic response testing was demonstrated on the integrated SAFE-100a heat pipe cooled, electrically heated reactor and heat exchanger hardware, utilizing a one-group solution to the point kinetics equations to simulate the expected neutronic response of the system (Bragg-Sitton, 2005). The current paper applies the same testing methodology to a direct drive gas cooled reactor system, demonstrating the applicability of the testing methodology to any reactor type and demonstrating the variation in system response characteristics in different reactor concepts. In each testing application, core power transients were controlled by a point kinetics model with reactivity feedback based on core average temperature; the neutron generation time and the temperature feedback coefficient are provided as model inputs. Although both system designs utilize a fast spectrum reactor, the method of cooling the reactor differs significantly, leading to a variable system response that can be demonstrated and assessed in a non-nuclear test facility.

  4. Dielectric Heaters for Testing Spacecraft Nuclear Reactors

    Science.gov (United States)

    Sims, William Herbert; Bitteker, Leo; Godfroy, Thomas

    2006-01-01

    A document proposes the development of radio-frequency-(RF)-driven dielectric heaters for non-nuclear thermal testing of the cores of nuclear-fission reactors for spacecraft. Like the electrical-resistance heaters used heretofore for such testing, the dielectric heaters would be inserted in the reactors in place of nuclear fuel rods. A typical heater according to the proposal would consist of a rod of lossy dielectric material sized and shaped like a fuel rod and containing an electrically conductive rod along its center line. Exploiting the dielectric loss mechanism that is usually considered a nuisance in other applications, an RF signal, typically at a frequency .50 MHz and an amplitude between 2 and 5 kV, would be applied to the central conductor to heat the dielectric material. The main advantage of the proposal is that the wiring needed for the RF dielectric heating would be simpler and easier to fabricate than is the wiring needed for resistance heating. In some applications, it might be possible to eliminate all heater wiring and, instead, beam the RF heating power into the dielectric rods from external antennas.

  5. Dynamic Response Testing in an Electrically Heated Reactor Test Facility

    Science.gov (United States)

    Bragg-Sitton, Shannon M.; Morton, T. J.

    2006-01-01

    Non-nuclear testing can be a valuable tool in the development of a space nuclear power or propulsion system. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Standard testing allows one to fully assess thermal, heat transfer, and stress related attributes of a given system, but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. The integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and fueled nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response characteristics, and assess potential design improvements at a relatively small fiscal investment. Initial system dynamic response testing was demonstrated on the integrated SAFE-100a heat pipe (HP) cooled, electrically heated reactor and heat exchanger hardware, utilizing a one-group solution to the point kinetics equations to simulate the expected neutronic response of the system. Reactivity feedback calculations were then based on a bulk reactivity feedback coefficient and measured average core temperature. This paper presents preliminary results from similar dynamic testing of a direct drive gas cooled reactor system (DDG), demonstrating the applicability of the testing methodology to any reactor type and demonstrating the variation in system response characteristics in different reactor concepts. Although the HP and DDG designs both utilize a fast spectrum reactor, the method of cooling the reactor differs significantly, leading to a variable system response that can be demonstrated and assessed in a non-nuclear test facility. Planned system upgrades to allow implementation of higher fidelity dynamic testing are also discussed. Proposed DDG

  6. Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hogerton, John

    1964-01-01

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

  7. General Electric Reactor Protection System Unavailability, 1984-1995

    Energy Technology Data Exchange (ETDEWEB)

    C. D. Gentillon; D. Rasmuson (USNRC); H. Hamzehee; M. B. Calley; S. A. Eide; T. Wierman (INEEL)

    1999-08-01

    An analysis was performed of the safety-related performance of the reactor protection system (RPS) at U.S. General Electric commercial reactors during the period 1984 through 1995. RPS operational data were collected from the Nuclear Plant Reliability Data System and Licensee Event Reports. A risk-based analysis was performed on the data to estimate the observed unavailability of the RPS, based on a fault tree model of the system. Results were compared with existing unavailability estimates from Individual Plant Examinations and other reports.

  8. General Electric Reactor Protection System Unavailability, 1984--1995

    Energy Technology Data Exchange (ETDEWEB)

    Eide, Steven Arvid; Calley, Michael Brennan; Gentillon, Cynthia Ann; Wierman, Thomas Edward; Hamzehee, H.; Rasmuson, D.

    1999-08-01

    An analysis was performed of the safety-related performance of the reactor protection system (RPS) at U. S. General Electric commercial reactors during the period 1984 through 1995. RPS operational data were collected from the Nuclear Plant Reliability Data System and Licensee Event Reports. A risk-based analysis was performed on the data to estimate the observed unavailability of the RPS, based on a fault tree model of the system. Results were compared with existing unavailability estimates from Individual Plant Examinations and other reports.

  9. Pellet bed reactor concept for nuclear electric propulsion

    Science.gov (United States)

    El-Genk, Mohamed S.; Morley, Nicholas J.; Juhasz, Albert

    1993-01-01

    For Nuclear Electric Propulsion (NEP) applications, gas cooled nuclear reactors with dynamic energy conversion systems offer high specific power and low total mass. This paper describes the Pellet Bed Reactor (PeBR) concept for potential NEP missions to Mars. The helium cooled, 75-80 MWt PeBR, consists of a single annular fuel region filled with a randomly packed bed of spherical fuel pellets, is designed for multiple starts, and offers unique safety and operation features. Each fuel pellet, about 8-10 mm in diameter, is composed of hundreds of TRISO type fuel microspheres embedded in a graphite matrix for a full retention of fission products. To eliminate the likelihood of a single-point failure, the annular core of the PeBR is divided into three 120° sectors. Each sector is self contained and separate and capable of operating and being cooled on its own and in cooperation with either one or two other sectors. Each sector is coupled to a separate, 5 MWe Closed Brayton Cycle (CBC) energy conversion unit and is subcritical for safe handling and launching. In the event of a failure of the cooling system of a core sector, the reactor power level may be reduced, allowing adjacent sectors to convect the heat away using their own cooling system, thus maintaining reactor operation. Also, due to the absence of an internal core structure in the PeBR core, fueling of the reactor can easily be performed either at the launch facility or in orbit, and refueling can be accomplished in orbit as needed to extend the power system lifetime

  10. Light weight space power reactors for nuclear electric propulsion

    Energy Technology Data Exchange (ETDEWEB)

    Ludewig, H.; Mughabghab, S.; Lazareth, O.; Perkins, K.; Schmidt, E.; Powell, J.R.

    1991-01-01

    A Nuclear Electric Propulsion (NEP) unit capable of propelling a manned vehicle to MARS will be required to have a value of {alpha} (kg/kWe) which is less than five. In order to meet this goal the reactor mass, and thus its contribution to the value of {alpha} will have to be minimized. In this paper a candidate for such a reactor is described. It consists of a gas cooled Particle Bed Reactor (PBR), with specially chosen materials which allow it to operate at an exit temperature of approximately 2000 K. One of the unique features of a PBR is the direct cooling of particulate fuel by the working fluid. This feature allows for high power densities, highest possible gas exit temperatures, for a given fuel temperature and because of the thin particle bed a low pressure drop. The PBR's described in this paper will have a ceramic moderator (Be{sub 2}C), ZrC coated fuel particles and a carbon/carbon hot frit. All the reactors will be designed with sufficient fissile loading to operate at full power for seven years. The burn up possible with particulate fuel is approximately 30%--50%. These rector designs achieve a value of {alpha} less than unity in the power range of interest (5 MWe). 5 refs., 3 figs.

  11. Performance tests for integral reactor nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, Dong-Seong; Yim, Jeong-Sik; Lee, Chong-Tak; Kim, Han-Soo; Koo, Yang-Hyun; Lee, Byung-Ho; Cheon, Jin-Sik; Oh, Je-Yong

    2006-02-15

    An integral type reactor SMART plans to utilize metallic Zr-U fuel which is Zr-based alloy with 34{approx}38 wt% U. In order to verify the technologies for the design and manufacturing of the fuel and get a license, performance tests were carried out. Experimental Fuel Assembly (EFA) manufactured in KAERI is being successfully irradiated in the MIR reactor of RIAR from September 4 2004, and it has achieved burnup of 0.21 g/cc as of January 25 2006. Thermal properties of irradiated Zr-U fuel were measured. Up to the phase transformation temperature, thermal diffusivity increased linearly in proportion to temperature. However its dependence on the burnup was not significant. RIA tests with 4 unirradiated Zr-U fuel rods were performed in Kurchatov Institute to establish a safety criterion. In the case of the un-irradiated Zr-U fuel, the energy deposition during the control rod ejection accident should be less than 172 cal/g to prevent the failure accompanying fuel fragmentation and dispersal. Finally the irradiation tests of fuel rods have been performed at HANARO. The HITE-2 test was successfully completed up to a burnup of 0.31 g/cc. The HITE-3 test began in February 2004 and will be continued up to a target burnup of 0.6 g/cc.

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

    Science.gov (United States)

    Fletcher, C. D.

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

  13. Comprehensive Technical Report, General Electric Direct-Air-Cycle Aircraft Nuclear Propulsion Program; Aircraft Nuclear Propulsion Application Studies

    Energy Technology Data Exchange (ETDEWEB)

    Comassar, S.

    1962-04-30

    This volume is one of twenty-one summarizing the Aircraft Nuclear Propulsion Program of the General Electric Company. This portion describes the studies of advanced applications of nuclear reactors that were performed, including various types of aircraft, missiles, space vehicles, ships, and portable power plants.

  14. Comprehensive Technical Report, General Electric Direct-Air-Cycle Aircraft Nuclear Propulsion Program; Aircraft Nuclear Propulsion Application Studies

    Energy Technology Data Exchange (ETDEWEB)

    Comassar, S.

    1962-04-30

    This volume is one of twenty-one summarizing the Aircraft Nuclear Propulsion Program of the General Electric Company. This portion describes the studies of advanced applications of nuclear reactors that were performed, including various types of aircraft, missiles, space vehicles, ships, and portable power plants.

  15. Nuclear reactor physics

    CERN Document Server

    Stacey, Weston M

    2010-01-01

    Nuclear reactor physics is the core discipline of nuclear engineering. Nuclear reactors now account for a significant portion of the electrical power generated worldwide, and new power reactors with improved fuel cycles are being developed. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. The second edition of this successful comprehensive textbook and reference on basic and advanced nuclear reactor physics has been completely updated, revised and enlarged to include the latest developme

  16. 10 CFR 110.26 - General license for the export of nuclear reactor components.

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false General license for the export of nuclear reactor components. 110.26 Section 110.26 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) EXPORT AND IMPORT OF NUCLEAR EQUIPMENT AND MATERIAL Licenses § 110.26 General license for the export of nuclear...

  17. Testing piezoelectric sensors in a nuclear reactor environment

    Science.gov (United States)

    Reinhardt, Brian T.; Suprock, Andy; Tittmann, Bernhard

    2017-02-01

    Several Department of Energy Office of Nuclear Energy (DOE-NE) programs, such as the Fuel Cycle Research and Development (FCRD), Advanced Reactor Concepts (ARC), Light Water Reactor Sustainability, and Next Generation Nuclear Power Plants (NGNP), are investigating new fuels, materials, and inspection paradigms for advanced and existing reactors. A key objective of such programs is to understand the performance of these fuels and materials during irradiation. In DOE-NE's FCRD program, ultrasonic based technology was identified as a key approach that should be pursued to obtain the high-fidelity, high-accuracy data required to characterize the behavior and performance of new candidate fuels and structural materials during irradiation testing. The radiation, high temperatures, and pressure can limit the available tools and characterization methods. In this work piezoelectric transducers capable of making these measurements are developed. Specifically, three piezoelectric sensors (Bismuth Titanate, Aluminum Nitride, and Zinc Oxide) are tested in the Massachusetts Institute of Technology Research reactor to a fast neutron fluence of 8.65×1020 nf/cm2. It is demonstrated that Bismuth Titanate is capable of transduction up to 5 × 1020 nf/cm2, Zinc Oxide is capable of transduction up to at least 6.27 × 1020 nf/cm2, and Aluminum Nitride is capable of transduction up to at least 8.65 × 1020 nf/cm2.

  18. 压水堆核电站核三级屏蔽电泵试验鉴定程序%Testing of Nuclear Three-stage Shielded Electric Pump for Pressurized Water Reactor Nuclear Power Station

    Institute of Scientific and Technical Information of China (English)

    李作维; 孙悦; 薄大威

    2015-01-01

    This paper analyzes the necessity of obtaining the design/manufacture license of three generation nuclear power plants with two generation of pressurized water reactor nuclear power plant. Describe the function and structure of the three stage pump. Especially introduce the program and test method of the two generation of nuclear power station in three gener-ation pressurized water reactor nuclear power station.%分析了二代加压水堆核电站核三级屏蔽电泵取得民用核安全设备 (屏蔽泵) 设计/制造许可证的必要性, 对核三级屏蔽电泵的功能、 规格及结构进行了详细的阐述, 并重点针对二代加压水堆核电站核三级屏蔽电泵试验鉴定的程序及试验方法进行了介绍, 为核三级屏蔽电泵鉴定提供了依据.

  19. 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.

  20. High-temperature nuclear reactor power plant cycle for hydrogen and electricity production – numerical analysis

    Directory of Open Access Journals (Sweden)

    Dudek Michał

    2016-01-01

    Full Text Available High temperature gas-cooled nuclear reactor (called HTR or HTGR for both electricity generation and hydrogen production is analysed. The HTR reactor because of the relatively high temperature of coolant could be combined with a steam or gas turbine, as well as with the system for heat delivery for high-temperature hydrogen production. However, the current development of HTR’s allows us to consider achievable working temperature up to 750°C. Due to this fact, industrial-scale hydrogen production using copper-chlorine (Cu-Cl thermochemical cycle is considered and compared with high-temperature electrolysis. Presented calculations show and confirm the potential of HTR’s as a future solution for hydrogen production without CO2 emission. Furthermore, integration of a hightemperature nuclear reactor with a combined cycle for electricity and hydrogen production may reach very high efficiency and could possibly lead to a significant decrease of hydrogen production costs.

  1. Radioactive waste shipments to Hanford Retrievable Storage from the General Electric Vallecitos Nuclear Center, Pleasanton, California

    Energy Technology Data Exchange (ETDEWEB)

    Vejvoda, E.J.; Pottmeyer, J.A.; DeLorenzo, D.S.; Weyns-Rollosson, M.I. [Los Alamos Technical Associates, Inc., NM (United States); Duncan, D.R. [Westinghouse Hanford Co., Richland, WA (United States)

    1993-10-01

    During the next two decades the transuranic (TRU) wastes now stored in the burial trenches and storage facilities at the Hanford Site are to be retrieved, processed at the Waste Receiving and Processing Facility, and shipped to the Waste Isolation Pilot Plant near Carlsbad, New Mexico for final disposal. Approximately 3.8% of the TRU waste to be retrieved for shipment to WIPP was generated at the General Electric (GE) Vallecitos Nuclear Center (VNC) in Pleasanton, California and shipped to the Hanford Site for storage. The purpose of this report is to characterize these radioactive solid wastes using process knowledge, existing records, and oral history interviews. The waste was generated almost exclusively from the activities, of the Plutonium Fuels Development Laboratory and the Plutonium Analytical Laboratory. Section 2.0 provides further details of the VNC physical plant, facility operations, facility history, and current status. The solid radioactive wastes were associated with two US Atomic Energy Commission/US Department of Energy reactor programs -- the Fast Ceramic Reactor (FCR) program, and the Fast Flux Test Reactor (FFTR) program. These programs involved the fabrication and testing of fuel assemblies that utilized plutonium in an oxide form. The types and estimated quantities of waste resulting from these programs are discussed in detail in Section 3.0. A detailed discussion of the packaging and handling procedures used for the VNC radioactive wastes shipped to the Hanford Site is provided in Section 4.0. Section 5.0 provides an in-depth look at this waste including the following: weight and volume of the waste, container types and numbers, physical description of the waste, radiological components, hazardous constituents, and current storage/disposal locations.

  2. Space Nuclear Reactor Engineering

    Energy Technology Data Exchange (ETDEWEB)

    Poston, David Irvin [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-03-06

    We needed to find a space reactor concept that could be attractive to NASA for flight and proven with a rapid turnaround, low-cost nuclear test. Heat-pipe-cooled reactors coupled to Stirling engines long identified as the easiest path to near-term, low-cost concept.

  3. Multi-reactor power system configurations for multimegawatt nuclear electric propulsion

    Science.gov (United States)

    George, Jeffrey A.

    1991-01-01

    A modular, multi-reactor power system and vehicle configuration for piloted nuclear electric propulsion (NEP) missions to Mars is presented. Such a design could provide enhanced system and mission reliability, allowing a comfortable safety margin for early manned flights, and would allow a range of piloted and cargo missions to be performed with a single power system design. Early use of common power modules for cargo missions would also provide progressive flight experience and validation of standardized systems for use in later piloted applications. System and mission analysis are presented to compare single and multi-reactor configurations for piloted Mars missions. A conceptual design for the Hydra modular multi-reactor NEP vehicle is presented.

  4. Space Molten Salt Reactor Concept for Nuclear Electric Propulsion and Surface Power

    Science.gov (United States)

    Eades, M.; Flanders, J.; McMurray, N.; Denning, R.; Sun, X.; Windl, W.; Blue, T.

    Students at The Ohio State University working under the NASA Steckler Grant sought to investigate how molten salt reactors with fissile material dissolved in a liquid fuel medium can be applied to space applications. Molten salt reactors of this kind, built for non-space applications, have demonstrated high power densities, high temperature operation without pressurization, high fuel burn up and other characteristics that are ideal for space fission systems. However, little research has been published on the application of molten salt reactor technology to space fission systems. This paper presents a conceptual design of the Space Molten Salt Reactor (SMSR), which utilizes molten salt reactor technology for Nuclear Electric Propulsion (NEP) and surface power at the 100 kWe to 15 MWe level. Central to the SMSR design is a liquid mixture of LiF, BeF2 and highly enriched U235F4 that acts as both fuel and core coolant. In brief, some of the positive characteristics of the SMSR are compact size, simplified core design, high fuel burn up percentages, proliferation resistant features, passive safety mechanisms, a considerable body of previous research, and the possibility for flexible mission architecture.

  5. Subsize specimen testing of nuclear reactor pressure vessel material

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, A.S. [Missouri Univ., Rolla, MO (United States). Materials Research Center; Rosinski, S.T. [Sandia National Labs., Albuquerque, NM (United States); Cannon, N.S. [Westinghouse Hanford Co., Richland, WA (United States); Hamilton, M.L. [Pacific Northwest Lab., Richland, WA (United States)

    1991-12-31

    A new methodology is proposed to correlate the upper shelf energy (USE) of full size and subsize Charpy specimens of a nuclear reactor pressure vessel plate material, A533B. The methodology appears to be more satisfactory than the methodologies proposed earlier. USE of a notched-only specimen is partitioned into macro-crack initiation and crack propagation energies. USE of a notched and precracked specimen provides the crack propagation energy. {Delta}USE, the difference between the USE`s of notched-only and precracked specimens, is an estimate of the crack initiation energy. {Delta}USE was normalized by a factor involving the dimensions of the Charpy specimen and the stress concentration factor at the notch root. The normalized values of the {Delta}USE were found to be invariant with specimen size.

  6. Subsize specimen testing of nuclear reactor pressure vessel material

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, A.S. (Missouri Univ., Rolla, MO (United States). Materials Research Center); Rosinski, S.T. (Sandia National Labs., Albuquerque, NM (United States)); Cannon, N.S. (Westinghouse Hanford Co., Richland, WA (United States)); Hamilton, M.L. (Pacific Northwest Lab., Richland, WA (United States))

    1991-01-01

    A new methodology is proposed to correlate the upper shelf energy (USE) of full size and subsize Charpy specimens of a nuclear reactor pressure vessel plate material, A533B. The methodology appears to be more satisfactory than the methodologies proposed earlier. USE of a notched-only specimen is partitioned into macro-crack initiation and crack propagation energies. USE of a notched and precracked specimen provides the crack propagation energy. [Delta]USE, the difference between the USE's of notched-only and precracked specimens, is an estimate of the crack initiation energy. [Delta]USE was normalized by a factor involving the dimensions of the Charpy specimen and the stress concentration factor at the notch root. The normalized values of the [Delta]USE were found to be invariant with specimen size.

  7. A pilot level decision analysis of thermionic reactor development strategy for nuclear electric propulsion

    Science.gov (United States)

    Menke, M. M.; Judd, B. R.

    1973-01-01

    The development policy for thermionic reactors to provide electric propulsion and power for space exploration was analyzed to develop a logical procedure for selecting development alternatives that reflect the technical feasibility, JPL/NASA project objectives, and the economic environment of the project. The partial evolution of a decision model from the underlying philosophy of decision analysis to a deterministic pilot phase is presented, and the general manner in which this decision model can be employed to examine propulsion development alternatives is illustrated.

  8. Electric Power quality Analysis in research reactor: Impacts on nuclear safety assessment and electrical distribution reliability

    Energy Technology Data Exchange (ETDEWEB)

    Touati, Said; Chennai, Salim; Souli, Aissa [Nuclear Research Centre of Birine, Ain Oussera, Djelfa Province (Algeria)

    2015-07-01

    The increased requirements on supervision, control, and performance in modern power systems make power quality monitoring a common practise for utilities. Large databases are created and automatic processing of the data is required for fast and effective use of the available information. Aim of the work presented in this paper is the development of tools for analysis of monitoring power quality data and in particular measurements of voltage and currents in various level of electrical power distribution. The study is extended to evaluate the reliability of the electrical system in nuclear plant. Power Quality is a measure of how well a system supports reliable operation of its loads. A power disturbance or event can involve voltage, current, or frequency. Power disturbances can originate in consumer power systems, consumer loads, or the utility. The effect of power quality problems is the loss power supply leading to severe damage to equipments. So, we try to track and improve system reliability. The assessment can be focused on the study of impact of short circuits on the system, harmonics distortion, power factor improvement and effects of transient disturbances on the Electrical System during motor starting and power system fault conditions. We focus also on the review of the Electrical System design against the Nuclear Directorate Safety Assessment principles, including those extended during the last Fukushima nuclear accident. The simplified configuration of the required system can be extended from this simple scheme. To achieve these studies, we have used a demo ETAP power station software for several simulations. (authors)

  9. Modeling and Testing of Non-Nuclear, Highpower Simulated Nuclear Thermal Rocket Reactor Elements

    Science.gov (United States)

    Kirk, Daniel R.

    2005-01-01

    When the President offered his new vision for space exploration in January of 2004, he said, "Our third goal is to return to the moon by 2020, as the launching point for missions beyond," and, "With the experience and knowledge gained on the moon, we will then be ready to take the next steps of space exploration: human missions to Mars and to worlds beyond." A human mission to Mars implies the need to move large payloads as rapidly as possible, in an efficient and cost-effective manner. Furthermore, with the scientific advancements possible with Project Prometheus and its Jupiter Icy Moons Orbiter (JIMO), (these use electric propulsion), there is a renewed interest in deep space exploration propulsion systems. According to many mission analyses, nuclear thermal propulsion (NTP), with its relatively high thrust and high specific impulse, is a serious candidate for such missions. Nuclear rockets utilize fission energy to heat a reactor core to very high temperatures. Hydrogen gas flowing through the core then becomes superheated and exits the engine at very high exhaust velocities. The combination of temperature and low molecular weight results in an engine with specific impulses above 900 seconds. This is almost twice the performance of the LOX/LH2 space shuttle engines, and the impact of this performance would be to reduce the trip time of a manned Mars mission from the 2.5 years, possible with chemical engines, to about 12-14 months.

  10. A design study of reactor core optimization for direct nuclear heat-to-electricity conversion in a space power reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yoshikawa, Hidekazu; Takahashi, Makoto; Shimoda, Hiroshi; Takeoka, Satoshi [Kyoto Univ. (Japan); Nakagawa, Masayuki; Kugo, Teruhiko

    1998-01-01

    To propose a new design concept of a nuclear reactor used in the space, research has been conducted on the conceptual design of a new nuclear reactor on the basis of the following three main concepts: (1) Thermionic generation by thermionic fuel elements (TFE), (2) reactivity control by rotary reflector, and (3) reactor cooling by liquid metal. The outcomes of the research are: (1) A calculation algorithm was derived for obtaining convergent conditions by repeating nuclear characteristic calculation and thermal flow characteristic calculation for the space nuclear reactor. (2) Use of this algorithm and the parametric study established that a space nuclear reactor using 97% enriched uranium nitride as the fuel and lithium as the coolant and having a core with a radius of about 25 cm, a height of about 50 cm and a generation efficiency of about 7% can probably be operated continuously for at least more than ten years at 100 kW only by reactivity control by rotary reflector. (3) A new CAD/CAE system was developed to assist design work to optimize the core characteristics of the space nuclear reactor comprehensively. It is composed of the integrated design support system VINDS using virtual reality and the distributed system WINDS to collaboratively support design work using Internet. (N.H.)

  11. Nuclear reactor neutron shielding

    Energy Technology Data Exchange (ETDEWEB)

    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.

  12. Ultrahigh temperature vapor core reactor-MHD system for space nuclear electric power

    Science.gov (United States)

    Maya, Isaac; Anghaie, Samim; Diaz, Nils J.; Dugan, Edward T.

    1991-01-01

    The conceptual design of a nuclear space power system based on the ultrahigh temperature vapor core reactor with MHD energy conversion is presented. This UF4 fueled gas core cavity reactor operates at 4000 K maximum core temperature and 40 atm. Materials experiments, conducted with UF4 up to 2200 K, demonstrate acceptable compatibility with tungsten-molybdenum-, and carbon-based materials. The supporting nuclear, heat transfer, fluid flow and MHD analysis, and fissioning plasma physics experiments are also discussed.

  13. Thermal-hydraulic analysis of nuclear reactors

    CERN Document Server

    Zohuri, Bahman

    2015-01-01

    This text covers the fundamentals of thermodynamics required to understand electrical power generation systems and the application of these principles to nuclear reactor power plant systems. It is not a traditional general thermodynamics text, per se, but a practical thermodynamics volume intended to explain the fundamentals and apply them to the challenges facing actual nuclear power plants systems, where thermal hydraulics comes to play.  Written in a lucid, straight-forward style while retaining scientific rigor, the content is accessible to upper division undergraduate students and aimed at practicing engineers in nuclear power facilities and engineering scientists and technicians in industry, academic research groups, and national laboratories. The book is also a valuable resource for students and faculty in various engineering programs concerned with nuclear reactors. This book also: Provides extensive coverage of thermal hydraulics with thermodynamics in nuclear reactors, beginning with fundamental ...

  14. Multimegawatt nuclear electric propulsion with gaseous and vapor core reactors with MHD

    Science.gov (United States)

    Knight, Travis; Anghaie, Samim; Smith, Blair; Houts, Michael

    2001-02-01

    This study investigated the development of a system concept for space power generation and nuclear electric propulsion based on a fissioning plasma core reactor (FPCR) with magnetohydrodynamic (MHD) power conversion system, coupled to a magnetoplasmadynamic (MPD) thruster. The FPCR is a liquid-vapor core reactor concept operating with metallic uranium or uranium tetrafluoride (UF4) vapor as the fissioning fuel and alkali metals or their fluorides as working fluid in a closed Rankine cycle with MHD energy conversion. Candidate working fluids include K, Li, Na, KF, LiF, NaF, etc. The system features core outlet temperatures of 3000 to 4000 K at pressures of about 1 to 10 MPa, MHD temperatures of 2000 to 3000 K, and radiator temperatures of 1200 to 2000 K. This combination of parameters offers the potential for low total system specific mass in the range of 0.4 to 0.6 kg/kWe. The MHD output could be coupled with minimal power conditioning to the variable specific impulse magnetoplasma rocket (VASIMR), MPD thrusters or other types of thruster for producing thrust at very high specific impulse (Isp=1500 to 10,000 s). .

  15. Analytical evaluation on loss of off-side electric power simulation of the High Temperature Engineering Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Takeda, Takeshi; Nakagawa, Shigeaki; Tachibana, Yukio; Takada, Eiji; Kunitomi, Kazuhiko [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    2000-03-01

    A rise-to-power test of the high temperature engineering test reactor (HTTR) started on September 28 in 1999 for establishing and upgrading the technological basis for the high temperature gas-cooled reactor (HTGR). A loss of off-site electric power test of the HTTR from the normal operation under 15 and 30 MW thermal power will be carried out in the rise-to-power test. Analytical evaluations on transient behaviors of the reactor and plant during the loss of off-site electric power were conducted. These estimations are proposed as benchmark problems for the IAEA coordinated research program on 'Evaluation of HTGR Performance'. This report describes an event scenario of transient during the loss of off-site electric power, the outline of major components and system, detailed thermal and nuclear data set for these problems and pre-estimation results of the benchmark problems by an analytical code 'ACCORD' for incore and plant dynamics of the HTGR. (author)

  16. Nuclear Rocket Engine Reactor

    CERN Document Server

    Lanin, Anatoly

    2013-01-01

    The development of a nuclear rocket engine reactor (NRER ) is presented in this book. The working capacity of an active zone NRER under mechanical and thermal load, intensive neutron fluxes, high energy generation (up to 30 MBT/l) in a working medium (hydrogen) at temperatures up to 3100 K is displayed. Design principles and bearing capacity of reactors area discussed on the basis of simulation experiments and test data of a prototype reactor. Property data of dense constructional, porous thermal insulating and fuel materials like carbide and uranium carbide compounds in the temperatures interval 300 - 3000 K are presented. Technological aspects of strength and thermal strength resistance of materials are considered. The design procedure of possible emergency processes in the NRER is developed and risks for their origination are evaluated. Prospects of the NRER development for pilotless space devices and piloted interplanetary ships are viewed.

  17. The Next Generation Nuclear Plant Graphite Creep Experiment Irradiation in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Blaine Grover

    2010-10-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Program will be irradiating six gas reactor graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These graphite irradiations are being accomplished to support development of the next generation reactors in the United States. The graphite experiments will be irradiated over the next six to eight years to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data, including irradiation creep, at different temperatures and loading conditions to support design of the Next Generation Nuclear Plant (NGNP) Very High Temperature Gas Reactor, as well as other future gas reactors. The experiments will each consist of a single capsule that will contain six stacks of graphite specimens, with half of the graphite specimens in each stack under a compressive load, while the other half of the specimens will not be subjected to a compressive load during irradiation. The six stacks will have differing compressive loads applied to the top half of each pair of specimen stacks, while a seventh stack will not have a compressive load. The specimens will be irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There will also be the capability of sampling the sweep gas effluent to determine if any oxidation or off-gassing of the specimens occurs during initial start-up of

  18. Acoustic emission monitoring of hot functional testing: Watts Bar Unit 1 Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hutton, P.H.; Dawson, J.F.; Friesel, M.A.; Harris, J.C.; Pappas, R.A.

    1984-06-01

    Acoustic emission (AE) monitoring of selected pressure boundary areas at TVA's Watts Bar, Unit 1 Nuclear Power Plant during hot functional preservice testing is described in this report. The report deals with background, methodology, and results. The work discussed here is a major milestone in a program supported by NRC to develop and demonstrate application of AE monitoring for continuous surveillance of reactor pressure boundaries to detect and evaluate growing flaws. The subject work demonstrated that anticipated problem areas can be overcome. Work is continuing toward AE monitoring during reactor operation.

  19. In-reactor tests of the nuclear light bulb rocket concept

    Science.gov (United States)

    Gauntt, R. O.; Slutz, S. A.; Latham, T. S.; Roman, W. C.; Rogers, R. J.

    1992-07-01

    An overview is given of the closed-cycle Gas Core Nuclear Rocket outlining scenarios for its use in short-duration Mars missions and results of Nuclear Light Bulb (NLB) tests. Isothermal and nonnuclear tests are described which confirmed the fundamental concepts behind the NLB. NLB reference-engine performance characteristics are given for hypothetical engines that could be used for manned Mars missions. Vehicle/propulsion sizing is based on a Mars mission with three trans-Mars impulse burns, capture and escape burns, and a total mission duration of 600 days. The engine would have a specific impulse of 1870 seconds, a 412-kN thrust, and a thrust/weight ratio of 1.3. Reactor tests including small-scale in-reactor tests are shown to be prerequisites for studying: (1) fluid mechanical confinement of the gaseous nuclear fuel; (2) buffer gas separation and circulation; and (3) the minimization of transparent wall-heat loading. The reactor tests are shown to be critical for establishing the feasibility of the NLB concept.

  20. Advanced Demonstration and Test Reactor Options Study

    Energy Technology Data Exchange (ETDEWEB)

    Petti, David Andrew [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hill, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Gehin, J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States); Heidet, F. [Argonne National Lab. (ANL), Argonne, IL (United States); Kinsey, J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Grandy, Christopher [Argonne National Lab. (ANL), Argonne, IL (United States); Qualls, A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hoffman, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Croson, D. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-01-01

    Global efforts to address climate change will require large-scale decarbonization of energy production in the United States and elsewhere. Nuclear power already provides 20% of electricity production in the United States (U.S.) and is increasing in countries undergoing rapid growth around the world. Because reliable, grid-stabilizing, low emission electricity generation, energy security, and energy resource diversity will be increasingly valued, nuclear power’s share of electricity production has a potential to grow. In addition, there are non electricity applications (e.g., process heat, desalination, hydrogen production) that could be better served by advanced nuclear systems. Thus, the timely development, demonstration, and commercialization of advanced nuclear reactors could diversify the nuclear technologies available and offer attractive technology options to expand the impact of nuclear energy for electricity generation and non-electricity missions. The purpose of this planning study is to provide transparent and defensible technology options for a test and/or demonstration reactor(s) to be built to support public policy, innovation and long term commercialization within the context of the Department of Energy’s (DOE’s) broader commitment to pursuing an “all of the above” clean energy strategy and associated time lines. This planning study includes identification of the key features and timing needed for advanced test or demonstration reactors to support research, development, and technology demonstration leading to the commercialization of power plants built upon these advanced reactor platforms. This planning study is consistent with the Congressional language contained within the fiscal year 2015 appropriation that directed the DOE to conduct a planning study to evaluate “advanced reactor technology options, capabilities, and requirements within the context of national needs and public policy to support innovation in nuclear energy

  1. Proving test on the seismic reliability of nuclear power plant: PWR reactor containment vessel

    Energy Technology Data Exchange (ETDEWEB)

    Akiyama, Hiroshi; Yoshikawa, Teiichi; Ohno, Tokue; Yoshikawa, Eiji.

    1989-01-01

    Seismic reliability proving tests of nuclear power plant facilities are carried out by the Nuclear Power Engineering Test Center, using the large-scale, high-performance vibration table of Tadotsu Engineering Laboratory, and sponsored by the Ministry of International Trade and Industry. In 1982, the seismic reliability proving test of a PWR containment vessel was conducted using a test component of reduced scale 1/3.7. As a result of this test, the test component proved to have structural soundness against earthquakes, and at the same time its stable function was proved by leak tests which were carried out before and after the vibration test. In 1983, the detailed analysis and evaluation of these test results were carried out, and the analysis methods for evaluating strength against earthquakes were established. The seismic analysis and evaluation on the actual containment vessel were then performed using these analysis methods, and the safety and reliability of the PWR reactor containment vessel were confirmed.

  2. Application of a Virtual Reactivity Feedback Control Loop in Non-Nuclear Testing of a Fast Spectrum Reactor

    Science.gov (United States)

    Bragg-Sitton, Shannon M.; Forsbacka, Matthew

    2004-01-01

    For a compact, fast-spectrum reactor, reactivity feedback is dominated by core deformation at elevated temperature. Given the use of accurate deformation measurement techniques, it is possible to simulate nuclear feedback in non-nuclear electrically heated reactor tests. Implementation of simulated reactivity feedback in response to measured deflection is being tested at the NASA Marshall Space Flight Center Early Flight Fission Test Facility (EFF-TF). During tests of the SAFE-100 reactor prototype, core deflection was monitored using a high resolution camera. "virtual" reactivity feedback was accomplished by applying the results of Monte Carlo calculations (MCNPX) to core deflection measurements; the computational analysis was used to establish the reactivity worth of van'ous core deformations. The power delivered to the SAFE-100 prototype was then dusted accordingly via kinetics calculations, The work presented in this paper will demonstrate virtual reactivity feedback as core power was increased from 1 kilowatt(sub t), to 10 kilowatts(sub t), held approximately constant at 10 kilowatts (sub t), and then allowed to decrease based on the negative thermal reactivity coefficient.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1982-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1989-12-01

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

  5. Electrically Heated Testing of the Kilowatt Reactor Using Stirling Technology (KRUSTY) Experiment Using a Depleted Uranium Core

    Science.gov (United States)

    Briggs, Maxwell H.; Gibson, Marc A.; Sanzi, James

    2017-01-01

    The Kilopower project aims to develop and demonstrate scalable fission-based power technology for systems capable of delivering 110 kW of electric power with a specific power ranging from 2.5 - 6.5 Wkg. This technology could enable high power science missions or could be used to provide surface power for manned missions to the Moon or Mars. NASA has partnered with the Department of Energys National Nuclear Security Administration, Los Alamos National Labs, and Y-12 National Security Complex to develop and test a prototypic reactor and power system using existing facilities and infrastructure. This technology demonstration, referred to as the Kilowatt Reactor Using Stirling TechnologY (KRUSTY), will undergo nuclear ground testing in the summer of 2017 at the Nevada Test Site. The 1 kWe variation of the Kilopower system was chosen for the KRUSTY demonstration. The concept for the 1 kWe flight system consist of a 4 kWt highly enriched Uranium-Molybdenum reactor operating at 800 degrees Celsius coupled to sodium heat pipes. The heat pipes deliver heat to the hot ends of eight 125 W Stirling convertors producing a net electrical output of 1 kW. Waste heat is rejected using titanium-water heat pipes coupled to carbon composite radiator panels. The KRUSTY test, based on this design, uses a prototypic highly enriched uranium-molybdenum core coupled to prototypic sodium heat pipes. The heat pipes transfer heat to two Advanced Stirling Convertors (ASC-E2s) and six thermal simulators, which simulate the thermal draw of full scale power conversion units. Thermal simulators and Stirling engines are gas cooled. The most recent project milestone was the completion of non-nuclear system level testing using an electrically heated depleted uranium (non-fissioning) reactor core simulator. System level testing at the Glenn Research Center (GRC) has validated performance predictions and has demonstrated system level operation and control in a test configuration that replicates the one

  6. CALMOS: Innovative device for the measurement of nuclear heating in material testing reactors

    Energy Technology Data Exchange (ETDEWEB)

    Carcreff, H. [Alternative Energies and Atomic Energy Commission CEA, Saclay Center, DEN/DANS/DRSN/SIREN, Gif Sur Yvette, 91191 (France)

    2011-07-01

    An R and D program has been carried out since 2002 in order to improve gamma heating measurements in the 70 MWth OSIRIS Material Testing Reactor operated by CEA's Nuclear Energy Div. at the Saclay research center. Throughout this program an innovative calorimetric probe associated to a specific handling system has been designed in order to make measurements both along the fissile height and on the upper part of the core, where nuclear heating rates still remain high. Two mock-ups of the probe were manufactured and tested in 2005 and 2009 in ex-core area of OSIRIS reactor for the process validation, while a displacement system has been especially designed to move the probe axially. A final probe has been designed thanks to modeling results and to preliminary measurements obtained with mock-ups irradiated to a heating level of 2W/g, This paper gives an overview of the development, describes the calorimetric probe, and expected advantages such as the possibility to use complementary methods to get the nuclear heating measurement. Results obtained with mock-ups irradiated in ex-core area of the reactor are presented and discussed. (authors)

  7. Comprehensive Technical Report, General Electric Direct-Air-Cycle Aircraft Nuclear Propulsion Program, Program Summary and References

    Energy Technology Data Exchange (ETDEWEB)

    Thornton, G.; Rothstein, A.J.

    1962-06-28

    This is one of twenty-one volumes sumarizing the Aircraft Nuclear Propulsion Program of the General Electric Company. This volume discusses the background to the General Electric program, and summarizes the various direct-air-cycle nuclear test assemblies and power plants that were developed. Because of the requirements of high performance, low weight, and small size, vast improvements in existing technology were required to meet the flight objectives. The technological progress achieved during the program is also summarized. The last appendix contains a compilation of the abstracts, tables of contents, and reference lists of the other twenty volumes.

  8. Nuclear safety considerations in the conceptual design of a fast reactor for space electric power and propulsion

    Science.gov (United States)

    Hsieh, T.-M.; Koenig, D. R.

    1977-01-01

    Some nuclear safety aspects of a 3.2 mWt heat pipe cooled fast reactor with out-of-core thermionic converters are discussed. Safety related characteristics of the design including a thin layer of B4C surrounding the core, the use of heat pipes and BeO reflector assembly, the elimination of fuel element bowing, etc., are highlighted. Potential supercriticality hazards and countermeasures are considered. Impacts of some safety guidelines of space transportation system are also briefly discussed, since the currently developing space shuttle would be used as the primary launch vehicle for the nuclear electric propulsion spacecraft.

  9. Monte Carlo Simulation Study of a Differential Calorimeter Measuring the Nuclear Heating in Material Testing Reactors

    Science.gov (United States)

    Amharrak, H.; Reynard-Carette, C.; Lyoussi, A.; Carette, M.; Brun, J.; De Vita, C.; Fourmentel, D.; Villard, J.-F.; Guimbal, P.

    2016-02-01

    The nuclear heating measurements in Material Testing Reactors (MTRs) are crucial for the study of nuclear materials and fuels under irradiation. The reference measurements of this nuclear heating are especially performed by a differential calorimeter including a graphite sample material. Then these measurements are used for other materials, other geometries, or other experimental conditions in order to predict the nuclear heating and thermal conditions induced in the irradiation devices. This paper will present new simulations with MCNP Monte-Carlo transport code to determine the gamma heating profile inside the calorimeter. The whole complex geometry of the sensor has been considered. We use as an input source in the model, the photon spectra calculated in various positions of CARMEN-1 irradiation program in OSIRIS reactor. After a description of the differential calorimeter device, the MCNP modeling used for the calculations of radial profile of nuclear heating inside the calorimeter elements will be introduced. The obtained results of different simulations will be detailed and discussed in this paper. The charged particle equilibrium inside the calorimeter elements will be studied. Then we will focus on parametric studies of the various components of the calorimeter. The influence of source type will be also took into account. Moreover the influence of the material used for the sample will be described.

  10. Monte Carlo Simulation Study of a Differential Calorimeter Measuring the Nuclear Heating in Material Testing Reactors

    Directory of Open Access Journals (Sweden)

    Amharrak H.

    2016-01-01

    Full Text Available The nuclear heating measurements in Material Testing Reactors (MTRs are crucial for the study of nuclear materials and fuels under irradiation. The reference measurements of this nuclear heating are especially performed by a differential calorimeter including a graphite sample material. Then these measurements are used for other materials, other geometries, or other experimental conditions in order to predict the nuclear heating and thermal conditions induced in the irradiation devices. This paper will present new simulations with MCNP Monte-Carlo transport code to determine the gamma heating profile inside the calorimeter. The whole complex geometry of the sensor has been considered. We use as an input source in the model, the photon spectra calculated in various positions of CARMEN-1 irradiation program in OSIRIS reactor. After a description of the differential calorimeter device, the MCNP modeling used for the calculations of radial profile of nuclear heating inside the calorimeter elements will be introduced. The obtained results of different simulations will be detailed and discussed in this paper. The charged particle equilibrium inside the calorimeter elements will be studied. Then we will focus on parametric studies of the various components of the calorimeter. The influence of source type will be also took into account. Moreover the influence of the material used for the sample will be described.

  11. Random processes in nuclear reactors

    CERN Document Server

    Williams, M M R

    1974-01-01

    Random Processes in Nuclear Reactors describes the problems that a nuclear engineer may meet which involve random fluctuations and sets out in detail how they may be interpreted in terms of various models of the reactor system. Chapters set out to discuss topics on the origins of random processes and sources; the general technique to zero-power problems and bring out the basic effect of fission, and fluctuations in the lifetime of neutrons, on the measured response; the interpretation of power reactor noise; and associated problems connected with mechanical, hydraulic and thermal noise sources

  12. Bayesian Zero-Failure (BAZE) reliability demonstration testing procedure for components of nuclear reactor safety systems

    Energy Technology Data Exchange (ETDEWEB)

    Waller, R.A.

    1977-06-01

    A Bayesian-Zero-Failure (BAZE) reliability demonstration testing procedure is presented. The method is developed for an exponential failure-time model and a gamma prior distribution on the failure-rate. A simple graphical approach using percentiles is used to fit the prior distribution. The procedure is given in an easily applied step-by-step form which does not require the use of a computer for its implementation. The BAZE approach is used to obtain sample test plans for selected components of nuclear reactor safety systems.

  13. Dynamic parameters test of Haiyang Nuclear Power Engineering in reactor areas, China

    Science.gov (United States)

    Zhou, N.; Zhao, S.; Sun, L.

    2012-12-01

    Haiyang Nuclear Power Project is located in Haiyang city, China. It consists of 6×1000MW AP1000 Nuclear Power generator sets. The dynamic parameters of the rockmass are essential for the design of the nuclear power plant. No.1 and No.2 reactor area are taken as research target in this paper. Sonic logging, single hole and cross-hole wave velocity are carried out respectively on the site. There are four types of rock lithology within the measured depth. They are siltstone, fine sandstone, shale and allgovite. The total depth of sonic logging is 409.8m and 2049 test points. The sound wave velocity of the rocks are respectively 5521 m/s, 5576m/s, 5318 m/s and 5576 m/s. Accroding to the statistic data, among medium weathered fine sandstone, fairly broken is majority, broken and relatively integrity are second, part of integrity. Medium weathered siltstone, relatively integrity is mojority, fairly broken is second. Medium weathered shale, fairly broken is majority, broken and relatively integrity for the next and part of integrity. Slight weathered fine sandstone, siltstone, shale and allgovite, integrity is the mojority, relatively integrity for the next, part of fairly broken.The single hole wave velocity tests are set in two boreholesin No.1 reactor area and No.2 reactor area respectively. The test depths of two holes are 2-24m, and the others are 2-40m. The wave velocity data are calculated at different depth in each holes and dynamic parameters. According to the test statistic data, the wave velocity and the dynamic parameter values of rockmass are distinctly influenced by the weathering degree. The test results are list in table 1. 3 groups of cross hole wave velocity tests are set for No.1 and 2 reactor area, No.1 reactor area: B16, B16-1, B20(Direction:175°, depth: 100m); B10, B10-1, B11(269°, 40m); B21, B21-1, B17(154°, 40m); with HB16, HB10, HB21 as trigger holes; No.2 reactor area: B47, B47-1, HB51(176°, 100m); B40, B40-1, B41(272°, 40m); B42, B42-1, B

  14. Review of Byzantine General Problems in the Reactor Protection System of Korean Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Eungse; Kim, Yungoo [Korea Hydro and Nuclear Power Co. Ltd. Central Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    This kind of complex error scenario is known as a Byzantine General Problem and the error is called a Byzantine Error (BE). The BE is now considered as one of the plausible common-cause failure in the nuclear power plant's (NPP) computer systems. The Reactor Protection System (RPS) in the Korean NPP consists of multiple redundant digital computer systems to increase system availability and redundancy. This system architecture is inherited form the well proved analog system's architecture. Failure modes and effects on the RPS system functions are reviewed when a BE assumed in the system's decision making logic path. For this review, one channel and two channels of BE problems in the RPS trip function are considered. If a BE occurs in any one channel of the RPS, the systems trip function has no harm affects from the BE. If two BEs occur in any channels of the RPS, the systems trip function may or may not work properly. This BE review method can be applied to other decision-making parts of the protection system in NPP.

  15. Verifying seismic design of nuclear reactors by testing. Volume 1: test plan

    Energy Technology Data Exchange (ETDEWEB)

    1979-07-20

    This document sets forth recommendations for a verification program to test the ability of operational nuclear power plants to achieve safe shutdown immediately following a safe-shutdown earthquake. The purpose of the study is to develop a program plan to provide assurance by physical demonstration that nuclear power plants are earthquake resistant and to allow nuclear power plant operators to (1) decide whether tests should be conducted on their facilities, (2) specify the tests that should be performed, and (3) estimate the cost of the effort to complete the recommended test program.

  16. The outlook for application of powerful nuclear thermionic reactor - powered space electric jet propulsion engines

    Energy Technology Data Exchange (ETDEWEB)

    Semyonov, Y.P.; Bakanov, Y.A.; Synyavsky, V.V.; Yuditsky, V.D. [Rocket-Space Corp. `Energia`, Moscow (Russian Federation)

    1997-12-31

    This paper summarizes main study results for application of powerful space electric jet propulsion unit (EJPUs) which is powered by Nuclear Thermionic Power Unit (NTPU). They are combined in Nuclear Power/Propulsion Unit (NPPU) which serves as means of spacecraft equipment power supply and spacecraft movement. Problems the paper deals with are the following: information satellites delivery and their on-orbit power supply during 10-15 years, removal of especially hazardous nuclear wastes, mining of asteroid resources and others. Evaluations on power/time/mass relationship for this type of mission are given. EJPU parameters are compatible with Russian existent or being under development launch vehicle. (author)

  17. Report on emergency electrical power supply systems for nuclear fuel cycle and reactor facilities security systems

    Energy Technology Data Exchange (ETDEWEB)

    1977-01-01

    The report includes information that will be useful to those responsible for the planning, design and implementation of emergency electric power systems for physical security and special nuclear materials accountability systems. Basic considerations for establishing the system requirements for emergency electric power for security and accountability operations are presented. Methods of supplying emergency power that are available at present and methods predicted to be available in the future are discussed. The characteristics of capacity, cost, safety, reliability and environmental and physical facility considerations of emergency electric power techniques are presented. The report includes basic considerations for the development of a system concept and the preparation of a detailed system design.

  18. Nuclear reactor PBMR and cogeneration; Reactor nuclear PBMR y cogeneracion

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez S, J. R.; Alonso V, G., E-mail: ramon.ramirez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2013-10-15

    In recent years the nuclear reactor designs for the electricity generation have increased their costs, so that at the moment costs are managed of around the 5000 US D for installed kw, reason for which a big nuclear plant requires of investments of the order of billions of dollars, the designed reactors as modular of low power seek to lighten the initial investment of a big reactor dividing the power in parts and dividing in modules the components to lower the production costs, this way it can begin to build a module and finished this to build other, differing the long term investment, getting less risk therefore in the investment. On the other hand the reactors of low power can be very useful in regions where is difficult to have access to the electric net being able to take advantage of the thermal energy of the reactor to feed other processes like the water desalination or the vapor generation for the processes industry like the petrochemical, or even more the possible hydrogen production to be used as fuel. In this work the possibility to generate vapor of high quality for the petrochemical industry is described using a spheres bed reactor of high temperature. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1982-03-01

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

  20. Non-Nuclear Testing of Compact Reactor Technologies at NASA MSFC

    Science.gov (United States)

    Houts, Michael G.; Pearson, J. Boise; Godfroy, Thomas J.

    2011-01-01

    Safe, reliable, compact, autonomous, long-life fission systems have numerous potential applications, both terrestrially and in space. Technologies and facilities developed in support of these systems could be useful to a variety of concepts. At moderate power levels, fission systems can be designed to operate for decades without the need for refueling. In addition, fast neutron damage to cladding and structural materials can be maintained at an acceptable level. Nuclear design codes have advanced to the stage where high confidence in the behavior and performance of a system can be achieved prior to initial testing. To help ensure reactor affordability, an optimal strategy must be devised for development and qualification. That strategy typically involves a combination of non-nuclear and nuclear testing. Non-nuclear testing is particularly useful for concepts in which nuclear operating characteristics are well understood and nuclear effects such as burnup and radiation damage are not likely to be significant. To be mass efficient, a SFPS must operate at higher coolant temperatures and use different types of power conversion than typical terrestrial reactors. The primary reason is the difficulty in rejecting excess heat to space. Although many options exist, NASA s current reference SFPS uses a fast spectrum, pumped-NaK cooled reactor coupled to a Stirling power conversion subsystem. The reference system uses technology with significant terrestrial heritage while still providing excellent performance. In addition, technologies from the SFPS system could be applicable to compact terrestrial systems. Recent non-nuclear testing at NASA s Early Flight Fission Test Facility (EFF-TF) has helped assess the viability of the reference SFPS and evaluate methods for system integration. In July, 2011 an Annular Linear Induction Pump (ALIP) provided by Idaho National Laboratory was tested at the EFF-TF to assess performance and verify suitability for use in a10 kWe technology

  1. [The Chinese nuclear test and 'atoms for peace' as a measure for preventing nuclear armament of Japan: the nuclear non-proliferation policy of the United States and the introduction of light water reactors into Japan, 1964-1968].

    Science.gov (United States)

    Yamazaki, Masakatsu

    2014-07-01

    Japan and the United States signed in 1968 a new atomic energy agreement through which US light-water nuclear reactors, including those of the Fukushima Daiichi Nuclear Power Plant of Tokyo Electric Power Company, were to be introduced into Japan. This paper studies the history of negotiations for the 1968 agreement using documents declassified in the 1990s in the US and Japan. After the success of the Chinese nuclear test in October 1964, the United States became seriously concerned about nuclear armament of other countries in Asia including Japan. Expecting that Japan would not have its own nuclear weapons, the US offered to help the country to demonstrate its superiority in some fields of science including peaceful nuclear energy to counter the psychological effect of the Chinese nuclear armament. Driven by his own political agenda, the newly appointed Prime Minister Eisaku Sato responded to the US expectation favorably. When he met in January 1965 with President Johnson, Sato made it clear that Japan would not pursue nuclear weapons. Although the US continued its support after this visit, it nevertheless gave priority to the control of nuclear technology in Japan through the bilateral peaceful nuclear agreement. This paper argues that the 1968 agreement implicitly meant a strategic measure to prevent Japan from going nuclear and also a tactic to persuade Japan to join the Nuclear Non -Proliferation Treaty.

  2. Nuclear reactor design

    CERN Document Server

    2014-01-01

    This book focuses on core design and methods for design and analysis. It is based on advances made in nuclear power utilization and computational methods over the past 40 years, covering core design of boiling water reactors and pressurized water reactors, as well as fast reactors and high-temperature gas-cooled reactors. The objectives of this book are to help graduate and advanced undergraduate students to understand core design and analysis, and to serve as a background reference for engineers actively working in light water reactors. Methodologies for core design and analysis, together with physical descriptions, are emphasized. The book also covers coupled thermal hydraulic core calculations, plant dynamics, and safety analysis, allowing readers to understand core design in relation to plant control and safety.

  3. Nuclear reactors and fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2014-07-01

    fulfill its mission that is to contribute in improving the quality of life of the Brazilian people. The nuclear fuel cycle is a series of steps involved in the production and use of fuel for nuclear reactors. The Laboratories of Chemistry and Environmental Diagnosis Center, CQMA, support the demand of Nuclear Fuel Cycle Program providing chemical characterization of uranium compounds and other related materials. In this period the Research Reactor Center (CRPq) concentrated efforts on improving equipment and systems to enable the IEA-R1 research reactor to operate at higher power, increasing the capacity of radioisotopes production, samples irradiation, tests and experiments. (author)

  4. Advanced Test Reactor National Scientific User Facility: Addressing advanced nuclear materials research

    Energy Technology Data Exchange (ETDEWEB)

    John Jackson; Todd Allen; Frances Marshall; Jim Cole

    2013-03-01

    The Advanced Test Reactor National Scientific User Facility (ATR NSUF), based at the Idaho National Laboratory in the United States, is supporting Department of Energy and industry research efforts to ensure the properties of materials in light water reactors are well understood. The ATR NSUF is providing this support through three main efforts: establishing unique infrastructure necessary to conduct research on highly radioactive materials, conducting research in conjunction with industry partners on life extension relevant topics, and providing training courses to encourage more U.S. researchers to understand and address LWR materials issues. In 2010 and 2011, several advanced instruments with capability focused on resolving nuclear material performance issues through analysis on the micro (10-6 m) to atomic (10-10 m) scales were installed primarily at the Center for Advanced Energy Studies (CAES) in Idaho Falls, Idaho. These instruments included a local electrode atom probe (LEAP), a field-emission gun scanning transmission electron microscope (FEG-STEM), a focused ion beam (FIB) system, a Raman spectrometer, and an nanoindentor/atomic force microscope. Ongoing capability enhancements intended to support industry efforts include completion of two shielded, irradiation assisted stress corrosion cracking (IASCC) test loops, the first of which will come online in early calendar year 2013, a pressurized and controlled chemistry water loop for the ATR center flux trap, and a dedicated facility intended to house post irradiation examination equipment. In addition to capability enhancements at the main site in Idaho, the ATR NSUF also welcomed two new partner facilities in 2011 and two new partner facilities in 2012; the Oak Ridge National Laboratory, High Flux Isotope Reactor (HFIR) and associated hot cells and the University California Berkeley capabilities in irradiated materials analysis were added in 2011. In 2012, Purdue University’s Interaction of Materials

  5. Nuclear Reactors and Technology

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-01-01

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

  6. Radioactive target needs for nuclear reactor physics and nuclear astrophysics

    OpenAIRE

    Jurado, B.; Barreau, G.; Bacri, C. O.

    2010-01-01

    Nuclear Instruments and Methods in Physics Research Section A - In press.; Nuclear reaction cross sections of short-lived nuclei are key inputs for new generation nuclear reactor simulations and for models describing the nucleosynthesis of elements. After discussing various topics of nuclear astrophysics and reactor physics where the demand of nuclear data on unstable nuclei is strong, we describe the general characteristics of the targets needed to measure the requested data. In some cases t...

  7. Spent nuclear fuel discharges from US reactors 1992

    Energy Technology Data Exchange (ETDEWEB)

    1994-05-05

    This report provides current statistical data on every fuel assembly irradiated in commercial nuclear reactors operating in the United States. It also provides data on the current inventories and storage capacities of those reactors to a wide audience, including Congress, Federal and State agencies, the nuclear and electric industries and the general public. It uses data from the mandatory, ``Nuclear Fuel Data`` survey, Form RW-859 for 1992 and historical data collected by the Energy Information Administration (EIA) on previous Form RW-859 surveys. The report was prepared by the EIA under a Memorandum of Understanding with the Office of Civilian Radioactive Waste Management.

  8. A cermet fuel reactor for nuclear thermal propulsion

    Science.gov (United States)

    Kruger, Gordon

    1991-01-01

    Work on the cermet fuel reactor done in the 1960's by General Electric (GE) and the Argonne National Laboratory (ANL) that had as its goal the development of systems that could be used for nuclear rocket propulsion as well as closed cycle propulsion system designs for ship propulsion, space nuclear propulsion, and other propulsion systems is reviewed. It is concluded that the work done in the 1960's has demonstrated that we can have excellent thermal and mechanical performance with cermet fuel. Thousands of hours of testing were performed on the cermet fuel at both GE and AGL, including very rapid transients and some radiation performance history. We conclude that there are no feasibility issues with cermet fuel. What is needed is reactivation of existing technology and qualification testing of a specific fuel form. We believe this can be done with a minimum development risk.

  9. Summary of experimental tests of elastomeric seismic isolation bearings for use in nuclear reactor plants

    Energy Technology Data Exchange (ETDEWEB)

    Seidensticker, R.W.; Chang, Y.W.; Kulak, R.F.

    1992-05-01

    This paper describes an experimental test program for isolator bearings which was developed to help establish the viability of using laminated elastomer bearings for base isolation of nuclear reactor plants. The goal of the test program is to determine the performance characteristics of laminated seismic isolation bearings under a wide range of loadings. Tests were performed on scale-size laminated seismic isolators both within the design shear strain range to determine the response of the bearing under expected earthquake loading conditions, and beyond the design range to determine failure modes and to establish safety margins. Three types of bearings, each produced from a different manufacturer, have been tested: (1) high shape factor-high damping-high shear modulus bearings; (2) medium shape factor-high damping-high shear modulus bearings; and (3) medium shape factor-high damping-low shear modulus bearings. All of these tests described in this report were performed at the Earthquake Engineering Research Center at the University of California, Berkeley, with technical assistance from ANL. The tests performed on the three types of bearings have confirmed the high performance characteristics of the high damping-high and low shear modulus elastomeric bearings. The bearings have shown that they are capable of having extremely large shear strains before failure occurs. The most common failure mechanism was the debonding of the top steel plate from the isolators. This failure mechanism can be virtually eliminated by improved manufacturing quality control. The most important result of the failure test of the isolators is the fact that bearings can sustain large horizontal displacement, several times larger than the design value, with failure. Their performance in moderate and strong earthquakes will be far superior to conventional structures.

  10. Summary of experimental tests of elastomeric seismic isolation bearings for use in nuclear reactor plants

    Energy Technology Data Exchange (ETDEWEB)

    Seidensticker, R.W.; Chang, Y.W.; Kulak, R.F.

    1992-01-01

    This paper describes an experimental test program for isolator bearings which was developed to help establish the viability of using laminated elastomer bearings for base isolation of nuclear reactor plants. The goal of the test program is to determine the performance characteristics of laminated seismic isolation bearings under a wide range of loadings. Tests were performed on scale-size laminated seismic isolators both within the design shear strain range to determine the response of the bearing under expected earthquake loading conditions, and beyond the design range to determine failure modes and to establish safety margins. Three types of bearings, each produced from a different manufacturer, have been tested: (1) high shape factor-high damping-high shear modulus bearings; (2) medium shape factor-high damping-high shear modulus bearings; and (3) medium shape factor-high damping-low shear modulus bearings. All of these tests described in this report were performed at the Earthquake Engineering Research Center at the University of California, Berkeley, with technical assistance from ANL. The tests performed on the three types of bearings have confirmed the high performance characteristics of the high damping-high and low shear modulus elastomeric bearings. The bearings have shown that they are capable of having extremely large shear strains before failure occurs. The most common failure mechanism was the debonding of the top steel plate from the isolators. This failure mechanism can be virtually eliminated by improved manufacturing quality control. The most important result of the failure test of the isolators is the fact that bearings can sustain large horizontal displacement, several times larger than the design value, with failure. Their performance in moderate and strong earthquakes will be far superior to conventional structures.

  11. Nuclear research reactors in Brazil

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

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

  12. Nuclear research reactors in Brazil

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

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

  13. An overview of future sustainable nuclear power reactors

    Directory of Open Access Journals (Sweden)

    Andreas Poullikkas

    2013-01-01

    Full Text Available In this paper an overview of the current and future nuclear power reactor technologies is carried out. In particular, the nuclear technology is described and the classification of the current and future nuclear reactors according to their generation is provided. The analysis has shown that generation II reactors currently in operation all around the world lack significantly in safety precautions and are prone to loss of coolant accident (LOCA. In contrast, generation III reactors, which are an evolution of generation II reactors, incorporate passive or inherent safety features that require no active controls or operational intervention to avoid accidents in the event of malfunction, and may rely on gravity, natural convection or resistance to high temperatures. Today, partly due to the high capital cost of large power reactors generating electricity and partly due to the consideration of public perception, there is a shift towards the development of smaller units. These may be built independently or as modules in a larger complex, with capacity added incrementally as required. Small reactors most importantly benefit from reduced capital costs, simpler units and the ability to produce power away from main grid systems. These factors combined with the ability of a nuclear power plant to use process heat for co-generation, make the small reactors an attractive option. Generally, modern small reactors for power generation are expected to have greater simplicity of design, economy of mass production and reduced installation costs. Many are also designed for a high level of passive or inherent safety in the event of malfunction. Generation III+ designs are generally extensions of the generation III concept, which include advanced passive safety features. These designs can maintain the safe state without the use of any active control components. Generation IV reactors, which are future designs that are currently under research and development, will

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

    CERN Document Server

    International Electrotechnical Commission. Geneva

    1998-01-01

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

  15. Design of the Next Generation Nuclear Plant Graphite Creep Experiments for Irradiation in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    S. Blaine Grover

    2009-05-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Program will be irradiating six gas reactor graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These graphite irradiations are being accomplished to support development of the next generation reactors in the United States. The graphite experiments will be irradiated over the next six to eight years to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data at different temperatures and loading conditions to support design of the Next Generation Nuclear Plant Very High Temperature Gas Reactor, as well as other future gas reactors. The experiments will each consist of a single capsule that will contain seven separate stacks of graphite specimens. Six of the specimen stacks will have half of their graphite specimens under a compressive load, while the other half of the specimens will not be subjected to a compressive load during irradiation. The six stacks will be organized into pairs with a different compressive load being applied to the top half of each pair of specimen stacks. The seventh stack will not have a compressive load on the graphite specimens during irradiation. The specimens will be irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There will also be the capability of sampling the sweep gas effluent to determine if any

  16. Advanced gas cooled nuclear reactor materials evaluation and development program. Progress report, July 1--September 30, 1978

    Energy Technology Data Exchange (ETDEWEB)

    1978-11-24

    Results of work performed from July 1, 1978 through September 30, 1978 on the Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program are presented. Candidate alloys were evaluated for Very High Temperature Reactor Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the affect of simulated reactor primary coolant (Helium containing small amounts of various other gases), the high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. The activities associated with the characterization of the materials for the screening test program are reported, i.e., test specimen preparation, information from the materials characterization tests performed by General Electric, and the status of the simulated reactor helium supply system, testing equipment, and gas chemistry analysis instrumentation and equipment. The status of the data management system is presented.

  17. Direct test of the time-independence of fundamental nuclear constants using the Oklo natural reactor

    CERN Document Server

    Shlyakhter, A I

    The positions of neutron resonances have been shown to be highly sensitive to the variation of fundamental nuclear constants. The analysis of the measured isotopic shifts in the natural fossil reactor at Oklo gives the following restrictions on the possible rates of the interaction constants variation: strong ~2x10^-19 yr^-1, electromagnetic ~5x10^-18 yr^-1, weak ~10^-12 yr^-1. These limits permit to exclude all the versions of nuclear constants contemporary variation discussed in the literature. URL: http://alexonline.info >. For more recent analyses see hep-ph/9606486, hep-ph/0205206 and astro-ph/0204069 .

  18. Radiation-induced electrical degradation experiments in the Japan materials testing reactor

    Science.gov (United States)

    Farnum, Eugene H.; Shikama, Tatsuo; Narui, Minoru; Sagawa, Tsutomu; Scarborough, Kent

    1996-02-01

    An experiment to measure radiation-induced electrical degradation (RIED) in a sapphire sample and in three MgO-insulated cables was conducted at the JMTR light water reactor. The materials were irradiated at about 260°C to a fluence of 3 × 1024 n/m 2 ( E > 1 MeV) with an applied DC electric field between 100 kV/m and 500 kV/m. Even though the results for the sapphire sample are somewhat ambiguous because of an unexplained offset current of about 0.6 μA substantial degradation was not observed in the sapphire: instead, radiation-induced conductivity (RIC) seemed to decrease slightly during the experiment. Substantial increase in leakage current, that increased with applied electric field, occurred in the MgO-insulated cables. This increased conductivity disappeared when the reactor was shut down and sample temperature returned to ambient. However, the physical degradation apparently remained in the material while the reactor was off because restarting the irradiation brought the conductivity back to its previous, degraded, reactor-on value. This effect is different from the RIED effect reported by Hodgson but is similar to previous results reported by Shikama et al. Considerable data were taken to determine the sample temperature and leakage currents during the irradiation.

  19. Turbulent jet erosion of a stably stratified gas layer in a nuclear reactor test containment

    Energy Technology Data Exchange (ETDEWEB)

    Ishay, Liel [Department of Mechanical Engineering, Ben-Gurion University of the Negev, Beer-Sheva 84105 (Israel); Bieder, Ulrich [Commissariat à l’énergie atomique et aux énergies alternatives, Centre de SACLAY DEN/SAC/DANS/DM2S/STMF/LMSF, F-91191 Gif-sur-Yvette (France); Ziskind, Gennady [Department of Mechanical Engineering, Ben-Gurion University of the Negev, Beer-Sheva 84105 (Israel); Rashkovan, Alex, E-mail: rashbgu@gmail.com [Physics Department, Nuclear Research Center Negev (NRCN), PO Box 9001, Beer-Sheva 84190 (Israel)

    2015-10-15

    Highlights: • We model stably stratified layer erosion by vertical turbulent round jet. • Separate effect studies are performed as a platform for choosing modeling approach. • A test performed in MISTRA facility, CEA, Saclay is modeled using Fluent and Trio-U codes. • The proposed modeling approach showed good agreement with the MISTRA facility LOWMA-3 test. - Abstract: A number of integral and separate effect experiments were performed in the last two decades for validation of containment computational tools. The main goal of these benchmark experiments was to assess the ability of turbulence models and computational fluid dynamics codes to predict hydrogen concentration distribution and steam condensation rate in a nuclear reactor containment in the course of severe accidents. It appears from the published literature that the predictive capability of the existing computational tools still needs to be improved. This work examines numerically the temporal evolution of helium concentration in the experiment called LOWMA-3, performed in the MISTRA facility of CEA-Saclay, France. In the experiment, helium is used to mimic hydrogen of a real-case accident. The aim of this separate effect experiment, where steam condensation was not involved, is to predict helium concentration field. The conditions of the experiment are such that both the momentum transport and molecular diffusion contributions to the mixing process are of the same order of magnitude (Fr ∼ 1). A commercial CFD code, Fluent, and a CEA in-house code, Trio-U, are used for flow and helium concentration fields temporal evolution prediction in the present study. The preliminary separate effect studies provide guidance to an optimal modeling approach for the LOWMA-3 experiment. Temporal evolution of helium concentration in the stratification layer is shown, and a comparison to the experiment is discussed. It is shown that correct modeling of the round jet flowfield is essential for a reliable

  20. General Nuclear Medicine

    Science.gov (United States)

    ... Physician Resources Professions Site Index A-Z General Nuclear Medicine Nuclear medicine imaging uses small amounts of ... limitations of General Nuclear Medicine? What is General Nuclear Medicine? Nuclear medicine is a branch of medical ...

  1. Parametric study of the energy deposition inside the calorimeter measuring the nuclear heating in Material Testing Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Amharrak, H., E-mail: hicham.amharrak@im2np.fr [Aix Marseille Université, CNRS, Université de Toulon, IM2NP UMR 7334, 13397, Marseille (France); Reynard-Carette, C. [Aix Marseille Université, CNRS, Université de Toulon, IM2NP UMR 7334, 13397, Marseille (France); Lyoussi, A. [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 Saint Paul lez Durance (France); Carette, M.; Brun, J.; De Vita, C. [Aix Marseille Université, CNRS, Université de Toulon, IM2NP UMR 7334, 13397, Marseille (France); Fourmentel, D.; Villard, J-F. [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 Saint Paul lez Durance (France)

    2015-11-01

    The nuclear heating measurements in Material Testing Reactors (MTRs) are crucial for the study of nuclear materials and fuels under irradiation. The reference measurements of this nuclear heating are especially performed by a differential calorimeter including a graphite sample material and two calorimetric cells. Then these measurements are used for other experimental conditions in order to predict the nuclear heating and thermal conditions induced in the irradiation devices. This paper will present simulations with MCNP5 Monte-Carlo transport code (using ENDF/B-VI nuclear data library) to evaluate the nuclear heating inside the calorimeter during irradiation campaigns of the CARMEN-1P mock-up inside OSIRIS reactor periphery (MTR based on Saclay, France). The whole complete geometry of the sensor has been considered. The calculation method corresponds to a calculation in two steps. Consequently, we used as an input source in the model, the neutron and photon spectra calculated in various experimental locations tested during the irradiation campaign (H9, H10, H11, D9). After a description of the differential calorimeter sensor, the MCNP5 model used for the calculations of nuclear heating inside the calorimeter elements is introduced by two quantities: KERMA and energy deposition rate per mass unit. The Charged Particle Equilibrium (CPE) inside the calorimeter elements is studied. The contribution of prompt gamma and neutron is determined. A comparison between this total nuclear heating calculation and the experimental results in a graphite sample will be made. Then parametric studies performed on the influence of the various calorimeter components on the nuclear heating are presented and discussed. The studies of the influence of the nature of materials, the sensor jacket, the source type and the comparison of the results obtained for the two calorimetric cells leads to some proposals for the sensor improvement.

  2. Research on the improvement of nuclear safety -Thermal hydraulic tests for reactor safety system-

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Moon Kee; Park, Choon Kyung; Yang, Sun Kyoo; Chun, Se Yung; Song, Chul Hwa; Jun, Hyung Kil; Jung, Heung Joon; Won, Soon Yun; Cho, Yung Roh; Min, Kyung Hoh; Jung, Jang Hwan; Jang, Suk Kyoo; Kim, Bok Deuk; Kim, Wooi Kyung; Huh, Jin; Kim, Sook Kwan; Moon, Sang Kee; Lee, Sang Il [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-06-01

    The present research aims at the development of the thermal hydraulic verification test technology for the safety system of the conventional and advanced nuclear power plant and the development of the advanced thermal hydraulic measuring techniques. In this research, test facilities simulating the primary coolant system and safety system are being constructed for the design verification tests of the existing and advanced nuclear power plant. 97 figs, 14 tabs, 65 refs. (Author).

  3. Radiation-induced electrical degradation experiments in the Japan materials testing reactor

    Energy Technology Data Exchange (ETDEWEB)

    Farnum, E.; Scharborough, K. [Los Alamos National Lab., NM (United States); Shikama, Tatsuo [and others

    1995-04-01

    The objective of this experiment is to determine the extent of degradation during neutron irradiation of electrical and optical properties of candidate dielectric materials. The goals are to identify promising dielectrics for ITER and other fusion machines for diagnostic applications and establish the basis for optimization of candidate materials. An experiment to measure radiation-induced electrical degradation (REID) in sapphire and MgO-insulated cables was conducted at the JMTR light water reactor. The materials were irradiated at about 260 {degree}C to a fluence of 3{times}10{sup 24} n/m{sup 2} (E>1 MeV) with an applied DC electric field between 100 kV/m and 500 kV/m.

  4. Nuclear reactor effluent monitoring

    Energy Technology Data Exchange (ETDEWEB)

    Minns, J.L.; Essig, T.H. [Nuclear Regulatory Commission, Washington, DC (United States)

    1993-12-31

    Radiological environmental monitoring and effluent monitoring at nuclear power plants is important both for normal operations, as well as in the event of an accident. During normal operations, environmental monitoring verifies the effectiveness of in-plant measures for controlling the release of radioactive materials in the plant. Following an accident, it would be an additional mechanism for estimating doses to members of the general public. This paper identifies the U.S. Nuclear Regulatory Commission (NRC) regulatory basis for requiring radiological environmental and effluent monitoring, licensee conditions for effluent and environmental monitoring, NRC independent oversight activities, and NRC`s program results.

  5. Nuclear data requirements for fusion reactor nucleonics

    Energy Technology Data Exchange (ETDEWEB)

    Bhat, M.R.; Abdou, M.A.

    1980-01-01

    Nuclear data requirements for fusion reactor nucleonics are reviewed and the present status of data are assessed. The discussion is divided into broad categories dealing with data for Fusion Materials Irradiation Test Facility (FMIT), D-T Fusion Reactors, Alternate Fuel Cycles and the Evaluated Data Files that are available or would be available in the near future.

  6. FY15 Status Report: CIRFT Testing of Spent Nuclear Fuel Rods from Boiler Water Reactor Limerick

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wang, Hong [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jiang, Hao [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-06-01

    The objective of this project is to perform a systematic study of used nuclear fuel (UNF, also known as spent nuclear fuel [SNF]) integrity under simulated transportation environments using the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT) hot-cell testing technology developed at Oak Ridge National Laboratory (ORNL) in August 2013. Under Nuclear Regulatory Commission (NRC) sponsorship, ORNL completed four benchmark tests, four static tests, and twelve dynamic or cycle tests on H. B. Robinson (HBR) high burn-up (HBU) fuel. The clad of the HBR fuels was made of Zircaloy-4. Testing was continued in fiscal year (FY) 2014 using Department of Energy (DOE) funds. The additional CIRFT was conducted on three HBR rods (R3, R4, and R5) in which two specimens failed and one specimen was tested to over 2.23 10⁷ cycles without failing. The data analysis on all the HBR UNF rods demonstrated that it is necessary to characterize the fatigue life of the UNF rods in terms of (1) the curvature amplitude and (2) the maximum absolute of curvature extremes. The maximum extremes are significant because they signify the maximum of tensile stress for the outer fiber of the bending rod. CIRFT testing has also addressed a large variation in hydrogen content on the HBR rods. While the load amplitude is the dominant factor that controls the fatigue life of bending rods, the hydrogen content also has an important effect on the lifetime attained at each load range tested. In FY 15, ten SNF rod segments from BWR Limerick were tested using ORNL CIRFT, with one under static and nine dynamic loading conditions. Under static unidirectional loading, a moment of 85 N·m was obtained at maximum curvature 4.0 m⁻¹. The specimen did not show any sign of failure in three repeated loading cycles to almost same maximum curvature. Ten cyclic tests were conducted with amplitude varying from 15.2 to 7.1 N·m. Failure was observed in nine of the tested rod specimens. The cycles to failure were

  7. FASTER Test Reactor Preconceptual Design Report

    Energy Technology Data Exchange (ETDEWEB)

    Grandy, C. [Argonne National Lab. (ANL), Argonne, IL (United States); Belch, H. [Argonne National Lab. (ANL), Argonne, IL (United States); Brunett, A. J. [Argonne National Lab. (ANL), Argonne, IL (United States); Heidet, F. [Argonne National Lab. (ANL), Argonne, IL (United States); Hill, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hoffman, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Jin, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Mohamed, W. [Argonne National Lab. (ANL), Argonne, IL (United States); Moisseytsev, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Passerini, S. [Argonne National Lab. (ANL), Argonne, IL (United States); Sienicki, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Sumner, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Vilim, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hayes, S. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-03-31

    The FASTER test reactor plant is a sodium-cooled fast spectrum test reactor that provides high levels of fast and thermal neutron flux for scientific research and development. The 120MWe FASTER reactor plant has a superheated steam power conversion system which provides electrical power to a local grid allowing for recovery of operating costs for the reactor plant.

  8. Severe accident testing of electrical penetration assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Clauss, D.B. (Sandia National Labs., Albuquerque, NM (USA))

    1989-11-01

    This report describes the results of tests conducted on three different designs of full-size electrical penetration assemblies (EPAs) that are used in the containment buildings of nuclear power plants. The objective of the tests was to evaluate the behavior of the EPAs under simulated severe accident conditions using steam at elevated temperature and pressure. Leakage, temperature, and cable insulation resistance were monitored throughout the tests. Nuclear-qualified EPAs were produced from D. G. O'Brien, Westinghouse, and Conax. Severe-accident-sequence analysis was used to generate the severe accident conditions (SAC) for a large dry pressurized-water reactor (PWR), a boiling-water reactor (BWR) Mark I drywell, and a BWR Mark III wetwell. Based on a survey conducted by Sandia, each EPA was matched with the severe accident conditions for a specific reactor type. This included the type of containment that a particular EPA design was used in most frequently. Thus, the D. G. O'Brien EPA was chosen for the PWR SAC test, the Westinghouse was chosen for the Mark III test, and the Conax was chosen for the Mark I test. The EPAs were radiation and thermal aged to simulate the effects of a 40-year service life and loss-of-coolant accident (LOCA) before the SAC tests were conducted. The design, test preparations, conduct of the severe accident test, experimental results, posttest observations, and conclusions about the integrity and electrical performance of each EPA tested in this program are described in this report. In general, the leak integrity of the EPAs tested in this program was not compromised by severe accident loads. However, there was significant degradation in the insulation resistance of the cables, which could affect the electrical performance of equipment and devices inside containment at some point during the progression of a severe accident. 10 refs., 165 figs., 16 tabs.

  9. Electric failure on the reactor n.3 of the nuclear power plant of Dampierre; Defaillance electrique sur le reacteur n. 3 de la centrale nucleaire de Dampierre

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2007-05-15

    This note of information resumes the progress of the electric failure on the reactor n.3 of the nuclear power plant of Dampierre, the organization during the incident, it establishes then a comparison with the incident arisen to Forsmark in 2006 and reminds that it lead in an inspection on behalf of the Asn which noticed that all the procedures had been respected by the operators and did not noticed any abnormality in the maintenance. This event was classified at the level 1 of the international nuclear event scale (INES). (N.C.)

  10. Teaching About Nature's Nuclear Reactors

    CERN Document Server

    Herndon, J M

    2005-01-01

    Naturally occurring nuclear reactors existed in uranium deposits on Earth long before Enrico Fermi built the first man-made nuclear reactor beneath Staggs Field in 1942. In the story of their discovery, there are important lessons to be learned about scientific inquiry and scientific discovery. Now, there is evidence to suggest that the Earth's magnetic field and Jupiter's atmospheric turbulence are driven by planetary-scale nuclear reactors. The subject of planetocentric nuclear fission reactors can be a jumping off point for stimulating classroom discussions about the nature and implications of planetary energy sources and about the geomagnetic field. But more importantly, the subject can help to bring into focus the importance of discussing, debating, and challenging current thinking in a variety of areas.

  11. FASTER test reactor preconceptual design report summary

    Energy Technology Data Exchange (ETDEWEB)

    Grandy, C. [Argonne National Lab. (ANL), Argonne, IL (United States); Belch, H. [Argonne National Lab. (ANL), Argonne, IL (United States); Brunett, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Heidet, F. [Argonne National Lab. (ANL), Argonne, IL (United States); Hill, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hoffman, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Jin, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Mohamed, W. [Argonne National Lab. (ANL), Argonne, IL (United States); Moisseytsev, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Passerini, S. [Argonne National Lab. (ANL), Argonne, IL (United States); Sienicki, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Sumner, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Vilim, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hayes, Steven [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-02-29

    The FASTER reactor plant is a sodium-cooled fast spectrum test reactor that provides high levels of fast and thermal neutron flux for scientific research and development. The 120MWe FASTER reactor plant has a superheated steam power conversion system which provides electrical power to a local grid allowing for recovery of operating costs for the reactor plant.

  12. Nuclear Reactor Engineering Analysis Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Carlos Chavez-Mercado; Jaime B. Morales-Sandoval; Benjamin E. Zayas-Perez

    1998-12-31

    The Nuclear Reactor Engineering Analysis Laboratory (NREAL) is a sophisticated computer system with state-of-the-art analytical tools and technology for analysis of light water reactors. Multiple application software tools can be activated to carry out different analyses and studies such as nuclear fuel reload evaluation, safety operation margin measurement, transient and severe accident analysis, nuclear reactor instability, operator training, normal and emergency procedures optimization, and human factors engineering studies. An advanced graphic interface, driven through touch-sensitive screens, provides the means to interact with specialized software and nuclear codes. The interface allows the visualization and control of all observable variables in a nuclear power plant (NPP), as well as a selected set of nonobservable or not directly controllable variables from conventional control panels.

  13. External Attachment of Titanium Sheathed Thermocouples to Zirconium Nuclear Fuel Rods For The Loss-Of-Fluid-Test (LOFT) Reactor

    Science.gov (United States)

    Welty, Richard K.

    1980-10-01

    The Exxon Nuclear Company, Inc. acting as a Subcontractor to EG&G Idaho Inc.3 Idaho National Engineering Laboratory, Idaho Falls, Idaho, has developed a welding process to attach titanium sheathed thermocouples to the outside of the zircaloy clad fuel rods. The fuel rods and thermocouples are used to test simulated loss-of-coolant-accident (LOCA) conditions in a pressurized water reactor (LOFT Reactor, Idaho National Laboratory). The design goals were to (1) reliably attach thermocouples to the zircaloy fuel rods, (2) achieve or exceed a life expectancy of 6,000 hours of reactor operation in a borated water environment of 316°C at 2260 psi, (3) provide and sustain repeatable physical and metallurgical properties in the instrumented rods subjected to transient temperatures up to 1538°C with blowdown, shock, loading, and fast quench. A laser beam was selected as the optimum welding process because of the extremely high energy input per unit volume that can be achieved allowing local fusion of a small area irrespective of the difference in material thickness to be joined. A commercial pulsed laser and energy control system was installed along with specialized welding fixtures. Laser room facility requirements and tolerances were established. Performance qualifications and detailed welding procedures were also developed. Product performance tests were conducted to assure that engineering design requirements could be met on a production basis. Irradiation tests showed no degradation of thermocouples or weld structure. Fast thermal cycle and heater rod blowdown reflood tests were made to subject the weldments to high temperatures, high pressure steam, and fast water quench cycles. From the behavior of these tests, it was concluded that the attachment welds would survive a series of reactor safety tests.

  14. The Birth of Nuclear-Generated Electricity

    Science.gov (United States)

    1999-09-01

    The Experimental Breeder Reactor-I (EBR-I), built in Idaho in 1949, generated the first usable electricity from nuclear power on December 20, 1951. More importantly, the reactor was used to prove that it was possible to create more nuclear fuel in the reactor than it consumed during operation -- fuel breeding. The EBR-I facility is now a National Historic Landmark open to the public.

  15. Fundamentals of Nuclear Reactor Physics

    CERN Document Server

    Lewis, E E

    2008-01-01

    This new streamlined text offers a one-semester treatment of the essentials of how the fission nuclear reactor works, the various approaches to the design of reactors, and their safe and efficient operation. The book includes numerous worked-out examples and end-of-chapter questions to help reinforce the knowledge presented. This textbook offers an engineering-oriented introduction to nuclear physics, with a particular focus on how those physics are put to work in the service of generating nuclear-based power, particularly the importance of neutron reactions and neutron behavior. Engin

  16. Complex risk analysis for loss of electric power in liquid metal nuclear reactor by system dynamics (SD) method

    Energy Technology Data Exchange (ETDEWEB)

    Woo, Tae Ho [Seoul National Univ. (Korea, Republic of). Dept. of Nuclear Engineering

    2012-07-15

    The power stabilization of the nuclear power plants (NPPs) is investigated in the aspect of the liquid metal coolant. The quantification of the risk analysis is performed by the system dynamics (SD) method which is processed by the feedback and accumulation complex algorithms. The Vensim software package is used for the simulations, which is supported by the Monte-Carlo method. There are 2 kinds of considerations as the economic and safety properties. The result shows the stability of the operations when the power can be decided. This shows the higher efficiency of the reactor. The failure frequency is 16/60 = 27%. In the event of Power Stabilized, the failure event is in the quite lower frequency rate. The commercial use of the reactor is important in the operations. (orig.)

  17. Advanced Carbothermal Electric Reactor Project

    Data.gov (United States)

    National Aeronautics and Space Administration — The overall objective of the Phase 1 effort was to demonstrate the technical feasibility of the Advanced Carbothermal Electric (ACE) Reactor concept. Unlike...

  18. Advanced Carbothermal Electric Reactor Project

    Data.gov (United States)

    National Aeronautics and Space Administration — ORBITEC proposes to develop the Advanced Carbothermal Electric (ACE) reactor to efficiently extract oxygen from lunar regolith. Unlike state-of-the-art carbothermal...

  19. Nuclear reactor composite fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Burgess, Donn M. (Richland, WA); Marr, Duane R. (West Richland, WA); Cappiello, Michael W. (Richland, WA); Omberg, Ronald P. (Richland, WA)

    1980-01-01

    A core and composite fuel assembly for a liquid-cooled breeder nuclear reactor including a plurality of elongated coextending driver and breeder fuel elements arranged to form a generally polygonal bundle within a thin-walled duct. The breeder elements are larger in cross section than the driver elements, and each breeder element is laterally bounded by a number of the driver elements. Each driver element further includes structure for spacing the driver elements from adjacent fuel elements and, where adjacent, the thin-walled duct. A core made up of the fuel elements can advantageously include fissile fuel of only one enrichment, while varying the effective enrichment of any given assembly or core region, merely by varying the relative number and size of the driver and breeder elements.

  20. Reactor Simulator Testing

    Science.gov (United States)

    Schoenfeld, Michael P.; Webster, Kenny L.; Pearson, Boise J.

    2013-01-01

    As part of the Nuclear Systems Office Fission Surface Power Technology Demonstration Unit (TDU) project, a reactor simulator test loop (RxSim) was design & built to perform integrated testing of the TDU components. In particular, the objectives of RxSim testing was to verify the operation of the core simulator, the instrumentation and control system, and the ground support gas and vacuum test equipment. In addition, it was decided to include a thermal test of a cold trap purification design and a pump performance test at pump voltages up to 150 V since the targeted mass flow rate of 1.75 kg/s was not obtained in the RxSim at the originally constrained voltage of 120 V. This paper summarizes RxSim testing. The gas and vacuum ground support test equipment performed effectively in NaK fill, loop pressurization, and NaK drain operations. The instrumentation and control system effectively controlled loop temperature and flow rates or pump voltage to targeted settings. The cold trap design was able to obtain the targeted cold temperature of 480 K. An outlet temperature of 636 K was obtained which was lower than the predicted 750 K but 156 K higher than the cold temperature indicating the design provided some heat regeneration. The annular linear induction pump (ALIP) tested was able to produce a maximum flow rate of 1.53 kg/s at 800 K when operated at 150 V and 53 Hz.

  1. Reactor antineutrinos and nuclear physics

    Science.gov (United States)

    Balantekin, A. B.

    2016-11-01

    Short-baseline reactor neutrino experiments successfully measured the neutrino parameters they set out to measure, but they also identified a shape distortion in the 5-7 MeV range as well as a reduction from the predicted value of the flux. Nuclear physics input into the calculations of reactor antineutrino spectra needs to be better refined if this anomaly is to be interpreted as due to sterile neutrino states.

  2. Nuclear reactor downcomer flow deflector

    Energy Technology Data Exchange (ETDEWEB)

    Gilmore, Charles B. (Greensburg, PA); Altman, David A. (Pittsburgh, PA); Singleton, Norman R. (Murrysville, PA)

    2011-02-15

    A nuclear reactor having a coolant flow deflector secured to a reactor core barrel in line with a coolant inlet nozzle. The flow deflector redirects incoming coolant down an annulus between the core barrel and the reactor vessel. The deflector has a main body with a front side facing the fluid inlet nozzle and a rear side facing the core barrel. The rear side of the main body has at least one protrusion secured to the core barrel so that a gap exists between the rear side of the main body adjacent the protrusion and the core barrel. Preferably, the protrusion is a relief that circumscribes the rear side of the main body.

  3. Fuel Fabrication and Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Karpius, Peter Joseph [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-02-02

    The uranium from the enrichment plant is still in the form of UF6. UF6 is not suitable for use in a reactor due to its highly corrosive chemistry as well as its phase diagram. UF6 is converted into UO2 fuel pellets, which are in turn placed in fuel rods and assemblies. Reactor designs are variable in moderators, coolants, fuel, performance etc.The dream of energy ‘too-cheap to meter’ is no more, and now the nuclear power industry is pushing ahead with advanced reactor designs.

  4. Results from the Operational Testing of the General Electric Smart Grid Capable Electric Vehicle Supply Equipment (EVSE)

    Energy Technology Data Exchange (ETDEWEB)

    Richard Barney Carlson; Don Scoffield; Brion Bennett

    2013-12-01

    The Idaho National Laboratory conducted testing and analysis of the General Electric (GE) smart grid capable electric vehicle supply equipment (EVSE), which was a deliverable from GE for the U.S. Department of Energy FOA-554. The Idaho National Laboratory has extensive knowledge and experience in testing advanced conductive and wireless charging systems though INL’s support of the U.S. Department of Energy’s Advanced Vehicle Testing Activity. This document details the findings from the EVSE operational testing conducted at the Idaho National Laboratory on the GE smart grid capable EVSE. The testing conducted on the EVSE included energy efficiency testing, SAE J1772 functionality testing, abnormal conditions testing, and charging of a plug-in vehicle.

  5. Reactor Simulator Testing

    Science.gov (United States)

    Schoenfeld, Michael P.; Webster, Kenny L.; Pearson, Boise Jon

    2013-01-01

    As part of the Nuclear Systems Office Fission Surface Power Technology Demonstration Unit (TDU) project, a reactor simulator test loop (RxSim) was design & built to perform integrated testing of the TDU components. In particular, the objectives of RxSim testing was to verify the operation of the core simulator, the instrumentation and control system, and the ground support gas and vacuum test equipment. In addition, it was decided to include a thermal test of a cold trap purification design and a pump performance test at pump voltages up to 150 V since the targeted mass flow rate of 1.75 kg/s was not obtained in the RxSim at the originally constrained voltage of 120 V. This paper summarizes RxSim testing. The gas and vacuum ground support test equipment performed effectively in NaK fill, loop pressurization, and NaK drain operations. The instrumentation and control system effectively controlled loop temperature and flow rates or pump voltage to targeted settings. The cold trap design was able to obtain the targeted cold temperature of 480 K. An outlet temperature of 636 K was obtained which was lower than the predicted 750 K but 156 K higher than the cold temperature indicating the design provided some heat regeneration. The annular linear induction pump (ALIP) tested was able to produce a maximum flow rate of 1.53 kg/s at 800 K when operated at 150 V and 53 Hz. Keywords: fission, space power, nuclear, liquid metal, NaK.

  6. Nozzle for electric dispersion reactor

    Science.gov (United States)

    Sisson, Warren G.; Basaran, Osman A.; Harris, Michael T.

    1995-01-01

    A nozzle for an electric dispersion reactor includes two concentric electrodes, the inner one of the two delivering disperse phase fluid into a continuous phase fluid. A potential difference generated by a voltage source creates a dispersing electric field at the end of the inner electrode.

  7. Uranium to Electricity: The Chemistry of the Nuclear Fuel Cycle

    Science.gov (United States)

    Settle, Frank A.

    2009-01-01

    The nuclear fuel cycle consists of a series of industrial processes that produce fuel for the production of electricity in nuclear reactors, use the fuel to generate electricity, and subsequently manage the spent reactor fuel. While the physics and engineering of controlled fission are central to the generation of nuclear power, chemistry…

  8. Inherently safe reactors and a second nuclear era.

    Science.gov (United States)

    Weinberg, A M; Spiewak, I

    1984-06-29

    The Swedish PIUS reactor and the German-American small modular high-temperature gas-cooled reactor are inherently safe-that is, their safety relies not upon intervention of humans or of electromechanical devices but on immutable principles of physics and chemistry. A second nuclear era may require commercialization and deployment of such inherently safe reactors, even though existing light-water reactors appear to be as safe as other well-accepted sources of central electricity, particularly hydroelectric dams.

  9. Synergistic Smart Fuel For In-pile Nuclear Reactor Measurements

    Energy Technology Data Exchange (ETDEWEB)

    James A. Smith; Dale K. Kotter; Randall A. Ali; Steven L . Garrett

    2013-10-01

    In March 2011, an earthquake of magnitude 9.0 on the Richter scale struck Japan with its epicenter on the northeast coast, near the Tohoku region. In addition to the immense physical destruction and casualties across the country, several nuclear power plants (NPP) were affected. It was the Fukushima Daiichi NPP that experienced the most severe and irreversible damage. The earthquake brought the reactors at Fukushima to an automatic shutdown and because the power transmission lines were damaged, emergency diesel generators (EDGs) were activated to ensure that there was continued cooling of the reactors and spent fuel pools. The situation was being successfully managed until the tsunami hit about forty-five minutes later with a maximum wave height of approximately 15 m. The influx of water submerged the EDGs, the electrical switchgear, and dc batteries, resulting in the total loss of power to the reactors.2 At this point, the situation became critical. There was a loss of the sensors and instrumentation within the reactor that could have provided valuable information to guide the operators to make informed decisions and avoid the unfortunate events that followed. In the light of these events, we have developed and tested a potential self-powered thermoacoustic system, which will have the ability to serve as a temperature sensor and can transmit data independently of electronic networks. Such a device is synergistic with the harsh environment of the nuclear reactor as it utilizes the heat from the nuclear fuel to provide the input power.

  10. Nuclear Data and the Oklo Natural Nuclear Reactors

    Science.gov (United States)

    Gould, C. R.; Sharapov, E. I.; Sonzogni, A. A.

    2014-04-01

    Data from the Oklo natural nuclear reactors have enabled some of the most sensitive terrestrial tests of time variation of dimensionless fundamental constants. The constraints on variation of αEM, the fine structure constant are particular good, but depend on the reliability of the nuclear data, and on the reliability of the modeling of the reactor environment. We briefly review the history of these tests and discuss our recent work in 1) attempting to better bound the temperatures at which the reactors operated, 2) investigating whether the γ-ray fluxes in the reactors could have contributed to changing lutetium isotopic abundances and 3) determining whether lanthanum isotopic data could provide an alternate estimate of the neutron fluence.

  11. Development and experimental validation of a calculation scheme for nuclear heating evaluation in the core of the OSIRIS material testing reactor

    Energy Technology Data Exchange (ETDEWEB)

    Malouch, F. [Saclay Center CEA, DEN/DANS/DM2S/SERMA, F-91191 Gif-sur-Yvette Cedex (France)

    2011-07-01

    The control of the temperature in material samples irradiated in a material testing reactor requires the knowledge of the nuclear heating caused by the energy deposition by neutrons and photons interacting in the irradiation device structures. Thus, a neutron-photonic three-dimensional calculation scheme has been developed to evaluate the nuclear heating in experimental devices irradiated in the core of the OSIRIS MTR reactor (CEA/Saclay Center). The aim is to obtain a predictive tool for the nuclear heating estimation in irradiation devices. This calculation scheme is mainly based on the TRIPOLI-4 three-dimensional continuous-energy Monte Carlo transport code, developed by CEA (Saclay Center). An experimental validation has been carried out on the basis of nuclear heating measurements performed in the OSIRIS core. After an overview of the experimental devices irradiated in the OSIRIS reactor, we present the calculation scheme and the first results of the experimental validation. (authors)

  12. Advanced gas cooled nuclear reactor materials evaluation and development program. Progress report, April 1--June 30, 1978

    Energy Technology Data Exchange (ETDEWEB)

    1978-08-31

    The objectives of the program are to evaluate candidate alloys for Very High Temperature Reactor Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the affect of simulated reactor primary coolant (Helium containing small amounts of various other gases), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in the report includes the activities associated with the procurement of the materials for the screening test program, information from vendor certification for the materials receiver, and preliminary information from the materials characterization tests performed by General Electric. The construction status of the simulated reactor helium supply system, testing equipment, and gas chemistry analysis instrumentation and equipment are discussed. The status of the data management system is also reviewed.

  13. Flow Components in a NaK Test Loop Designed to Simulate Conditions in a Nuclear Surface Power Reactor

    Science.gov (United States)

    Polzin, Kurt A.; Godfroy, Thomas J.

    2008-01-01

    A test loop using NaK as the working fluid is presently in use to study material compatibility effects on various components that comprise a possible nuclear reactor design for use on the lunar surface. A DC electromagnetic (EM) pump has been designed and implemented as a means of actively controlling the NaK flow rate through the system and an EM flow sensor is employed to monitor the developed flow rate. These components allow for the matching of the flow rate conditions in test loops with those that would be found in a full-scale surface-power reactor. The design and operating characteristics of the EM pump and flow sensor are presented. In the EM pump, current is applied to a set of electrodes to produce a Lorentz body force in the fluid. A measurement of the induced voltage (back-EMF) in the flow sensor provides the means of monitoring flow rate. Both components are compact, employing high magnetic field strength neodymium magnets thermally coupled to a water-cooled housing. A vacuum gap limits the heat transferred from the high temperature NaK tube to the magnets and a magnetically-permeable material completes the magnetic circuit. The pump is designed to produce a pressure rise of 5 psi, and the flow sensor's predicted output is roughly 20 mV at the loop's nominal flow rate of 0.5 GPM.

  14. Gaseous fuel nuclear reactor research

    Science.gov (United States)

    Schwenk, F. C.; Thom, K.

    1975-01-01

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

  15. Nuclear Reactors and Technology; (USA)

    Energy Technology Data Exchange (ETDEWEB)

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

    1991-01-01

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

  16. Nuclear stress test

    Science.gov (United States)

    ... Persantine stress test; Thallium stress test; Stress test - nuclear; Adenosine stress test; Regadenoson stress test; CAD - nuclear stress; Coronary artery disease - nuclear stress; Angina - nuclear ...

  17. Proliferation Resistant Nuclear Reactor Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Gray, L W; Moody, K J; Bradley, K S; Lorenzana, H E

    2011-02-18

    Global appetite for fission power is projected to grow dramatically this century, and for good reason. Despite considerable research to identify new sources of energy, fission remains the most plentiful and practical alternative to fossil fuels. The environmental challenges of fossil fuel have made the fission power option increasingly attractive, particularly as we are forced to rely on reserves in ecologically fragile or politically unstable corners of the globe. Caught between a globally eroding fossil fuel reserve as well as the uncertainty and considerable costs in the development of fusion power, most of the world will most likely come to rely on fission power for at least the remainder of the 21st century. Despite inevitable growth, fission power faces enduring challenges in sustainability and security. One of fission power's greatest hurdles to universal acceptance is the risk of potential misuse for nefarious purposes of fissionable byproducts in spent fuel, such as plutonium. With this issue in mind, we have discussed intrinsic concepts in this report that are motivated by the premise that the utility, desirability, and applicability of nuclear materials can be reduced. In a general sense, the intrinsic solutions aim to reduce or eliminate the quantity of existing weapons usable material; avoid production of new weapons-usable material through enrichment, breeding, extraction; or employ engineering solutions to make the fuel cycle less useful or more difficult for producing weapons-usable material. By their nature, these schemes require modifications to existing fuel cycles. As such, the concomitants of these modifications require engagement from the nuclear reactor and fuel-design community to fully assess their effects. Unfortunately, active pursuit of any scheme that could further complicate the spread of domestic nuclear power will probably be understandably unpopular. Nevertheless, the nonproliferation and counterterrorism issues are paramount

  18. Cold nuclear fusion reactor and nuclear fusion rocket

    Directory of Open Access Journals (Sweden)

    Huang Zhenqiang

    2013-10-01

    Full Text Available "Nuclear restraint inertial guidance directly hit the cold nuclear fusion reactor and ion speed dc transformer" [1], referred to as "cold fusion reactor" invention patents, Chinese Patent Application No. CN: 200910129632.7 [2]. The invention is characterized in that: at room temperature under vacuum conditions, specific combinations of the installation space of the electromagnetic field, based on light nuclei intrinsic magnetic moment and the electric field, the first two strings of the nuclei to be bound fusion on the same line (track of. Re-use nuclear spin angular momentum vector inherent nearly the speed of light to form a super strong spin rotation gyro inertial guidance features, to overcome the Coulomb repulsion strong bias barrier to achieve fusion directly hit. Similar constraints apply nuclear inertial guidance mode for different speeds and energy ion beam mixing speed, the design of ion speed dc transformer is cold fusion reactors, nuclear fusion engines and such nuclear power plants and power delivery systems start important supporting equipment, so apply for a patent merger

  19. Conceptual Design of a Nuclear Reactor Dedicated for Desalination

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Yong Hun; Moon, Jang Sik; Jeong, Yong Hoon [Korea Adavanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2014-05-15

    The many advantages of nuclear desalination, the nuclear safety issues still remain a perennial problem today. To respond to such needs, the development of a desalination-dedicated nuclear reactor with maximized safety features was proposed. From the feasibility study, the desalination-dedicated reactor was found to be a good solution for meeting future water demand during the winter season in some countries like UAE by decoupling water and electricity supply. The economic analysis results indicated that under certain conditions, the desalination-dedicated reactor can produce freshwater at lower cost than the target nuclear cogeneration reactor using steam extraction technologies. A conceptual design of the desalination-dedicated nuclear reactor is in progress. The design features of the desalination-dedicated nuclear reactor could significantly enhance safety, reliability, and simplicity, and facilitate the extensive use of innovative passive safety systems. These maximized safety features of desalination-dedicated reactor could provide advanced capabilities for passive reactor shutdown and residual heat removal, and eventually prevent radioactivity release into the environment. The conceptual design achieved will provide a foothold for the future commercialization of the desalination-dedicated nuclear reactor and eventually help to address both a serious water crisis and nuclear safety issues.

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

    Energy Technology Data Exchange (ETDEWEB)

    Konzek, G.J.

    1983-07-01

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

  1. Development of Electrical Capacitance Sensors for Accident Tolerant Fuel (ATF) Testing at the Transient Reactor Test (TREAT) Facility

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Maolong; Ryals, Matthew; Ali, Amir; Blandford, Edward; Jensen, Colby; Condie, Keith; Svoboda, John; O' Brien, Robert

    2016-08-01

    A variety of instruments are being developed and qualified to support the Accident Tolerant Fuels (ATF) program and future transient irradiations at the Transient Reactor Test (TREAT) facility at Idaho National Laboratory (INL). The University of New Mexico (UNM) is working with INL to develop capacitance-based void sensors for determining the timing of critical boiling phenomena in static capsule fuel testing and the volume-averaged void fraction in flow-boiling in-pile water loop fuel testing. The static capsule sensor developed at INL is a plate-type configuration, while UNM is utilizing a ring-type capacitance sensor. Each sensor design has been theoretically and experimentally investigated at INL and UNM. Experiments are being performed at INL in an autoclave to investigate the performance of these sensors under representative Pressurized Water Reactor (PWR) conditions in a static capsule. Experiments have been performed at UNM using air-water two-phase flow to determine the sensitivity and time response of the capacitance sensor under a flow boiling configuration. Initial measurements from the capacitance sensor have demonstrated the validity of the concept to enable real-time measurement of void fraction. The next steps include designing the cabling interface with the flow loop at UNM for Reactivity Initiated Accident (RIA) ATF testing at TREAT and further characterization of the measurement response for each sensor under varying conditions by experiments and modeling.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  3. Small nuclear power reactor emergency electric power supply system reliability comparative analysis; Analise da confiabilidade do sistema de suprimento de energia eletrica de emergencia de um reator nuclear de pequeno porte

    Energy Technology Data Exchange (ETDEWEB)

    Bonfietti, Gerson

    2003-07-01

    This work presents an analysis of the reliability of the emergency power supply system, of a small size nuclear power reactor. Three different configurations are investigated and their reliability analyzed. The fault tree method is used as the main tool of analysis. The work includes a bibliographic review of emergency diesel generator reliability and a discussion of the design requirements applicable to emergency electrical systems. The influence of common cause failure influences is considered using the beta factor model. The operator action is considered using human failure probabilities. A parametric analysis shows the strong dependence between the reactor safety and the loss of offsite electric power supply. It is also shown that common cause failures can be a major contributor to the system reliability. (author)

  4. Foundational development of an advanced nuclear reactor integrated safety code.

    Energy Technology Data Exchange (ETDEWEB)

    Clarno, Kevin (Oak Ridge National Laboratory, Oak Ridge, TN); Lorber, Alfred Abraham; Pryor, Richard J.; Spotz, William F.; Schmidt, Rodney Cannon; Belcourt, Kenneth (Ktech Corporation, Albuquerque, NM); Hooper, Russell Warren; Humphries, Larry LaRon

    2010-02-01

    This report describes the activities and results of a Sandia LDRD project whose objective was to develop and demonstrate foundational aspects of a next-generation nuclear reactor safety code that leverages advanced computational technology. The project scope was directed towards the systems-level modeling and simulation of an advanced, sodium cooled fast reactor, but the approach developed has a more general applicability. The major accomplishments of the LDRD are centered around the following two activities. (1) The development and testing of LIME, a Lightweight Integrating Multi-physics Environment for coupling codes that is designed to enable both 'legacy' and 'new' physics codes to be combined and strongly coupled using advanced nonlinear solution methods. (2) The development and initial demonstration of BRISC, a prototype next-generation nuclear reactor integrated safety code. BRISC leverages LIME to tightly couple the physics models in several different codes (written in a variety of languages) into one integrated package for simulating accident scenarios in a liquid sodium cooled 'burner' nuclear reactor. Other activities and accomplishments of the LDRD include (a) further development, application and demonstration of the 'non-linear elimination' strategy to enable physics codes that do not provide residuals to be incorporated into LIME, (b) significant extensions of the RIO CFD code capabilities, (c) complex 3D solid modeling and meshing of major fast reactor components and regions, and (d) an approach for multi-physics coupling across non-conformal mesh interfaces.

  5. Long term out-of-pile thermocouple tests in conditions representative for nuclear gas-cooled high temperature reactors

    OpenAIRE

    LAURIE Mathias; FOURREZ Stéphane; FUETTERER Michael; LAPETITE Jean-Marc; SADLI M.; MORICE Ronan; FAILLEAU G

    2013-01-01

    During irradiation tests at high temperature failure of commercial Inconel 600 sheathed thermocouples is commonly encountered. As instrumentation, in particular thermocouples are considered safety-relevant both for irradiation tests and for commercial reactors, JRC and THERMOCOAX joined forces to solve this issue by performing out-of-pile tests with thermocouples mimicking the environment encountered by high temperature reactor (HTR) in-core instrumentation. The objective was to screen innova...

  6. Fabrication and testing of a 4-node micro-pocket fission detector array for the Kansas State University TRIGA Mk. II research nuclear reactor

    Science.gov (United States)

    Reichenberger, Michael A.; Nichols, Daniel M.; Stevenson, Sarah R.; Swope, Tanner M.; Hilger, Caden W.; Unruh, Troy C.; McGregor, Douglas S.; Roberts, Jeremy A.

    2017-08-01

    Advancements in nuclear reactor core modeling and computational capability have encouraged further development of in-core neutron sensors. Micro-Pocket Fission Detectors (MPFDs) have been fabricated and tested previously, but successful testing of these prior detectors was limited to single-node operation with specialized designs. Described in this work is a modular, four-node MPFD array fabricated and tested at Kansas State University (KSU). The four sensor nodes were equally spaced to span the length of the fuel-region of the KSU TRIGA Mk. II research nuclear reactor core. The encapsulated array was filled with argon gas, serving as an ionization medium in the small cavities of the MPFDs. The unified design improved device ruggedness and simplified construction over previous designs. A 0.315-in. (8-mm) penetration in the upper grid plate of the KSU TRIGA Mk. II research nuclear reactor was used to deploy the array between fuel elements in the core. The MPFD array was coupled to an electronic support system which has been developed to support pulse-mode operation. Neutron-induced pulses were observed on all four sensor channels. Stable device operation was confirmed by testing under steady-state reactor conditions. Each of the four sensors in the array responded to changes in reactor power between 10 kWth and full power (750 kWth). Reactor power transients were observed in real-time including positive transients with periods of 5, 15, and 30 s. Finally, manual reactor power oscillations were observed in real-time.

  7. Methodology for the integral comparison of nuclear reactors: selecting a reactor for Mexico; Metodologia para la comparacion integral de reactores nucleares: seleccion de un reactor para Mexico

    Energy Technology Data Exchange (ETDEWEB)

    Reyes R, R.; Martin del Campo M, C. [UNAM, Facultad de Ingenieria, Laboratorio de Analisis de Ingenieria de Reactores Nucleares, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico)]. e-mail: ricarera@yahoo.com.mx

    2006-07-01

    In this work it was built a methodology to compare nuclear reactors of third generation that can be contemplated for future electric planning in Mexico. This methodology understands the selection of the reactors to evaluate eliminating the reactors that still are not sufficiently mature at this time of the study. A general description of each reactor together with their main ones characteristic is made. It was carried out a study for to select the group of parameters that can serve as evaluation indicators, which are the characteristics of the reactors with specific values for each considered technology, and it was elaborated an evaluation matrix indicators including the reactors in the columns and those indicators in the lines. Since that none reactor is the best in all the indicators were necessary to use a methodology for multi criteria taking decisions that we are presented. It was used the 'Fuzzy Logic' technique, the which is based in those denominated diffuse groups and in a system of diffuse inference based on heuristic rules in the way 'If Then consequence> ', where the linguistic values of the condition and of the consequence is defined by diffuse groups, it is as well as the rules always they transform a diffuse group into another. Later on they combine all the diffuse outputs to create a single output and an inverse transformation is made that it transfers the output from the diffuse domain to the real one. They were carried out two studies one with the entirety of the indicators and another in which the indicators were classified in three approaches: the first one refers to indicators related with the planning of the plants inside the context of the general electric sector, the second approach includes indicators related with the characteristics of the fuel and the third it considers indicators related with the acting of the plant in safety and environmental impact. This second study allowed us to know the qualities of

  8. Multivariable Feedback Control of Nuclear Reactors

    Directory of Open Access Journals (Sweden)

    Rune Moen

    1982-07-01

    Full Text Available Multivariable feedback control has been adapted for optimal control of the spatial power distribution in nuclear reactor cores. Two design techniques, based on the theory of automatic control, were developed: the State Variable Feedback (SVF is an application of the linear optimal control theory, and the Multivariable Frequency Response (MFR is based on a generalization of the traditional frequency response approach to control system design.

  9. Advanced Nuclear Reactor Concepts for China

    Energy Technology Data Exchange (ETDEWEB)

    Knoche, D.; Sassen, F.; Tietsch, W. [Westinghouse Electric Germany, Postfach 10 05 63, 68140 Mannheim (Germany); Yujie, Dong; Li, Cao [INET, Tsinghua University, 100084 Beijing (China)

    2008-07-01

    China is one of the fastest growing economies in the world. With 1.3 billion people China also has the largest population worldwide. The growing economy, the migration of people from rural areas to cities and the augmentation in living standard will drive the energy demand of China in the coming decades. At present the installed electrical power is about 500 GW. In the years 2004 and 2005 the added electrical capacity was around 60 GW per year. Chinas primary energy demand is covered mainly by the use of coal. Coal also will remain the main energy source in the coming decades in China. Nevertheless taking into account more and more environmental aspects and the goal to reduce dependencies on energy imports a better energy mix strategy is planed to change including at an increasing level the renewable and nuclear option. Present the nuclear park is characterised by a large variety of different types of reactors. With the AP-1000, EPR and the gas-cooled High Temperature Reactor (HTR) the spectrum of different reactor types will be further enlarged. (authors)

  10. Osiris, an irradiation reactor for material and nuclear fuel testing; Osiris, reacteur d'irradiation pour materiaux et combustibles

    Energy Technology Data Exchange (ETDEWEB)

    Loubiere, S.; Durande-Ayme, P. [CEA Saclay, Div. Nucleaire Energie, Dept. Reacteurs et Nucleaire Service, 91 - Gif-sur-Yvette (France)

    2005-07-01

    Since 1966 the Osiris reactor located at Saclay has been participating in French and international irradiation programs for research and development in the field of nuclear fuel and materials. Today the French atomic commission (Cea) pursues irradiation programs in support of existing reactors, mainly PWR, strengthening its own knowledge and the one of its clients on fuel and material behaviour under irradiation, pertaining to plant life-time issues and high burn-up. For instance important programs have been performed on pressure vessel steel aging, pellet-clad interaction, internal component aging and mox fuel qualification. With the arising of the Generation 4 research and development programs, the Osiris reactor has developed capacities to undertake material and fuel irradiation under high temperature conditions. Routine irradiations such as the doping of silicon or the production of radio-nuclides for medical or imaging purposes are made on a daily basis. The specificities of the Osiris reactor are presented in the first part of this paper while the second part focuses on the experimental devices available in Osiris to perform irradiation in light water reactor conditions and in high temperature reactor conditions and on their associated programs.

  11. Fast Reactor Subassembly Design Modifications for Increasing Electricity Generation Efficiency

    Energy Technology Data Exchange (ETDEWEB)

    R. Wigeland; K. Hamman

    2009-09-01

    Suggested for Track 7: Advances in Reactor Core Design and In-Core Management _____________________________________________________________________________________ Fast Reactor Subassembly Design Modifications for Increasing Electricity Generation Efficiency R. Wigeland and K. Hamman Idaho National Laboratory Given the ability of fast reactors to effectively transmute the transuranic elements as are present in spent nuclear fuel, fast reactors are being considered as one element of future nuclear power systems to enable continued use and growth of nuclear power by limiting high-level waste generation. However, a key issue for fast reactors is higher electricity cost relative to other forms of nuclear energy generation. The economics of the fast reactor are affected by the amount of electric power that can be produced from a reactor, i.e., the thermal efficiency for electricity generation. The present study is examining the potential for fast reactor subassembly design changes to improve the thermal efficiency by increasing the average coolant outlet temperature without increasing peak temperatures within the subassembly, i.e., to make better use of current technology. Sodium-cooled fast reactors operate at temperatures far below the coolant boiling point, so that the maximum coolant outlet temperature is limited by the acceptable peak temperatures for the reactor fuel and cladding. Fast reactor fuel subassemblies have historically been constructed using a large number of small diameter fuel pins contained within a tube of hexagonal cross-section, or hexcan. Due to this design, there is a larger coolant flow area next to the hexcan wall as compared to flow area in the interior of the subassembly. This results in a higher flow rate near the hexcan wall, overcooling the fuel pins next to the wall, and a non-uniform coolant temperature distribution. It has been recognized for many years that this difference in sodium coolant temperature was detrimental to achieving

  12. Heat for industry from nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kikoin, I.K.; Novikov, V.M.

    Two factors which incline nations toward the use of heat from nuclear reactors for industrial use are: 1) exhaustion of cheap fossil fuel resources, and 2) ecological problems associated both with extraction of fossil fuel from the earth and with its combustion. In addition to the usual problems that beset nuclear reactors, special problems associated with using heat from nuclear reactors in various industries are explored.

  13. REACTOR FUEL ELEMENTS TESTING CONTAINER

    Science.gov (United States)

    Whitham, G.K.; Smith, R.R.

    1963-01-15

    This patent shows a method for detecting leaks in jacketed fuel elements. The element is placed in a sealed tank within a nuclear reactor, and, while the reactor operates, the element is sparged with gas. The gas is then led outside the reactor and monitored for radioactive Xe or Kr. (AEC)

  14. Autonomous Control of Space Nuclear Reactors Project

    Data.gov (United States)

    National Aeronautics and Space Administration — Nuclear reactors to support future robotic and manned missions impose new and innovative technological requirements for their control and protection instrumentation....

  15. Generalized Nuclear Data

    Energy Technology Data Exchange (ETDEWEB)

    2017-03-15

    This software is code related to reading/writing/manipulating nuclear data in the Generalized Nuclear Data (GND) format, a new format for sharing nuclear data among institutions. In addition to the software and its documentation, notes and documentation from the WPEC Subgroup 43 will be included. WPEC Subgroup 43 is an international committee charged with creating the API for the GND format.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-12-31

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

  17. The current status of nuclear research reactor in Thailand

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-10-01

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

  18. Licensed reactor nuclear safety criteria applicable to DOE reactors

    Energy Technology Data Exchange (ETDEWEB)

    1991-04-01

    The Department of Energy (DOE) Order DOE 5480.6, Safety of Department of Energy-Owned Nuclear Reactors, establishes reactor safety requirements to assure that reactors are sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that adequately protects health and safety and is in accordance with uniform standards, guides, and codes which are consistent with those applied to comparable licensed reactors. This document identifies nuclear safety criteria applied to NRC (Nuclear Regulatory Commission) licensed reactors. The titles of the chapters and sections of USNRC Regulatory Guide 1.70, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, Rev. 3, are used as the format for compiling the NRC criteria applied to the various areas of nuclear safety addressed in a safety analysis report for a nuclear reactor. In each section the criteria are compiled in four groups: (1) Code of Federal Regulations, (2) US NRC Regulatory Guides, SRP Branch Technical Positions and Appendices, (3) Codes and Standards, and (4) Supplemental Information. The degree of application of these criteria to a DOE-owned reactor, consistent with their application to comparable licensed reactors, must be determined by the DOE and DOE contractor.

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

    Science.gov (United States)

    Corradini, M L

    2007-11-01

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

  20. Final Report for the 1st Surveillance Test of the Reactor Pressure Vessel Material (CAPSULE 2) of Ulchin Nuclear Power Plant Unit 3

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai (and others)

    2006-12-15

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 1st surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejon after the capsule was transported from Ulchin site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Ulchin unit 3 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsule 2 is 4.674E 18n/cm{sup 2}. The bias factor, the ratio of calculation/measurement, was 0.920 for the 1st testing and the calculational uncertainty,7.0% satisfied the requirement of USNRC Reg.Guide 1.190, 20%. The best estimated neutron fluence for reactor vessel inside surface was 3.913E 18n/cm{sup 2} based on the end of 6th fuel cycle and it was predicted that the fluences of vessel inside surface at 16 and 32EFPY would reach 9.249E 18 and 1.834E 19n/cm{sup 2} based on the current calculation. The result through this analysis for Ulchin unit 3 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life.

  1. Final Report for the 1st Surveillance Test of the Reactor Pressure Vessel Material (Capsule 2) of Ulchin Nuclear Power Plant Unit 4

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai (and others)

    2007-04-15

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 1st surveillance testing was performed completely by Korea Atomic Energy Research Institute at Daejon after the capsule was transported from Ulchin site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Ulchin Unit 4 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsule 2 is 4.306E+18n/cm{sup 2}. The bias factor, the ratio of calculation/measurement, was 0.918 for the 1st testing and the calculational uncertainty,7.0% satisfied the requirement of USNRC Reg.Guide 1.190, 20%. The best estimated neutron fluence for reactor vessel inside surface was 3.615E+18n/cm{sup 2} based on the end of 6th fuel cycle and it was predicted that the fluences of vessel inside surface at 16 and 32EFPY would reach 8.478E+18 and 1.673E+19n/cm{sup 2} based on the current calculation. The result through this analysis for Ulchin Unit 4 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life.

  2. Plutonium Discharge Rates and Spent Nuclear Fuel Inventory Estimates for Nuclear Reactors Worldwide

    Energy Technology Data Exchange (ETDEWEB)

    Brian K. Castle; Shauna A. Hoiland; Richard A. Rankin; James W. Sterbentz

    2012-09-01

    This report presents a preliminary survey and analysis of the five primary types of commercial nuclear power reactors currently in use around the world. Plutonium mass discharge rates from the reactors’ spent fuel at reload are estimated based on a simple methodology that is able to use limited reactor burnup and operational characteristics collected from a variety of public domain sources. Selected commercial reactor operating and nuclear core characteristics are also given for each reactor type. In addition to the worldwide commercial reactors survey, a materials test reactor survey was conducted to identify reactors of this type with a significant core power rating. Over 100 material or research reactors with a core power rating >1 MW fall into this category. Fuel characteristics and spent fuel inventories for these material test reactors are also provided herein.

  3. Decommissioning of the Nuclear Reactors R2 and R2-0 at Studsvik, Sweden. General Data as called for under Article 37 of the Euratom Treaty

    Energy Technology Data Exchange (ETDEWEB)

    2009-01-15

    This document describes the plans for decommissioning of the nuclear research and material test reactors R2 and R2-0, situated at the Studsvik site close to the city of Nykoeping, Sweden. The purpose of the document is to serve as information for the European Commission, and to fulfil the requirements of Article 37 of the Euratom Treaty. Studsvik is situated on the Baltic coast, about 20 km east of Nykoeping and 80 km southwest of Stockholm. The site comprises the reactors R2 and R2-0 and several facilities for material investigation and radioactive waste treatment and storage. The reactors were used for a number of different purposes from 1960 until June 2005, when they were shut down following a decision by the operator. Decommissioning of the reactor facility is planned to be completed in 2016 after dismantling and conditioning of radioactive parts and demolition of the facility. Solid and liquid radioactive wastes from the dismantling activities will be treated and stored on-site awaiting final disposal. The waste treatment facilities, which are situated in other buildings at the Studsvik site, are planned to continue operation during and after the decommissioning of the reactor facility. All nuclear fuel has been transferred to a separate storage facility and is being shipped to the US according to existing agreements. The objective of the planned dismantling activities is to achieve clearance of the facility to make it possible to either demolish the buildings or use them for other purposes. The operator has divided the planning for dismantling and demolition of the facility into three phases [1]: Dismantling 1, including primary system decontamination, dismantling of the reactors with systems in the reactor pool, draining, cleaning and temporary covering of the reactor pool. This phase has begun and is due to last till approximately December 2009. Dismantling 2, including dismantling of systems in the reactor facility, removal of equipment, radiological

  4. Experimental determination of nuclear parameters for RP-0 reactor core; Determinacion experimental de los parametros nucleares para el nucleo tipo MTR del reactor nuclear RP-0

    Energy Technology Data Exchange (ETDEWEB)

    Cajacuri, Rafael A. [Sao Paulo Univ., SP (Brazil). Inst. de Fisica

    2000-07-01

    In the nuclear reactor for investigations RP-0 which is in Lima, Peru, that is a open pool class reactor with 1 to 10 watts of power and as a nuclear fuel uranium 238 enriched to 20% constituted by elements of Material Testing Reactor fuel class. This has reflectors of graphite and moderator of water demineralized. In 1996/1997 was measured in this reactor the following parameters: position of the control bar that make critic the reactor, critic height of moderator, excess of reactivity of the nucleus, parameter of reactivity for vacuum, parameter of reactivity for temperature, reactivity of its control bar, levels of doses in the reactor. (author)

  5. Closed Brayton cycle power conversion systems for nuclear reactors :

    Energy Technology Data Exchange (ETDEWEB)

    Wright, Steven A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Lipinski, Ronald J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Vernon, Milton E. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sanchez, Travis [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2006-04-01

    This report describes the results of a Sandia National Laboratories internally funded research program to study the coupling of nuclear reactors to gas dynamic Brayton power conversion systems. The research focused on developing integrated dynamic system models, fabricating a 10-30 kWe closed loop Brayton cycle, and validating these models by operating the Brayton test-loop. The work tasks were performed in three major areas. First, the system equations and dynamic models for reactors and Closed Brayton Cycle (CBC) systems were developed and implemented in SIMULINKTM. Within this effort, both steady state and dynamic system models for all the components (turbines, compressors, reactors, ducting, alternators, heat exchangers, and space based radiators) were developed and assembled into complete systems for gas cooled reactors, liquid metal reactors, and electrically heated simulators. Various control modules that use proportional-integral-differential (PID) feedback loops for the reactor and the power-conversion shaft speed were also developed and implemented. The simulation code is called RPCSIM (Reactor Power and Control Simulator). In the second task an open cycle commercially available Capstone C30 micro-turbine power generator was modified to provide a small inexpensive closed Brayton cycle test loop called the Sandia Brayton test-Loop (SBL-30). The Capstone gas-turbine unit housing was modified to permit the attachment of an electrical heater and a water cooled chiller to form a closed loop. The Capstone turbine, compressor, and alternator were used without modification. The Capstone systems nominal operating point is 1150 K turbine inlet temperature at 96,000 rpm. The annular recuperator and portions of the Capstone control system (inverter) and starter system also were reused. The rotational speed of the turbo-machinery is controlled by adjusting the alternator load by using the electrical grid as the load bank. The SBL-30 test loop was operated at

  6. Long Term Out-of-pile Thermocouple Tests in Conditions Representative for Nuclear Gas-cooled High Temperature Reactors

    OpenAIRE

    LAURIE Mathias; FOURREZ Stephane; FUETTERER Michael; LAPETITE Jean-Marc

    2011-01-01

    During irradiation tests at high temperature, failure of commercial Inconel 600 sheathed thermocouples is commonly encountered. To understand and remediate this problem, out-ofpile tests were performed with thermocouples in carburizing atmospheres which can be assumed to be at least locally representative for High Temperature Reactors. The objective was to screen those thermocouples which would consecutively be used under irradiation. Two such screening tests have been performed with a set of...

  7. Licensed reactor nuclear safety criteria applicable to DOE reactors

    Energy Technology Data Exchange (ETDEWEB)

    1993-11-01

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

  8. Nuclear fission reactors from thousand of million years; Reactores de fision nuclear de hace miles de millones de anos

    Energy Technology Data Exchange (ETDEWEB)

    Bulbulian G, S.; Ordonez R, E.; Fernandez V, S.M. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico)

    2005-07-01

    This book is about nuclear reactors, not only of the industrial ones that work to provide electric power, neither of those experimental ones as the first one that worked in Chicago in the first half of the XX Century but, mainly, of those that worked in the Earth thousands of millions of years ago. The book examines what happened in last geologic times, when the natural uranium had a different constitution to the current one. We will give you information on the nuclear fission reactors that worked in Gabon, Africa. A discussion of the radioactive isotopes formed during the operation of those reactors and its behavior until our days is presented. (Author)

  9. The Design of a Nuclear Reactor

    Indian Academy of Sciences (India)

    2016-09-01

    The aim of this largely pedagogical article is toemploy pre-college physics to arrive at an understanding of a system as complex as a nuclear reactor. We focus on three key issues: the fuelpin, the moderator, and lastly the dimensions ofthe nuclear reactor.

  10. Actinide transmutation in nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bultman, J.H.

    1995-01-17

    An optimization method is developed to maximize the burning capability of the ALMR while complying with all constraints imposed on the design for reliability and safety. This method leads to a maximal transuranics enrichment, which is being limited by constraints on reactivity. The enrichment can be raised by using the neutrons less efficiently by increasing leakage from the fuel. With the developed optimization method, a metallic and an oxide fueled ALMR were optimized. Both reactors perform equally well considering the burning of transuranics. However, metallic fuel has a much higher heat conductivity coefficient, which in general leads to better safety characteristics. In search of a more effective waste transmuter, a modified Molten Salt Reactor was designed. A MSR operates on a liquid fuel salt which makes continuous refueling possible, eliminating the issue of the burnup reactivity loss. Also, a prompt negative reactivity feedback is possible for an overmoderated reactor design, even when the Doppler coefficient is positive, due to the fuel expansion with fuel temperature increase. Furthermore, the molten salt fuel can be reprocessed based on a reduction process which is not sensitive to the short-lived spontaneously fissioning actinides. (orig./HP).

  11. Design and axial optimization of nuclear fuel for BWR reactors; Diseno y optimizacion axial de combustible nuclear para reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Garcia V, M.A

    2006-07-01

    In the present thesis, the modifications made to the axial optimization system based on Tabu Search (BT) for the axial design of BWR fuel type are presented, developed previously in the Nuclear Engineering Group of the UNAM Engineering Faculty. With the modifications what is mainly looked is to consider the particular characteristics of the mechanical design of the GE12 fuel type, used at the moment in the Laguna Verde Nucleo electric Central (CNLV) and that it considers the fuel bars of partial longitude. The information obtained in this thesis will allow to plan nuclear fuel reloads with the best conditions to operate in a certain cycle guaranteeing a better yield and use in the fuel burnt, additionally people in charge in the reload planning will be favored with the changes carried out to the system for the design and axial optimization of nuclear fuel, which facilitate their handling and it reduces their execution time. This thesis this developed in five chapters that are understood in the following way in general: Chapter 1: It approaches the basic concepts of the nuclear energy, it describes the physical and chemical composition of the atoms as well as that of the uranium isotopes, the handling of the uranium isotope by means of the nuclear fission until arriving to the operation of the nuclear reactors. Chapter 2: The nuclear fuel cycle is described, the methods for its extraction, its conversion and its enrichment to arrive to the stages of the nuclear fuel management used in the reactors are described. Beginning by the radial design, the axial design and the core design of the nuclear reactor related with the fuel assemblies design. Chapter 3: the optimization methods of nuclear fuel previously used are exposed among those that are: the genetic algorithms method, the search methods based on heuristic rules and the application of the tabu search method, which was used for the development of this thesis. Chapter 4: In this part the used methodology to the

  12. Long term out-of-pile thermocouple tests in conditions representative for nuclear gas-cooled high temperature reactors

    Energy Technology Data Exchange (ETDEWEB)

    Laurie, M., E-mail: mathias.laurie@ec.europa.eu [European Commission, Joint Research Centre, Institut für Transurane, Hermann-von-Helmholtz-Platz 1, 76344, Eggenstein-Leopoldshafen (Germany); Fourrez, S. [THERMOCOAX SAS, Rue du Pré Neuf, 61100 Saint Georges des Groseillers (France); Fütterer, M.A.; Lapetite, J.M. [European Commission, Joint Research Centre, Institut für Transurane, Hermann-von-Helmholtz-Platz 1, 76344, Eggenstein-Leopoldshafen (Germany); Sadli, M.; Morice, R.; Failleau, G. [Laboratoire commun de métrologie LNE-Cnam, 61 rue du Landy, F-92310 Saint-Denis (France)

    2014-05-01

    During irradiation tests at high temperature failure of commercial Inconel 600 sheathed thermocouples is commonly encountered. As instrumentation, in particular thermocouples are considered safety-relevant both for irradiation tests and for commercial reactors, JRC and THERMOCOAX joined forces to solve this issue by performing out-of-pile tests with thermocouples mimicking the environment encountered by high temperature reactor (HTR) in-core instrumentation. The objective was to screen innovative sheathed thermocouples which would consecutively be tested under irradiation. Two such screening tests have been performed in high temperature environment (i.e. temperature in the range 1100–1150 °C) with purposely contaminated helium atmosphere (mainly CH{sub 4}, CO, CO{sub 2}, O{sub 2} impurities) representative for high temperature reactor carburizing atmospheres. The first set of thermocouples embedded in graphite (mainly conventional N type thermocouples and thermocouples with innovative sheaths) was tested in a dedicated furnace at THERMOCOAX lab with helium flushing. The second out-of-pile test at JRC with a partly different set of thermocouples replicated the original test for comparison. Performance indicators such as thermal drift, insulation resistance and loop resistance were monitored. Through these long-term screening tests the effect of several parameters were investigated: niobium sleeves, bending, diameter, sheath composition as well as the chemical environment. SEM examinations were performed to analyze local damage (bending zone, sheath). The present paper describes the two tests, sums up data collected during these tests in terms of thermocouple behavior and describes further instrumentation testing work with fixed point mini cells for qualification under irradiation.

  13. Exploring new coolants for nuclear breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lafuente, A., E-mail: anlafuente@etsii.upm.e [ETSII-UPM, c/Jose Gutierrez Abascal, 2, 28006 Madrid (Spain); Piera, M. [ETSII:UNED, c/Juan del Rosal, 12, 28040 Madrid (Spain)

    2010-06-15

    Breeder reactors are considered a unique tool for fully exploiting natural nuclear resources. In current Light Water Reactors (LWR), only 0.5% of the primary energy contained in the nuclei removed from a mine is converted into useful heat. The rest remains in the depleted uranium or spent fuel. The need to improve resource-efficiency has stimulated interest in Fast-Reactor-based fuel cycles, which can exploit a much higher fraction of the energy content of mined uranium by burning U-238, mainly after conversion into Pu-239. Thorium fuel cycles also offer several potential advantages over a uranium fuel cycle. The coolant initially selected for most of the FBR programs launched in the 1960s was sodium, which is still considered the best candidate for these reactors. However, Na-cooled FBRs have a positive void reactivity coefficient. Among other factors, this fundamental drawback has resulted in the canceled deployment of these reactors. Therefore, it seems reasonable to explore new options for breeder coolants. In this paper, a proposal is presented for a new molten salt (F{sub 2}Be) coolant that could overcome the safety issues related to the positive void reactivity coefficient of molten metal coolants. Although it is a very innovative proposal that would require an extensive R and D program, this paper presents the very appealing properties of this salt when using a specific type of fuel that is similar to that of pebble bed reactors. The F{sub 2}Be concept was studied over a typical MOX composition and extended to a thorium-based cycle. The general analysis took into account the requirements for criticality (opening the option of hybrid subcritical systems); the requirements for breeding; and the safety requirement of having a negative coolant void reactivity coefficient. A design window was found in the definition of a F{sub 2}Be cooled reactor where the safety requirement was met, unlike for molten metal-cooled reactors, which always have positive void

  14. Design of an Organic Simplified Nuclear Reactor

    OpenAIRE

    Koroush Shirvan; Eric Forrest

    2016-01-01

    Numerous advanced reactor concepts have been proposed to replace light water reactors ever since their establishment as the dominant technology for nuclear energy production. While most designs seek to improve cost competitiveness and safety, the implausibility of doing so with affordable materials or existing nuclear fuel infrastructure reduces the possibility of near-term deployment, especially in developing countries. The organic nuclear concept, first explored in the 1950s, offers an attr...

  15. Feasibility Study of Supercritical Light Water Cooled Reactors for Electric Power Production, Nuclear Energy Research Initiative Project 2001-001, Westinghouse Electric Co. Grant Number: DE-FG07-02SF22533, Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Philip E. MacDonald

    2005-01-01

    The supercritical water-cooled reactor (SCWR) is one of the six reactor technologies selected for research and development under the Generation IV program. SCWRs are promising advanced nuclear systems because of their high thermal efficiency (i.e., about 45% versus about 33% efficiency for current Light Water Reactors [LWRs]) and considerable plant simplification. SCWRs are basically LWRs operating at higher pressure and temperatures with a direct once-through cycle. Operation above the critical pressure eliminates coolant boiling, so the coolant remains single-phase throughout the system. Thus, the need for a pressurizer, steam generators, steam separators, and dryers is eliminated. The main mission of the SCWR is generation of low-cost electricity. It is built upon two proven technologies: LWRs, which are the most commonly deployed power generating reactors in the world, and supercritical fossil-fired boilers, a large number of which are also in use around the world. The reference SCWR design for the U.S. program is a direct cycle system operating at 25.0 MPa, with core inlet and outlet temperatures of 280 and 500 C, respectively. The coolant density decreases from about 760 kg/m3 at the core inlet to about 90 kg/m3 at the core outlet. The inlet flow splits with about 10% of the inlet flow going down the space between the core barrel and the reactor pressure vessel (the downcomer) and about 90% of the inlet flow going to the plenum at the top of the rector pressure vessel, to then flow down through the core in special water rods to the inlet plenum. Here it mixes with the feedwater from the downcomer and flows upward to remove the heat in the fuel channels. This strategy is employed to provide good moderation at the top of the core. The coolant is heated to about 500 C and delivered to the turbine. The purpose of this NERI project was to assess the reference U.S. Generation IV SCWR design and explore alternatives to determine feasibility. The project was

  16. Testing of Liquid Metal Components for Nuclear Surface Power Systems

    Science.gov (United States)

    Polzin, K. A.; Pearson, J. B.; Godfroy, T. J.; Schoenfeld, M.; Webster, K.; Briggs, M. H.; Geng, S. M.; Adkins, H. E.; Werner, J. E.

    2010-01-01

    The capability to perform testing at both the module/component level and in near prototypic reactor configurations using a non-nuclear test methodology allowed for evaluation of two components critical to the development of a potential nuclear fission power system for the lunar surface. A pair of 1 kW Stirling power convertors, similar to the type that would be used in a reactor system to convert heat to electricity, were integrated into a reactor simulator system to determine their performance using pumped NaK as the hot side working fluid. The performance in the pumped-NaK system met or exceed the baseline performance measurements where the converters were electrically heated. At the maximum hot-side temperature of 550 C the maximum output power was 2375 watts. A specially-designed test apparatus was fabricated and used to quantify the performance of an annular linear induction pump that is similar to the type that could be used to circulate liquid metal through the core of a space reactor system. The errors on the measurements were generally much smaller than the magnitude of the measurements, permitting accurate performance evaluation over a wide range of operating conditions. The pump produced flow rates spanning roughly 0.16 to 5.7 l/s (2.5 to 90 GPM), and delta p levels from less than 1 kPa to 90 kPa (greater than 0.145 psi to roughly 13 psi). At the nominal FSP system operating temperature of 525 C the maximum efficiency was just over 4%.

  17. Nuclear reactor kinetics and plant control

    CERN Document Server

    Oka, Yoshiaki

    2013-01-01

    Understanding time-dependent behaviors of nuclear reactors and the methods of their control is essential to the operation and safety of nuclear power plants. This book provides graduate students, researchers, and engineers in nuclear engineering comprehensive information on both the fundamental theory of nuclear reactor kinetics and control and the state-of-the-art practice in actual plants, as well as the idea of how to bridge the two. The first part focuses on understanding fundamental nuclear kinetics. It introduces delayed neutrons, fission chain reactions, point kinetics theory, reactivit

  18. The NASA Environmentally Responsible Aviation Project/General Electric Open Rotor Test Campaign

    Science.gov (United States)

    Van Zante, Dale

    2013-01-01

    The Open Rotor is a modern version of the UnDucted Fan (UDF) that was flight tested in the late 1980's through a partnership between NASA and General Electric (GE). Tests were conducted in the 9'x15' Low Speed Wind Tunnel and the 8'x6' Supersonic Wind Tunnel starting in late 2009 and completed in early 2012. Aerodynamic and acoustic data were obtained for takeoff, approach and cruise simulations. GE was the primary partner, but other organizations were involved such as Boeing and Airbus who provided additional hardware for fuselage simulations. This test campaign provided the acoustic and performance characteristics for modern open rotor blades designs." NASA and GE conducted joint systems analysis to evaluate how well new blade designs would perform on a B737 class aircraft, and compared the results to an advanced higher bypass ratio turbofan." Acoustic shielding experiments were performed at NASA GRC and Boeing LSAF facilities to provide data for noise estimates of unconventional aircraft configurations with Open Rotor propulsion systems." The work was sponsored by NASA's aeronautics programs, including the Subsonic Fixed Wing (SFW) and the Environmentally Responsible Aviation (ERA) projects."

  19. Exploring new coolants for nuclear breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lafuente, A. [ETSI Industriales-Universidad Politecnica de Madrid, C/Jose Gutierrez Abascal, 2. 28006 Madrid (Spain)

    2010-07-01

    Breeder reactors are considered the unique tool for fully exploiting the natural nuclear resources. In current LWR, only a 0.5% of the primary energy contained in the nuclei removed from the mine is converted into useful heat, with the rest remaining in the depleted uranium or in the spent fuel. The objective of resource-efficiency stimulated the interest in Fast- Reactor-based fuel cycles which can exploit a much higher fraction of the energy content of the mined uranium by burning U-238, mainly after conversion into Pu-239. Thorium fuel cycles would also offers several potential advantages over a uranium fuel cycle. The coolant initially chosen for most of the FBR programs launched in the 60's was sodium, which still is considered the best candidate for these reactors. However, Na-cooled FBR have a positive void reactivity coefficient, which has been among others, a fundamental drawback that has cancelled the deployment of these reactors. Therefore, it seems reasonable to explore totally new options on coolants for breeders. In this paper, a proposal is presented on a new molten salt (F{sub 2}Be) coolant that could overcome the safety issues related to the positive void reactivity coefficient of molten metal coolants. Although it is a very innovative proposal that would need an extensive R and D programme, this paper presents the very appealing properties of this salt, in the case of using a specific type of fuel, similar to that of pebble bed reactors. The concept will be studied over a typical MOX composition and extended to a Thorium-based cycle. The general analysis takes into account requirements for criticality (opening the option of hybrid subcritical systems); requirements for breeding; and the safety requirement of having a negative coolant void reactivity coefficient. A design window is found in the definition of a F{sub 2}Be cooled reactor where the safety requirement is met, unlike for molten metal cooled reactors which always have positive void

  20. Nuclear research reactors activities in INVAP

    Energy Technology Data Exchange (ETDEWEB)

    Ordonez, Juan Pablo [INVAP, Bariloche (Argentina)

    2013-07-01

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

  1. Cumulative damage fatigue tests on nuclear reactor Zircaloy-2 fuel tubes at room temperature and 300°C

    Science.gov (United States)

    Pandarinathan, P. R.; Vasudevan, P.

    1980-06-01

    Cumulative damage fatigue tests were conducted on the Zircaloy-2 fuel tubes at room temperature and 300°C on the modified Moore type, four-point-loaded, deflection-controlled, rotating bending fatigue testing machine. The cumulative cycle ratio at fracture for the Zircaloy-2 fuel tubes was found to depend on the sequence of loading, stress history, number of cycles of application of the pre-stress and the test temperature. A Hi-Lo type fatigue loading was found to be very much damaging at room temperature and this feature was not observed in the tests at 300°C. Results indicate significant differences in damage interaction and damage propagation under cumulative damage tests at room temperature and at 300°C. Block-loading fatigue tests are suggested as the best method to determine the life-time of Zircaloy-2 fuel tubes under random fatigue loading during their service in the reactor.

  2. The 4th surveillance test and evaluation of the reactor pressure vessel material (capsule W) of Younggwang nuclear power plant unit1

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byoung Chul; Lee, Sam Lai; Choi, Kwon Jae [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2001-08-01

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 4th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejeon after the capsule was transported from Yonggwang site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Yonggwang unit 1 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules U, V, X and W are 5.555E+18, 1.662E+19, 3.358E+19, and 4.521E+19 n/cm{sup 2}, respectively. The bias factor, the ratio of measurement versus calculation, was 0.859 for the 1st through 4th testing and the calculational uncertainty, 11.80% satisfied the requirement of USNRC Reg.Guide 1.190, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.551E+19 n/cm{sup 2} based on the end of 12th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 40, 48 and 56EFPY would reach 3.929E+19, 4.880E+19, 5.831E+19 and 6.782E+19 n/cm{sup 2} based on the current calculation. The result through this analysis for Yonggwang unit 1 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life. 4 refs., 41 figs., 35 tabs. (Author)

  3. Final report for the 5th surveillance test of the reactor pressure vessel material (capsule Y) of Yonggwang Nuclear Power Plant unit 2

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sam Lai; Kim, ByoungChul; Chang, Kee Ok (and others)

    2006-02-15

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 5th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Daejeon after the capsule was transported from Yonggwang site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Yonggwang unit 2 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules U, V, X, W and Y are 5.777E+18, 1.5371E+19, 3.7634E+19, 4.3045E+19, and 4.8662E+19n/cm{sup 2}, respectively. The bias factor, the ratio of calculation/measurement, was 0.953 for the 1st through 5th testing and the calculational uncertainty,7.2% satisfied the requirement of USNRC Reg.Guide 1.190, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.659E+19n/cm{sup 2} based on the end of 13th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 48, 56 and 64EFPY would reach 3.625E+19, 5.293E+19, 6.127E+19 and 6.960E+19n/cm{sup 2} based on the current calculation. The result through this analysis for Yonggwang unit 2 showed that there would be no problem for the Pressurized Thermal Shock(PTS) during the operation until design life.

  4. The 4th surveillance test and evaluation of the reactor pressure vessel material (capsule W) of Yonggwang nuclear power plant unit 2

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai; Choi, Kwon Jae; Gong, Un Sik; Chang, Jong Hwa; Joo, Yong Sun; Ahn, Sang Bok; Hong, Joon Hwa [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2001-02-01

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 4th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejeon after the capsule was transported from Kori site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Yonggwang unit 2 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules U, V, X and W are 5.762E+18, 1.5391E+19, 3.5119E+19, and 4.2610E+19 n/cm{sup 2}, respectively. The bias factor, the ratio of measurement versus calculation, was 0.899 for the 1st through 4th testing and the calculational uncertainty, 12.3% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.357E+19 n/cm{sup 2} based on the end of 11th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 40, 48 and 56EFPY would reach 3.525E+19, 4.337E+19, 5.148E+19 and 5.960E+19 n/cm{sup 2} based on the current calculation. The result through this analysis for Yonggwang unit 2 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life. 48 refs., 35 figs., 42 tabs. (Author)

  5. SCW Pressure-Channel Nuclear Reactor Some Design Features

    Science.gov (United States)

    Pioro, Igor L.; Khan, Mosin; Hopps, Victory; Jacobs, Chris; Patkunam, Ruban; Gopaul, Sandeep; Bakan, Kurtulus

    Concepts of nuclear reactors cooled with water at supercritical pressures were studied as early as the 1950s and 1960s in the USA and Russia. After a 30-year break, the idea of developing nuclear reactors cooled with SuperCritical Water (SCW) became attractive again as the ultimate development path for water cooling. The main objectives of using SCW in nuclear reactors are: 1) to increase the thermal efficiency of modern Nuclear Power Plants (NPPs) from 30-35% to about 45-48%, and 2) to decrease capital and operational costs and hence decrease electrical energy costs (˜1000 US/kW or even less). SCW NPPs will have much higher operating parameters compared to modern NPPs (pressure about 25 MPa and outlet temperature up to 625°C), and a simplified flow circuit, in which steam generators, steam dryers, steam separators, etc., can be eliminated. Also, higher SCW temperatures allow direct thermo-chemical production of hydrogen at low cost, due to increased reaction rates. Pressure-tube or pressure-channel SCW nuclear reactor concepts are being developed in Canada and Russia for some time. Some design features of the Canadian concept related to fuel channels are discussed in this paper. The main conclusion is that the development of SCW pressure-tube nuclear reactors is feasible and significant benefits can be expected over other thermal-energy systems.

  6. Hysteresis phenomenon in nuclear reactor dynamics

    Energy Technology Data Exchange (ETDEWEB)

    Pirayesh, Behnam; Pazirandeh, Ali [Islamic Azad Univ., Tehran (Iran, Islamic Republic of). Dept. of Nuclear Engineering, Science and Research Branch; Akbari, Monireh [Shahid Rajaee Teacher Training Univ., Tehran (Iran, Islamic Republic of). Dept. of Mathematics

    2017-05-15

    This paper applies a nonlinear analysis method to show that hysteresis phenomenon, due to the Saddle-node bifurcation, may occur in the nuclear reactor. This phenomenon may have significant effects on nuclear reactor dynamics and can even be the beginning of a nuclear reactor accident. A system of four dimensional nonlinear ordinary differential equations was considered to study the hysteresis phenomenon in a typical nuclear reactor. It should be noted that the reactivity was considered as a nonlinear function of state variables. The condition for emerging hysteresis was investigated using Routh-Hurwitz criterion and Sotomayor's theorem for saddle node bifurcation. A numerical analysis is also provided to illustrate the analytical results.

  7. Autonomous Control of Space Nuclear Reactors Project

    Data.gov (United States)

    National Aeronautics and Space Administration — Nuclear reactors to support future lunar and Mars robotic and manned missions impose new and innovative technological requirements for their control and protection...

  8. Design of an organic simplified nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shirvan, Koroush [Dept. of Nuclear Science and Engineering, Massachusetts Institute of Technology, Cambridge (United States); Forrest, Eric [Primary Standards Laboratory, Sandia National Laboratories, Albuquerque (United States)

    2016-08-15

    Numerous advanced reactor concepts have been proposed to replace light water reactors ever since their establishment as the dominant technology for nuclear energy production. While most designs seek to improve cost competitiveness and safety, the implausibility of doing so with affordable materials or existing nuclear fuel infrastructure reduces the possibility of near-term deployment, especially in developing countries. The organic nuclear concept, first explored in the 1950s, offers an attractive alternative to advanced reactor designs being considered. The advent of high temperature fluids, along with advances in hydrocracking and reforming technologies driven by the oil and gas industries, make the organic concept even more viable today. We present a simple, cost-effective, and safe small modular nuclear reactor for offshore underwater deployment. The core is moderated by graphite, zirconium hydride, and organic fluid while cooled by the organic fluid. The organic coolant enables operation near atmospheric pressure and use of plain carbon steel for the reactor tank and primary coolant piping system. The core is designed to mitigate the coolant degradation seen in early organic reactors. Overall, the design provides a power density of 40 kW/L, while reducing the reactor hull size by 40% compared with a pressurized water reactor while significantly reducing capital plant costs.

  9. MODERATOR ELEMENTS FOR UNIFORM POWER NUCLEAR REACTOR

    Science.gov (United States)

    Balent, R.

    1963-03-12

    This patent describes a method of obtaining a flatter flux and more uniform power generation across the core of a nuclear reactor. The method comprises using moderator elements having differing moderating strength. The elements have an increasing amount of the better moderating material as a function of radial and/or axial distance from the reactor core center. (AEC)

  10. Design of an Organic Simplified Nuclear Reactor

    Directory of Open Access Journals (Sweden)

    Koroush Shirvan

    2016-08-01

    Full Text Available Numerous advanced reactor concepts have been proposed to replace light water reactors ever since their establishment as the dominant technology for nuclear energy production. While most designs seek to improve cost competitiveness and safety, the implausibility of doing so with affordable materials or existing nuclear fuel infrastructure reduces the possibility of near-term deployment, especially in developing countries. The organic nuclear concept, first explored in the 1950s, offers an attractive alternative to advanced reactor designs being considered. The advent of high temperature fluids, along with advances in hydrocracking and reforming technologies driven by the oil and gas industries, make the organic concept even more viable today. We present a simple, cost-effective, and safe small modular nuclear reactor for offshore underwater deployment. The core is moderated by graphite, zirconium hydride, and organic fluid while cooled by the organic fluid. The organic coolant enables operation near atmospheric pressure and use of plain carbon steel for the reactor tank and primary coolant piping system. The core is designed to mitigate the coolant degradation seen in early organic reactors. Overall, the design provides a power density of 40 kW/L, while reducing the reactor hull size by 40% compared with a pressurized water reactor while significantly reducing capital plant costs.

  11. Evaluation of ductile-brittle transition behavior with neutron irradiation in nuclear reactor pressure vessel steels using small punch test

    Energy Technology Data Exchange (ETDEWEB)

    Kim, M. C.; Lee, B. S. [KAERI, Taejon (Korea, Republic of); Oh, Y. J. [Hanbat National Univ., Taejon (Korea, Republic of)

    2003-10-01

    A Small Punch (SP) test was performed to evaluate the ductile-brittle transition temperature before and after neutron irradiation in Reactor Pressure Vessel (RPV) steels produced by different manufacturing (refining) processes. The results were compared to the standard transition temperature shifts from the Charpy test and Master Curve fracture toughness test in accordance with the ASTM standard E1921. The samples were taken from 1/4t location of the vessel thickness and machined into a 10x10x0.5mm dimension. Irradiation of the samples was carried out in the research reactor at KAERI (HANARO) at about 290 .deg. C of the different fluence levels respectively. SP tests were performed in the temperature range of RT to -196 .deg. C using a 2.4mm diameter ball. For the materials before and after irradiation, SP transition temperatures (T{sub sp}), which are determined at the middle of the upper and lower SP energies, showed a linear correlation with the Charpy index temperature, T{sub 41J}. T{sub sp} from the irradiated samples was increased as the fluence level increased and was well within the deviation range of the unirradiated data. The TSP had a correlation with the reference temperature (T{sub 0}) from the master curve method using a pre-cracked Charpy V-notched (PCVN) specimen.

  12. Primary loop simulation of the SP-100 space nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Borges, Eduardo M.; Braz Filho, Francisco A.; Guimaraes, Lamartine N.F., E-mail: eduardo@ieav.cta.b, E-mail: fbraz@ieav.cta.b, E-mail: guimarae@ieav.cta.b [Instituto de Estudos Avancados (IEAv/DCTA) Sao Jose dos Campos, SP (Brazil)

    2011-07-01

    Between 1983 and 1992 the SP-100 space nuclear reactor development project for electric power generation in a range of 100 to 1000 kWh was conducted in the USA. Several configurations were studied to satisfy different mission objectives and power systems. In this reactor the heat is generated in a compact core and refrigerated by liquid lithium, the primary loops flow are controlled by thermoelectric electromagnetic pumps (EMTE), and thermoelectric converters produce direct current energy. To define the system operation point for an operating nominal power, it is necessary the simulation of the thermal-hydraulic components of the space nuclear reactor. In this paper the BEMTE-3 computer code is used to EMTE pump design performance evaluation to a thermalhydraulic primary loop configuration, and comparison of the system operation points of SP-100 reactor to two thermal powers, with satisfactory results. (author)

  13. Nonlinear Ultrasonic Measurements in Nuclear Reactor Environments

    Science.gov (United States)

    Reinhardt, Brian T.

    Several Department of Energy Office of Nuclear Energy (DOE-NE) programs, such as the Fuel Cycle Research and Development (FCRD), Advanced Reactor Concepts (ARC), Light Water Reactor Sustainability, and Next Generation Nuclear Power Plants (NGNP), are investigating new fuels, materials, and inspection paradigms for advanced and existing reactors. A key objective of such programs is to understand the performance of these fuels and materials during irradiation. In DOE-NE's FCRD program, ultrasonic based technology was identified as a key approach that should be pursued to obtain the high-fidelity, high-accuracy data required to characterize the behavior and performance of new candidate fuels and structural materials during irradiation testing. The radiation, high temperatures, and pressure can limit the available tools and characterization methods. In this thesis, two ultrasonic characterization techniques will be explored. The first, finite amplitude wave propagation has been demonstrated to be sensitive to microstructural material property changes. It is a strong candidate to determine fuel evolution; however, it has not been demonstrated for in-situ reactor applications. In this thesis, finite amplitude wave propagation will be used to measure the microstructural evolution in Al-6061. This is the first demonstration of finite amplitude wave propagation at temperatures in excess of 200 °C and during an irradiation test. Second, a method based on contact nonlinear acoustic theory will be developed to identify compressed cracks. Compressed cracks are typically transparent to ultrasonic wave propagation; however, by measuring harmonic content developed during finite amplitude wave propagation, it is shown that even compressed cracks can be characterized. Lastly, piezoelectric transducers capable of making these measurements are developed. Specifically, three piezoelectric sensors (Bismuth Titanate, Aluminum Nitride, and Zinc Oxide) are tested in the Massachusetts

  14. Mathematical Modeling for Simulation of Nuclear Reactor Analysis

    OpenAIRE

    Salah Ud-Din Khan; Shahab Ud-Din Khan

    2013-01-01

    In this paper, we have developed a mathematical model for the nuclear reactor analysis to be implemented in the nuclear reactor code. THEATRe is nuclear reactor analysis code which can only work for the cylindrical type fuel reactor and cannot applicable for the plate type fuel nuclear reactor. Therefore, the current studies encompasses on the modification of THEATRe code for the plate type fuel element. This mathematical model is applicable to the thermal analysis of the reactor which is ver...

  15. Electrical system regulations of the IEA-R1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mello, Jose Roberto de; Madi Filho, Tufic, E-mail: jrmello@ipen.br, E-mail: tmfilho@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2013-07-01

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

  16. Structural integrity of nuclear reactor pressure vessels

    Science.gov (United States)

    Knott, John F.

    2013-09-01

    The paper starts from concerns expressed by Sir Alan Cottrell, in the early 1970s, related to the safety of the pressurized water reactor (PWR) proposed at that time for the next phase of electrical power generation. It proceeds to describe the design and operation of nuclear generation plant and gives details of the manufacture of PWR reactor pressure vessels (RPVs). Attention is paid to stress-relief cracking and under-clad cracking, experienced with early RPVs, explaining the mechanisms for these forms of cracking and the means taken to avoid them. Particular note is made of the contribution of non-destructive inspection to structural integrity. Factors affecting brittle fracture in RPV steels are described: in particular, effects of neutron irradiation. The use of fracture mechanics to assess defect tolerance is explained, together with the failure assessment diagram embodied in the R6 procedure. There is discussion of the Master Curve and how it incorporates effects of irradiation on fracture toughness. Dangers associated with extrapolation of data to low probabilities are illustrated. The treatment of fatigue-crack growth is described, in the context of transients that may be experienced in the operation of PWR plant. Detailed attention is paid to the thermal shock associated with a large loss-of-coolant accident. The final section reviews the arguments advanced to justify 'Incredibility of Failure' and how these are incorporated in assessments of the integrity of existing plant and proposed 'new build' PWR pressure vessels.

  17. Irradiation tests of texture controlled cladding for pressurized water reactor in foreign reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ukawa, Kazunori [Hokkaido Electric Power Co., Inc., Sapporo (Japan); Abeta, Sadaaki

    1996-12-01

    PWR electric power companies and makers are promoting a high burn-up program for nuclear fuel aiming at reducing spent fuel. PCI tolerance performance of fuel clad is desirable to promote a high burnup program. It is clear that when we change the direction of a Zircaloy crystal structure by improving a manufacture process (texture control), PCI tolerance performance greatly improved. In this study, an improved clad was burnt in a R2 reactor to 61.5 GWd/t and a power ramp test was carried out in Sweden. Based on irradiation data, power ramp test data and post irradiation examination data, improvement of PCI tolerance performance was confirmed. (author)

  18. U.S. Nuclear Regulatory Commission

    Science.gov (United States)

    ... to page Search NRC Report a Safety Concern Nuclear Reactors Power Reactors Research & Test Reactors Operating Reactors ... Reactors Operator Licensing for New Reactors Research Activities Nuclear Reactor Quick Links Nuclear Materials Special Nuclear Material ...

  19. Improving the performance of the Egyptian second testing nuclear research reactor using interval type-2 fuzzy logic controller tuned by modified biogeography-based optimization

    Energy Technology Data Exchange (ETDEWEB)

    Sayed, M.M., E-mail: M.M.Sayed@ieee.org; Saad, M.S.; Emara, H.M.; Abou El-Zahab, E.E.

    2013-09-15

    Highlights: • A modified version of the BBO was proposed. • A novel method for interval type-2 FLC design tuned by MBBO was proposed. • The performance of the ETRR-2 was improved by using IT2FLC tuned by MBBO. -- Abstract: Power stabilization is a critical issue in nuclear reactors. The conventional proportional derivative (PD) controller is currently used in the Egyptian second testing research reactor (ETRR-2). In this paper, we propose a modified biogeography-based optimization (MBBO) algorithm to design the interval type-2 fuzzy logic controller (IT2FLC) to improve the performance of the Egyptian second testing research reactor (ETRR-2). Biogeography-based optimization (BBO) is a novel evolutionary algorithm that is based on the mathematical models of biogeography. Biogeography is the study of the geographical distribution of biological organisms. In the BBO model, problem solutions are represented as islands, and the sharing of features between solutions is represented as immigration and emigration between the islands. A modified version of the BBO is applied to design the IT2FLC to get the optimal parameters of the membership functions of the controller. We test the optimal IT2FLC obtained by modified biogeography-based optimization (MBBO) using the integral square error (ISE) and is compared with the currently used PD controller.

  20. Electrochemistry of Water-Cooled Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Macdonald, Dgiby; Urquidi-Macdonald, Mirna; Pitt, Jonathan

    2006-08-08

    This project developed a comprehensive mathematical and simulation model for calculating thermal hydraulic, electrochemical, and corrosion parameters, viz. temperature, fluid flow velocity, pH, corrosion potential, hydrogen injection, oxygen contamination, stress corrosion cracking, crack growth rate, and other important quantities in the coolant circuits of water-cooled nuclear power plants, including both Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs). The model is being used to assess the three major operational problems in Pressurized Water Reactors (PWR), which include mass transport, activity transport, and the axial offset anomaly, and provide a powerful tool for predicting the accumulation of SCC damage in BWR primary coolant circuits as a function of operating history. Another achievement of the project is the development of a simulation tool to serve both as a training tool for plant operators and as an engineering test-bed to evaluate new equipment and operating strategies (normal operation, cold shut down and others). The development and implementation of the model allows us to estimate the activity transport or "radiation fields" around the primary loop and the vessel, as a function of the operating parameters and the water chemistry.

  1. Reactor Simulator Integration and Testing

    Science.gov (United States)

    Schoenfield, M. P.; Webster, K. L.; Pearson, J. B.

    2013-01-01

    As part of the Nuclear Systems Office Fission Surface Power Technology Demonstration Unit (TDU) project, a reactor simulator (RxSim) test loop was designed and built to perform integrated testing of the TDU components. In particular, the objectives of RxSim testing were to verify the operation of the core simulator, the instrumentation and control system, and the ground support gas and vacuum test equipment. In addition, it was decided to include a thermal test of a cold trap purification design and a pump performance test at pump voltages up to 150 V because the targeted mass flow rate of 1.75 kg/s was not obtained in the RxSim at the originally constrained voltage of 120 V. This Technical Memorandum summarizes RxSim testing. The gas and vacuum ground support test equipment performed effectively in NaK fill, loop pressurization, and NaK drain operations. The instrumentation and control system effectively controlled loop temperature and flow rates or pump voltage to targeted settings. The cold trap design was able to obtain the targeted cold temperature of 480 K. An outlet temperature of 636 K was obtained, which was lower than the predicted 750 K but 156 K higher than the cold temperature, indicating the design provided some heat regeneration. The annular linear induction pump tested was able to produce a maximum flow rate of 1.53 kg/s at 800 K when operated at 150 V and 53 Hz.

  2. Reactivity control assembly for nuclear reactor. [LMFBR

    Science.gov (United States)

    Bollinger, L.R.

    1982-03-17

    This invention, which resulted from a contact with the United States Department of Energy, relates to a control mechanism for a nuclear reactor and, more particularly, to an assembly for selectively shifting different numbers of reactivity modifying rods into and out of the core of a nuclear reactor. It has been proposed heretofore to control the reactivity of a breeder reactor by varying the depth of insertion of control rods (e.g., rods containing a fertile material such as ThO/sub 2/) in the core of the reactor, thereby varying the amount of neutron-thermalizing coolant and the amount of neutron-capturing material in the core. This invention relates to a mechanism which can advantageously be used in this type of reactor control system.

  3. The role of nuclear reactors in space exploration and development

    Energy Technology Data Exchange (ETDEWEB)

    Lipinski, R.J.

    2000-07-01

    The United States has launched more than 20 radioisotopic thermoelectric generators (RTGs) into space over the past 30 yr but has launched only one nuclear reactor, and that was in 1965. Russia has launched more than 30 reactors. The RTGs use the heat of alpha decay of {sup 238}Pu for power and typically generate <1 kW of electricity. Apollo, Pioneer, Voyager, Viking, Galileo, Ulysses, and Cassini all used RTGs. Space reactors use the fission energy of {sup 235}U; typical designs are for 100 to 1000 kW of electricity. The only US space reactor launch (SNAP-10A) was a demonstration mission. One reason for the lack of space reactor use by the United States was the lack of space missions that required high power. But, another was the assumed negative publicity that would accompany a reactor launch. The net result is that all space reactor programs after 1970 were terminated before an operating space reactor could be developed, and they are now many years from recovering the ability to build them. Two major near-term needs for space reactors are the human exploration of Mars and advanced missions to and beyond the orbit of Jupiter. To help obtain public acceptance of space reactors, one must correct some of the misconceptions concerning space reactors and convey the following facts to the public and to decision makers: Space reactors are 1000 times smaller in power and size than a commercial power reactor. A space reactor at launch is only as radioactive as a pile of dirt 60 m (200 ft) across. A space reactor contains no plutonium at launch. It does not become significantly radioactive until it is turned on, and it will be engineered so that no launch accident can turn it on, even if that means fueling it after launch. The reactor will not be turned on until it is in a high stable orbit or even on an earth-escape trajectory for some missions. The benefits of space reactors are that they give humanity a stairway to the planets and perhaps the stars. They open a new

  4. Nuclear Power from Fission Reactors. An Introduction.

    Science.gov (United States)

    Department of Energy, Washington, DC. Technical Information Center.

    The purpose of this booklet is to provide a basic understanding of nuclear fission energy and different fission reaction concepts. Topics discussed are: energy use and production, current uses of fuels, oil and gas consumption, alternative energy sources, fossil fuel plants, nuclear plants, boiling water and pressurized water reactors, the light…

  5. U.S. Nuclear Power Reactor Plant Status

    Data.gov (United States)

    Nuclear Regulatory Commission — Demographic data on U.S. commercial nuclear power reactors, including: plant name/unit number, docket number, location, licensee, reactor/containment type, nuclear...

  6. Reactor neutrons in nuclear astrophysics

    Science.gov (United States)

    Reifarth, René; Glorius, Jan; Göbel, Kathrin; Heftrich, Tanja; Jentschel, Michael; Jurado, Beatriz; Käppeler, Franz; Köster, Ulli; Langer, Christoph; Litvinov, Yuri A.; Weigand, Mario

    2017-09-01

    The huge neutron fluxes offer the possibility to use research reactors to produce isotopes of interest, which can be investigated afterwards. An example is the half-lives of long-lived isotopes like 129I. A direct usage of reactor neutrons in the astrophysical energy regime is only possible, if the corresponding ions are not at rest in the laboratory frame. The combination of an ion storage ring with a reactor and a neutron guide could open the path to direct measurements of neutron-induced cross sections on short-lived radioactive isotopes in the astrophysically interesting energy regime.

  7. Test reactor risk assessment methodology

    Energy Technology Data Exchange (ETDEWEB)

    Jennings, R.H.; Rawlins, J.K.; Stewart, M.E.

    1976-04-01

    A methodology has been developed for the identification of accident initiating events and the fault modeling of systems, including common mode identification, as these methods are applied in overall test reactor risk assessment. The methods are exemplified by a determination of risks to a loss of primary coolant flow in the Engineering Test Reactor.

  8. Nuclear reactor composite fuel assembly. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Burgess, D.M.; Cappiello, M.W.; Marr, D.R.; Omberg, R.P.

    1980-11-25

    A core and composite fuel assembly are described for a liquid-cooled breeder nuclear reactor including a plurality of elongated coextending driver and breeder fuel elements arranged to form a generally polygonal bundle within a thin-walled duct. The breeder elements are larger in cross section than the driver elements, and each breeder element is laterally bounded by a number of the driver elements. Each driver element further includes structure for spacing the driver elements from adjacent fuel elements and, where adjacent, the thin-walled duct. A core made up of the fuel elements can advantageously include fissile fuel of only one enrichment, while varying the effective enrichment of any given assembly or core region, merely by varying the relative number and size of the driver and breeder elements.

  9. Hybrid reactors: Nuclear breeding or energy production?

    Energy Technology Data Exchange (ETDEWEB)

    Piera, Mireia [UNED, ETSII-Dp Ingenieria Energetica, c/Juan del Rosal 12, 28040 Madrid (Spain); Lafuente, Antonio; Abanades, Alberto; Martinez-Val, J.M. [ETSII-UPM, c/Jose Gutierrez Abascal 2, 28006 Madrid (Spain)

    2010-09-15

    After reviewing the long-standing tradition on hybrid research, an assessment model is presented in order to characterize the hybrid performance under different objectives. In hybrids, neutron multiplication in the subcritical blanket plays a major role, not only for energy production and nuclear breeding, but also for tritium breeding, which is fundamental requirement in fusion-fission hybrids. All three objectives are better achieved with high values of the neutron multiplication factor (k-eff) with the obvious and fundamental limitation that it cannot reach criticality under any event, particularly, in the case of a loss of coolant accident. This limitation will be very important in the selection of the coolant. Some general considerations will be proposed, as guidelines for assessing the hybrid potential in a given scenario. Those guidelines point out that hybrids can be of great interest for the future of nuclear energy in a framework of Sustainable Development, because they can contribute to the efficient exploitation of nuclear fuels, with very high safety features. Additionally, a proposal is presented on a blanket specially suited for fusion-fission hybrids, although this reactor concept is still under review, and new work is needed for identifying the most suitable blanket composition, which can vary depending on the main objective of the hybrid. (author)

  10. NASA's Nuclear Frontier: The Plum Brook Reactor Facility, 1941-2002

    Science.gov (United States)

    Bowles, Mark D.; Arrighi, Robert S.

    2004-01-01

    In 1953, President Eisenhower delivered a speech called "Atoms for Peace" to the United Nations General Assembly. He described the emergence of the atomic age and the weapons of mass destruction that were piling up in the storehouses of the American and Soviet nations. Although neither side was aiming for global destruction, Eisenhower wanted to "move out of the dark chambers of horrors into the light, to find a way by which the minds of men, the hopes of men, the souls of men everywhere, can move towards peace and happiness and well-being." One way Eisenhower hoped this could happen was by transforming the atom from a weapon of war into a useful tool for civilization. Many people believed that there were unprecedented opportunities for peaceful nuclear applications. These included hopeful visions of atomic-powered cities, cars, airplanes, and rockets. Nuclear power might also serve as an efficient way to generate electricity in space to support life and machines. Eisenhower wanted to provide scientists and engineers with "adequate amounts of fission- able material with which to test and develop their ideas." But, in attempting to devise ways to use atomic power for peaceful purposes, scientists realized how little they knew about the nature and effects of radiation. As a result, the United States began constructing nuclear test reactors to enable scientists to conduct research by producing neutrons.

  11. Nuclear reactor (1960); Reacteurs nucleaires (1960)

    Energy Technology Data Exchange (ETDEWEB)

    Maillard, M.L. [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires; Leo, M.B. [Electricite de France (EDF), 75 - Paris (France)

    1960-07-01

    The first French plutonium-making reactors G1, G2 and G3 built at Marcoule research center are linked to a power plant. The G1 electrical output does not offset the energy needed for operating this reactor. On the contrary, reactors G2 and G3 will each generate a net power of 25 to 30 MW, which will go into the EDF grid. This power is relatively small, but the information obtained from operation is great and will be helpful for starting up the power reactor EDF1, EDF2 and EDF3. The paper describes how, previous to any starting-up operation, the tests performed, especially those concerned with the power plant and the pressure vessel, have helped to bring the commissioning date closer. (author) [French] Les premiers reacteurs industriels plutonigenes francais G1 - G2 - G3 du Centre de Marcoule comportent une installation de recuperation d'energie. La production d'electricite de G1 ne compense pas l'energie depensee par ailleurs pour le fonctionnement de l'ensemble, par contre, G2 et G3 doivent fournir chacun une puissance de 25 a 30 MW au reseau national d'Electricite de France. Cette puissance est modeste, mais l'experience acquise grace a ces reacteurs est tres grande et c'est grace a elle qu'il nous sera possible de mettre en exploitation les reacteurs energetiques EDF1 - EDF2 - EDF3. Le memoire decrit comment, avant tout demarrage du reacteur, les essais effectues, en particulier ceux concernant l'installation de recuperation d'energie et le caisson, ont permis d'abreger la phase de montee en puissance. (auteur)

  12. Nuclear modules for space electric propulsion

    Science.gov (United States)

    Difilippo, F. C.

    1998-01-01

    Analysis of interplanetary cargo and piloted missions requires calculations of the performances and masses of subsystems to be integrated in a final design. In a preliminary and scoping stage the designer needs to evaluate options iteratively by using fast computer simulations. The Oak Ridge National Laboratory (ORNL) has been involved in the development of models and calculational procedures for the analysis (neutronic and thermal hydraulic) of power sources for nuclear electric propulsion. The nuclear modules will be integrated into the whole simulation of the nuclear electric propulsion system. The vehicles use either a Brayton direct-conversion cycle, using the heated helium from a NERVA-type reactor, or a potassium Rankine cycle, with the working fluid heated on the secondary side of a heat exchanger and lithium on the primary side coming from a fast reactor. Given a set of input conditions, the codes calculate composition. dimensions, volumes, and masses of the core, reflector, control system, pressure vessel, neutron and gamma shields, as well as the thermal hydraulic conditions of the coolant, clad and fuel. Input conditions are power, core life, pressure and temperature of the coolant at the inlet of the core, either the temperature of the coolant at the outlet of the core or the coolant mass flow and the fluences and integrated doses at the cargo area. Using state-of-the-art neutron cross sections and transport codes, a database was created for the neutronic performance of both reactor designs. The free parameters of the models are the moderator/fuel mass ratio for the NERVA reactor and the enrichment and the pitch of the lattice for the fast reactor. Reactivity and energy balance equations are simultaneously solved to find the reactor design. Thermalhydraulic conditions are calculated by solving the one-dimensional versions of the equations of conservation of mass, energy, and momentum with compressible flow.

  13. Nuclear modules for space electric propulsion

    Energy Technology Data Exchange (ETDEWEB)

    Difilippo, F.C.

    1998-12-31

    Analysis of interplanetary cargo and piloted missions requires calculations of the performances and masses of subsystems to be integrated in a final design. In a preliminary and scoping stage the designer needs to evaluate options iteratively by using fast computer simulations. The Oak Ridge National Laboratory (ORNL) has been involved in the development of models and calculational procedures for the analysis (neutronic and thermal hydraulic) of power sources for nuclear electric propulsion. The nuclear modules will be integrated into the whole simulation of the nuclear electric propulsion system. The vehicles use either a Brayton direct-conversion cycle, using the heated helium from a NERVA-type reactor, or a potassium Rankine cycle, with the working fluid heated on the secondary side of a heat exchanger and lithium on the primary side coming from a fast reactor. Given a set of input conditions, the codes calculate composition. dimensions, volumes, and masses of the core, reflector, control system, pressure vessel, neutron and gamma shields, as well as the thermal hydraulic conditions of the coolant, clad and fuel. Input conditions are power, core life, pressure and temperature of the coolant at the inlet of the core, either the temperature of the coolant at the outlet of the core or the coolant mass flow and the fluences and integrated doses at the cargo area. Using state-of-the-art neutron cross sections and transport codes, a database was created for the neutronic performance of both reactor designs. The free parameters of the models are the moderator/fuel mass ratio for the NERVA reactor and the enrichment and the pitch of the lattice for the fast reactor. Reactivity and energy balance equations are simultaneously solved to find the reactor design. Thermalhydraulic conditions are calculated by solving the one-dimensional versions of the equations of conservation of mass, energy, and momentum with compressible flow. 10 refs., 1 tab.

  14. Space Nuclear Thermal Propulsion (SNTP) tests

    Science.gov (United States)

    Allen, George C.

    1993-01-01

    Viewgraphs on the space nuclear thermal propulsion (SNTP) program are presented. The objective of the research is to develop advanced nuclear thermal propulsion (NTP) technology based on the particle bed reactor concept. A strong philosophical commitment exists in the industry/national laboratory team to emphasize testing in development activities. Nuclear testing currently underway to support development of SNTP technology is addressed.

  15. Research on friction coefficient of nuclear Reactor Vessel Internals Hold Down Spring: Stress coefficient test analysis method

    Energy Technology Data Exchange (ETDEWEB)

    Linjun, Xie, E-mail: linjunx@zjut.edu.cn [College of Mechanical Engineering, Zhejiang University of Technology, Hangzhou 310014 (China); Guohong, Xue; Ming, Zhang [Shanghai Nuclear Engineering Research & Design Institute, Shanghai 200233 (China)

    2016-08-01

    Graphical abstract: HDS stress coefficient test apparatus. - Highlights: • This paper performs mathematic deduction to the physical model of Hold Down Spring (HDS), establishes a mathematic model of axial load P and stress, stress coefficient and friction coefficient and designs a set of test apparatuses for simulating the pretightening process of the HDS for the first time according to a model similarity criterion. • The mathematical relation between the load and the strain is obtained about the HDS, and the mathematical model of the stress coefficient and the friction coefficient is established. So, a set of test apparatuses for obtaining the stress coefficient is designed according to the model scaling criterion and the friction coefficient of the K1000 HDS is calculated to be 0.336 through the obtained stress coefficient. • The relation curve between the theoretical load and the friction coefficient is obtained through analysis and indicates that the change of the friction coefficient f would influence the pretightening load under the condition of designed stress. The necessary pretightening load in the design process is calculated to be 5469 kN according to the obtained friction coefficient. Therefore, the friction coefficient and the pretightening load under the design conditions can provide accurate pretightening data for the analysis and design of the reactor HDS according to the operations. - Abstract: This paper performs mathematic deduction to the physical model of Hold Down Spring (HDS), establishes a mathematic model of axial load P and stress, stress coefficient and friction coefficient and designs a set of test apparatuses for simulating the pretightening process of the HDS for the first time according to a model similarity criterion. By carrying out tests and researches through a stress testing technique, P–σ curves in loading and unloading processes of the HDS are obtained and the stress coefficient k{sub f} of the HDS is obtained. So, the

  16. Long term out-of-pile thermocouple tests in conditions representative for nuclear gas-cooled high temperature reactors

    Energy Technology Data Exchange (ETDEWEB)

    Laurie, M. [European Commission, Joint Research Centre, Inst. for Energy, P.O. Box 2, NL-1755 ZG Petten (Netherlands); Fourrez, S. [THERMOCOAX SAS, BP 26, Planquivon, F-61438 Flers Cedex (France); Fuetterer, M. A.; Lapetite, J. M. [European Commission, Joint Research Centre, Inst. for Energy, P.O. Box 2, NL-1755 ZG Petten (Netherlands)

    2011-07-01

    During irradiation tests at high temperature, failure of commercial Inconel 600 sheathed thermocouples is commonly encountered. To understand and remedy this problem, out-of-pile tests were performed with thermocouples in carburizing atmospheres which can be assumed to be at least locally representative for High Temperature Reactors. The objective was to screen those thermocouples which would consecutively be used under irradiation. Two such screening tests have been performed with a set of thermocouples embedded in graphite (mainly conventional Type N thermocouples and thermocouples with innovative sheaths) in a dedicated furnace with helium flushing. Performance indicators such as thermal drift, insulation and loop resistance were monitored and compared to those from conventional Type N thermocouples. Several parameters were investigated: niobium sleeves, bending, thickness, sheath composition, temperature as well as the chemical environment. After the tests, Scanning Electron Microscopy (SEM) examinations were performed to analyze possible local damage in wires and in the sheath. The present paper describes the two experiments, summarizes results and outlines further work, in particular to further analyze the findings and to select suitable thermocouples for qualification under irradiation. (authors)

  17. Thermionic reactors for space nuclear power

    Science.gov (United States)

    Griaznov, Georgii M.; Zhabotinskii, Evgenii E.; Serbin, Victor I.; Zrodnikov, Anatolii V.; Pupko, Victor Ia.; Ponomarev-Stepnoi, Nikolai N.; Usov, V. A.; Nikolaev, Iu. V.

    Compact thermionic nuclear reactor systems with satisfactory mass performance are competitive with space nuclear power systems based on the organic Rankine and closed Brayton cycles. The mass characteristics of the thermionic space nuclear power system are better than that of the solar power system for power levels beyond about 10 kWe. Longlife thermionic fuel element requirements, including their optimal dimensions, and common requirements for the in-core thermionic reactor design are formulated. Thermal and fast in-core thermionic reactors are considered and the ranges of their sensible use are discussed. Some design features of the fast in-core thermionic reactors cores (power range to 1 MWe) including a choice of coolants are discussed. Mass and dimensional performance for thermionic nuclear power reactor system are assessed. It is concluded that thermionic space nuclear power systems are promising power supplies for spacecrafts and that a single basic type of thermionic fuel element may be used for power requirements ranging to several hundred kWe.

  18. Integral and Separate Effects Tests for Thermal Hydraulics Code Validation for Liquid-Salt Cooled Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Peterson, Per

    2012-10-30

    The objective of the 3-year project was to collect integral effects test (IET) data to validate the RELAP5-3D code and other thermal hydraulics codes for use in predicting the transient thermal hydraulics response of liquid salt cooled reactor systems, including integral transient response for forced and natural circulation operation. The reference system for the project is a modular, 900-MWth Pebble Bed Advanced High Temperature Reactor (PB-AHTR), a specific type of Fluoride salt-cooled High temperature Reactor (FHR). Two experimental facilities were developed for thermal-hydraulic integral effects tests (IETs) and separate effects tests (SETs). The facilities use simulant fluids for the liquid fluoride salts, with very little distortion to the heat transfer and fluid dynamics behavior. The CIET Test Bay facility was designed, built, and operated. IET data for steady state and transient natural circulation was collected. SET data for convective heat transfer in pebble beds and straight channel geometries was collected. The facility continues to be operational and will be used for future experiments, and for component development. The CIET 2 facility is larger in scope, and its construction and operation has a longer timeline than the duration of this grant. The design for the CIET 2 facility has drawn heavily on the experience and data collected on the CIET Test Bay, and it was completed in parallel with operation of the CIET Test Bay. CIET 2 will demonstrate start-up and shut-down transients and control logic, in addition to LOFC and LOHS transients, and buoyant shut down rod operation during transients. Design of the CIET 2 Facility is complete, and engineering drawings have been submitted to an external vendor for outsourced quality controlled construction. CIET 2 construction and operation continue under another NEUP grant. IET data from both CIET facilities is to be used for validation of system codes used for FHR modeling, such as RELAP5-3D. A set of

  19. Advanced gas cooled nuclear reactor materials evaluation and development program

    Energy Technology Data Exchange (ETDEWEB)

    1977-01-01

    Results of work performed from January 1, 1977 through March 31, 1977 on the Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program are presented. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Process Heat and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (impure Helium), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in this report includes progress to date on alloy selection for VHTR Nuclear Process Heat (NPH) applications and for DCHT applications. The present status on the simulated reactor helium loop design and on designs for the testing and analysis facilities and equipment is discussed.

  20. Optimally moderated nuclear fission reactor and fuel source therefor

    Science.gov (United States)

    Ougouag, Abderrafi M.; Terry, William K.; Gougar, Hans D.

    2008-07-22

    An improved nuclear fission reactor of the continuous fueling type involves determining an asymptotic equilibrium state for the nuclear fission reactor and providing the reactor with a moderator-to-fuel ratio that is optimally moderated for the asymptotic equilibrium state of the nuclear fission reactor; the fuel-to-moderator ratio allowing the nuclear fission reactor to be substantially continuously operated in an optimally moderated state.

  1. Introduction to the neutron kinetics of nuclear power reactors

    CERN Document Server

    Tyror, J G; Grant, P J

    2013-01-01

    An Introduction to the Neutron Kinetics of Nuclear Power Reactors introduces the reader to the neutron kinetics of nuclear power reactors. Topics covered include the neutron physics of reactor kinetics, feedback effects, water-moderated reactors, fast reactors, and methods of plant control. The reactor transients following faults are also discussed, along with the use of computers in the study of power reactor kinetics. This book is comprised of eight chapters and begins with an overview of the reactor physics characteristics of a nuclear power reactor and their influence on system design and

  2. Advanced nuclear reactor types and technologies

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

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

  4. Oklo reactors and implications for nuclear science

    CERN Document Server

    Davis, E D; Sharapov, E I

    2014-01-01

    We summarize the nuclear physics interests in the Oklo natural nuclear reactors, focusing particularly on developments over the past two decades. Modeling of the reactors has become increasingly sophisticated, employing Monte Carlo simulations with realistic geometries and materials that can generate both the thermal and epithermal fractions. The water content and the temperatures of the reactors have been uncertain parameters. We discuss recent work pointing to lower temperatures than earlier assumed. Nuclear cross sections are input to all Oklo modeling and we discuss a parameter, the $^{175}$Lu ground state cross section for thermal neutron capture leading to the isomer $^{176\\mathrm{m}}$ Lu, that warrants further investigation. Studies of the time dependence of dimensionless fundamental constants have been a driver for much of the recent work on Oklo. We critically review neutron resonance energy shifts and their dependence on the fine structure constant $\\alpha$ and the ratio $X_q=m_q/\\Lambda$ (where $m_...

  5. The role of integral experiments and nuclear cross section evaluations in space nuclear reactor design

    Science.gov (United States)

    Moses, David L.; McKnight, Richard D.

    The importance of the nuclear and neutronic properties of candidate space reactor materials to the design process has been acknowledged as has been the use of benchmark reactor physics experiments to verify and qualify analytical tools used in design, safety, and performance evaluation. Since June 1966, the Cross Section Evaluation Working Group (CSEWG) has acted as an interagency forum for the assessment and evaluation of nuclear reaction data used in the nuclear design process. CSEWG data testing has involved the specification and calculation of benchmark experiments which are used widely for commercial reactor design and safety analysis. These benchmark experiments preceded the issuance of the industry standards for acceptance, but the benchmarks exceed the minimum acceptance criteria for such data. Thus, a starting place has been provided in assuring the accuracy and uncertainty of nuclear data important to space reactor applications.

  6. Scoping calculations of power sources for nuclear electric propulsion

    Energy Technology Data Exchange (ETDEWEB)

    Difilippo, F.C. [Oak Ridge National Lab., TN (United States)

    1994-05-01

    This technical memorandum describes models and calculational procedures to fully characterize the nuclear island of power sources for nuclear electric propulsion. Two computer codes were written: one for the gas-cooled NERVA derivative reactor and the other for liquid metal-cooled fuel pin reactors. These codes are going to be interfaced by NASA with the balance of plant in order to making scoping calculations for mission analysis.

  7. Uso de detectores de neutrinos para el monitoreo de reactores nucleares Uso de detectores de neutrinos para el monitoreo de reactores nucleares

    Directory of Open Access Journals (Sweden)

    Gerardo Moreno

    2012-02-01

    Full Text Available Se estudia la factibilidad del uso de los detectores de antineutrinos para el monitoreo de reactores nucleares. Usando un modelo sencillo de cascada de fisión a dos componentes, se ilustra la dependencia del número de antineutrinos detectados a una distancia L del reactor según la composición nuclear del combustible. Se explica el principio de detección de neutrinos de reactores en base al decaimiento beta inverso y se describe como los detectores de neutrinos pueden emplearse para el monitoreo de la producción de materiales fisibles en el reactor. Se comenta como generalizar este análisis al caso real de un reactor nuclear in situ y uno de los principales experimentos internacionales dedicados a este propósito. We study the feasibility to use antineutrinos detectors for monitoring of nuclear reactors. Using a simple model of fission shower with two components, we illustrate how the numbers of antineutrinos detected at a distance L from the reactor depend on the composition of the nuclear combustible. We explain the principles of reactor neutrino detection using inverse beta decays and we describe how neutrinos detectors can be used for monitoring the production of fissile materials within the reactors. We comment how to generalize this analysis to the realistic case of a nuclear reactor in situ and one of the main international experiments dedicated to study the use of neutrinos detectors as nuclear safeguards.

  8. Microbial fouling community analysis of the cooling water system of a nuclear test reactor with emphasis on sulphate reducing bacteria.

    Science.gov (United States)

    Balamurugan, P; Joshi, M Hiren; Rao, T S

    2011-10-01

    Culture and molecular-based techniques were used to characterize bacterial diversity in the cooling water system of a fast breeder test reactor (FBTR). Techniques were selected for special emphasis on sulphate-reducing bacteria (SRB). Water samples from different locations of the FBTR cooling water system, in addition to biofilm scrapings from carbon steel coupons and a control SRB sample were characterized. Whole genome extraction of the water samples and SRB diversity by group specific primers were analysed using nested PCR and denaturing gradient gel electrophoresis (DGGE). The results of the bacterial assay in the cooling water showed that the total culturable bacteria (TCB) ranged from 10(3) to 10(5) cfu ml(-1); iron-reducing bacteria, 10(3) to 10(5) cfu ml(-1); iron oxidizing bacteria, 10(2) to 10(3) cfu ml(-1) and SRB, 2-29 cfu ml(-1). However, the counts of the various bacterial types in the biofilm sample were 2-3 orders of magnitude higher. SRB diversity by the nested PCR-DGGE approach showed the presence of groups 1, 5 and 6 in the FBTR cooling water system; however, groups 2, 3 and 4 were not detected. The study demonstrated that the PCR protocol influenced the results of the diversity analysis. The paper further discusses the microbiota of the cooling water system and its relevance in biofouling.

  9. Modular Lead-Bismuth Fast Reactors in Nuclear Power

    Directory of Open Access Journals (Sweden)

    Vladimir Petrochenko

    2012-09-01

    Full Text Available On the basis of the unique experience of operating reactors with heavy liquid metal coolant–eutectic lead-bismuth alloy in nuclear submarines, the concept of modular small fast reactors SVBR-100 for civilian nuclear power has been developed and validated. The features of this innovative technology are as follows: a monoblock (integral design of the reactor with fast neutron spectrum, which can operate using different types of fuel in various fuel cycles including MOX fuel in a self-providing mode. The reactor is distinct in that it has a high level of self-protection and passive safety, it is factory manufactured and the assembled reactor can be transported by railway. Multipurpose application of the reactor is presumed, primarily, it can be used for regional power to produce electricity, heat and for water desalination. The Project is being realized within the framework of state-private partnership with joint venture OJSC “AKME-Engineering” established on a parity basis by the State Atomic Energy Corporation “Rosatom” and the Limited Liability Company “EuroSibEnergo”.

  10. Testing of an Integrated Reactor Core Simulator and Power Conversion System with Simulated Reactivity Feedback

    Science.gov (United States)

    Bragg-Sitton, Shannon M.; Hervol, David S.; Godfroy, Thomas J.

    2010-01-01

    A Direct Drive Gas-Cooled (DDG) reactor core simulator has been coupled to a Brayton Power Conversion Unit (BPCU) for integrated system testing at NASA Glenn Research Center (GRC) in Cleveland, Ohio. This is a closed-cycle system that incorporates an electrically heated reactor core module, turboalternator, recuperator, and gas cooler. Nuclear fuel elements in the gas-cooled reactor design are replaced with electric resistance heaters to simulate the heat from nuclear fuel in the corresponding fast spectrum nuclear reactor. The thermodynamic transient behavior of the integrated system was the focus of this test series. In order to better mimic the integrated response of the nuclear-fueled system, a simulated reactivity feedback control loop was implemented. Core power was controlled by a point kinetics model in which the reactivity feedback was based on core temperature measurements; the neutron generation time and the temperature feedback coefficient are provided as model inputs. These dynamic system response tests demonstrate the overall capability of a non-nuclear test facility in assessing system integration issues and characterizing integrated system response times and response characteristics.

  11. 26 CFR 1.468A-1T - Nuclear decommissioning costs; general rules (temporary).

    Science.gov (United States)

    2010-04-01

    ... nuclear power plant means any nuclear power reactor that is used predominantly in the trade or business of the furnishing or sale of electric energy. Each unit (that is, nuclear reactor) located on a multi..., systems and components of a nuclear power plant that has permanently ceased the production of...

  12. Development of nuclear thermal hydraulic verification tests and evaluation technology - Development of the ultrasonic method for two-phase mixture level measurement in nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    No, Hee Cheon; Kim, Sang Jae; Kim, Hyung Tae; Moon, Young Min [Korea Advanced Institute of Science and Technology, Taejon (Korea)

    2000-04-01

    An ultrasonic method is developed for the measurement of the two-phase mixture level in the reactor vessel or steam generator. The ultrasonic method is selected among the several non-nuclear two-phase mixture level measurement methods through two steps of selection procedure. A commercial ultrasonic level measurement method is modified for application into the high temperature, pressure, and other conditions. The calculation method of the ultrasonic velocity is modified to consider the medium as the homogeneous mixture of air and steam, and to be applied into the high temperature and pressure conditions. The cross-correlation technique is adopted as a detection method to reduced the effects of the attenuation and the diffused reflection caused by surface fluctuation. The waveguides are developed to reduce the loss of echo and to remove the effects of obstructs. The present experimental study shows that the developed ultrasonic method measures the two-phase mixture level more accurately than the conventional methods do. 21 refs., 60 figs., 13 tabs. (Author)

  13. Nuclear reactor alignment plate configuration

    Energy Technology Data Exchange (ETDEWEB)

    Altman, David A; Forsyth, David R; Smith, Richard E; Singleton, Norman R

    2014-01-28

    An alignment plate that is attached to a core barrel of a pressurized water reactor and fits within slots within a top plate of a lower core shroud and upper core plate to maintain lateral alignment of the reactor internals. The alignment plate is connected to the core barrel through two vertically-spaced dowel pins that extend from the outside surface of the core barrel through a reinforcement pad and into corresponding holes in the alignment plate. Additionally, threaded fasteners are inserted around the perimeter of the reinforcement pad and into the alignment plate to further secure the alignment plate to the core barrel. A fillet weld also is deposited around the perimeter of the reinforcement pad. To accomodate thermal growth between the alignment plate and the core barrel, a gap is left above, below and at both sides of one of the dowel pins in the alignment plate holes through with the dowel pins pass.

  14. Review of Transient Fuel Test Results at Sandia National Laboratories and the Potential for Future Fast Reactor Fuel Transient Testing in the Annular Core Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wright, Steven A.; Pickard, Paul S.; Parma, Edward J.; Vernon, Milton E.; Kelly, John; Tikare, Veena [Sandia National Laboratories, Org 6872 MS-1146, PO Box 5800 Albuquerque, New Mexico 87185 (United States)

    2009-06-15

    Reactor driven transient tests of fast reactor fuels may be required to support the development and certification of new fuels for Fast Reactors. The results of the transient fuel tests will likely be needed to support licensing and to provide validation data to support the safety case for a variety of proposed fast fuel types and reactors. In general reactor driven transient tests are used to identify basic phenomenology during reactor transients and to determine the fuel performance limits and margins to failure during design basis accidents such as loss of flow, loss of heat sink, and reactivity insertion accidents. This paper provides a summary description of the previous Sandia Fuel Disruption and Transient Axial Relocation tests that were performed in the Annular Core Research Reactor (ACRR) for the U.S. Nuclear Regulatory Commission almost 25 years ago. These tests consisted of a number of capsule tests and flowing gas tests that used fission heating to disrupt fresh and irradiated MOX fuel. The behavior of the fuel disruption, the generation of aerosols and the melting and relocation of fuel and cladding was recorded on high speed cinematography. This paper will present videos of the fuel disruption that was observed in these tests which reveal stark differences in fuel behavior between fresh and irradiated fuel. Even though these tests were performed over 25 years ago, their results are still relevant to today's reactor designs. These types of transient tests are again being considered by the Advanced Fuel Cycle Initiative to support the Global Nuclear Energy Partnership because of the need to perform tests on metal fuels and transuranic fuels. Because the Annular Core Research Reactor is the only transient test facility available within the US, a brief summary of Sandia's continued capability to perform these tests in the ACRR will also be provided. (authors)

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

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

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

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

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2003-01-01

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

  17. Development of an educational nuclear research reactor simulator

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-12-15

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

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

    Science.gov (United States)

    Radulescu, Laura; Pavelescu, Margarit

    2010-01-01

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

  19. Development of a research nuclear reactor simulator using LABVIEW®

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

  20. Jules Horowitz Reactor, a new irradiation facility: Improving dosimetry for the future of nuclear experimentation

    Energy Technology Data Exchange (ETDEWEB)

    Gregoire, G.; Beretz, D.; Destouches, C. [CEA, DEN, DER/SPEX, F-13108 Saint-Paul-lez-Durance (France)

    2011-07-01

    Document available in abstract form only, full text of document follows: The Jules Horowitz Reactor (JHR) is an experimental reactor under construction at the French Nuclear Energy and Alternative Energies Commission (CEA) facility at Cadarache. It will achieve its first criticality by the end of 2014. Experiments that will be conducted at JHR will deal with fuel, cladding, and material behavior. The JHR will also produce medical radio-isotopes and doped silicon for the electronic industry. As a new irradiation facility, its instrumentation will benefit from recent improvements. Nuclear instrumentation will include reactor dosimetry, as it is a reference technique to determine neutron fluence in experimental devices or characterize irradiation locations. Reactor dosimetry has been improved with the progress of simulation tools and nuclear data, but at the same time the customer needs have increased: Experimental results must have reduced and assessed uncertainties. This is now a necessary condition to perform an experimental irradiation in a test reactor. Items improved, in the framework of a general upgrading of the dosimetry process based on uncertainty minimization, will include dosimeter, nuclear data, and modelling scheme. (authors)

  1. Current Abstracts Nuclear Reactors and Technology

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-01-01

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

  2. Exploratory Design of a Reactor/Fuel Cycle Using Spent Nuclear Fuel Without Conventional Reprocessing - 13579

    Energy Technology Data Exchange (ETDEWEB)

    Bertch, Timothy C.; Schleicher, Robert W.; Rawls, John D. [General Atomics 3550 General Atomics Court San Diego, CA 92130 (United States)

    2013-07-01

    General Atomics has started design of a waste to energy nuclear reactor (EM2) that can use light water reactor (LWR) spent nuclear fuel (SNF). This effort addresses two problems: using an advanced small reactor with long core life to reduce nuclear energy overnight cost and providing a disposal path for LWR SNF. LWR SNF is re-fabricated into new EM2 fuel using a dry voloxidation process modeled on AIROX/ OREOX processes which remove some of the fission products but no heavy metals. By not removing all of the fission products the fuel remains self-protecting. By not separating heavy metals, the process remains proliferation resistant. Implementation of Energy Multiplier Module (EM2) fuel cycle will provide low cost nuclear energy while providing a long term LWR SNF disposition path which is important for LWR waste confidence. With LWR waste confidence recent impacts on reactor licensing, an alternate disposition path is highly relevant. Centered on a reactor operating at 250 MWe, the compact electricity generating system design maximizes site flexibility with truck transport of all system components and available dry cooling features that removes the need to be located near a body of water. A high temperature system using helium coolant, electricity is efficiently produced using an asynchronous high-speed gas turbine while the LWR SNF is converted to fission products. Reactor design features such as vented fuel and silicon carbide cladding support reactor operation for decades between refueling, with improved fuel utilization. Beyond the reactor, the fuel cycle is designed so that subsequent generations of EM2 reactor fuel will use the previous EM2 discharge, providing its own waste confidence plus eliminating the need for enrichment after the first generation. Additional LWR SNF is added at each re-fabrication to replace the removed fission products. The fuel cycle uses a dry voloxidation process for both the initial LWR SNF re-fabrication and later for EM2

  3. Piezoelectric material for use in a nuclear reactor core

    Science.gov (United States)

    Parks, D. A.; Reinhardt, Brian; Tittmann, B. R.

    2012-05-01

    In radiation environments ultrasonic nondestructive evaluation has great potential for improving reactor safety and furthering the understanding of radiation effects and materials. In both nuclear power plants and materials test reactors, elevated temperatures and high levels of radiation present challenges to ultrasonic NDE methodologies. The challenges are primarily due to the degradation of the ultrasonic sensors utilized. We present results from the operation of a ultrasonic piezoelectric transducer, composed of bulk single crystal AlN, in a nuclear reactor core for over 120 MWHrs. The transducer was coupled to an aluminum cylinder and operated in pulse echo mode throughout the irradiation. In addition to the pulse echo testing impedance data were obtained. Further, the piezoelectric coefficient d33 was measured prior to irradiation and found to be 5.5 pC/N which is unchanged from as-grown samples, and in fact higher than the measured d33 for many as-grown samples.

  4. The MAUS nuclear space reactor with ion propulsion system

    Energy Technology Data Exchange (ETDEWEB)

    Mainardi, Enrico [DINCE - Dipartimento di Ingegneria Nucleare e Conversioni Energetiche, University of Rome ' La Sapienza' , C.so V. Emanuele II, 244, 00186 Rome (Italy)]. E-mail: mainardi@frascati.enea.it

    2006-06-01

    MAUS (Moltiplicatore Avanzato Ultracompatto Spaziale) is a nuclear reactor concept design capable to ensure a reliable, long-lasting, low-mass, compact energy supply needed for advanced, future space missions. The exploration of the solar system and the space beyond requires the development of nuclear energy generators for supplying electricity to space-bases, spacecrafts, probes or satellites, as well as for propelling ships in long space missions. For propulsion, the MAUS nuclear reactor could be used to power electric ion drive engines. An ion engine is able to build up to very high velocities, far greater than chemical propulsion systems, but has high power and long service requirements. The MAUS concept is described, together with the ion propulsion engine and together with the reference thermoionic process used to convert the thermal power into electricity. The design work has been performed at the Nuclear Engineering and Energy Conversion Department of the University of Rome 'La Sapienza' starting from 1992 on an issue submitted by the Italian Space Agency (ASI), in cooperation with the research laboratories of ENEA.

  5. The Maus nuclear space reactor with ion propulsion system

    Energy Technology Data Exchange (ETDEWEB)

    Enrico Mainardi [DINCE - Dipartimento di Ingegneria Nucleare e Conversioni Energetiche, University of Rome ' La Sapienza' , C.so V. EmanueleII, 244, 00186 Roma (Italy)

    2006-07-01

    MAUS (Moltiplicatore Avanzato Ultracompatto Spaziale) is a nuclear reactor concept design capable to ensure a reliable, long lasting, low mass, compact energy supply needed for advanced, future space missions. The exploration of the solar system and the space beyond requires the development of nuclear energy generators for supplying electricity to space-bases, spacecrafts, probes or satellites, as well as for propelling ships in long space missions. For propulsion, the MAUS nuclear reactor could be used to power electric ion drive engines. An ion engine is able to build up to very high velocities, far greater than chemical propulsion systems, but has high power and long service requirements. The MAUS concept is described, together with the ion propulsion engine and together with the reference thermionic process used to convert the thermal power into electricity. The design work has been performed at the Nuclear Engineering and Energy Conversion Department of the University of Rome 'La Sapienza' starting from 1992 on an issue submitted by the Italian Space Agency (ASI), in cooperation with the research laboratories of ENEA. (author)

  6. Instrumentation to Enhance Advanced Test Reactor Irradiations

    Energy Technology Data Exchange (ETDEWEB)

    J. L. Rempe; D. L. Knudson; K. G. Condie; J. E. Daw; S. C. Taylor

    2009-09-01

    The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support U.S. leadership in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR will support basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort is to prove new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors has been completed. Based on this review, recommendations are made with respect to what instrumentation is needed at the ATR and a strategy has been developed for obtaining these sensors. Progress toward implementing this strategy is reported in this document. It is anticipated that this report will be updated on an annual basis.

  7. FUEL COMPOSITION FOR NUCLEAR REACTORS

    Science.gov (United States)

    Andersen, J.C.

    1963-08-01

    A process for making refractory nuclear fuel elements involves heating uranium and silicon powders in an inert atmosphere to 1600 to 1800 deg C to form USi/sub 3/; adding silicon carbide, carbon, 15% by weight of nickel and aluminum, and possibly also molybdenum and silicon powders; shaping the mixture; and heating to 1700 to 2050 deg C again in an inert atmosphere. Information on obtaining specific compositions is included. (AEC)

  8. Passive heat-transfer means for nuclear reactors. [LMFBR

    Science.gov (United States)

    Burelbach, J.P.

    1982-06-10

    An improved passive cooling arrangement is disclosed for maintaining adjacent or related components of a nuclear reactor within specified temperature differences. Specifically, heat pipes are operatively interposed between the components, with the vaporizing section of the heat pipe proximate the hot component operable to cool it and the primary condensing section of the heat pipe proximate the other and cooler component operable to heat it. Each heat pipe further has a secondary condensing section that is located outwardly beyond the reactor confinement and in a secondary heat sink, such as air ambient the containment, that is cooler than the other reactor component. By having many such heat pipes, an emergency passive cooling system is defined that is operative without electrical power.

  9. Fuel-Cycle and Nuclear Material Disposition Issues Associated with High-Temperature Gas Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Shropshire, D.E.; Herring, J.S.

    2004-10-03

    The objective of this paper is to facilitate a better understanding of the fuel-cycle and nuclear material disposition issues associated with high-temperature gas reactors (HTGRs). This paper reviews the nuclear fuel cycles supporting early and present day gas reactors, and identifies challenges for the advanced fuel cycles and waste management systems supporting the next generation of HTGRs, including the Very High Temperature Reactor, which is under development in the Generation IV Program. The earliest gas-cooled reactors were the carbon dioxide (CO2)-cooled reactors. Historical experience is available from over 1,000 reactor-years of operation from 52 electricity-generating, CO2-cooled reactor plants that were placed in operation worldwide. Following the CO2 reactor development, seven HTGR plants were built and operated. The HTGR came about from the combination of helium coolant and graphite moderator. Helium was used instead of air or CO2 as the coolant. The helium gas has a significant technical base due to the experience gained in the United States from the 40-MWe Peach Bottom and 330-MWe Fort St. Vrain reactors designed by General Atomics. Germany also built and operated the 15-MWe Arbeitsgemeinschaft Versuchsreaktor (AVR) and the 300-MWe Thorium High-Temperature Reactor (THTR) power plants. The AVR, THTR, Peach Bottom and Fort St. Vrain all used fuel containing thorium in various forms (i.e., carbides, oxides, thorium particles) and mixtures with highly enriched uranium. The operational experience gained from these early gas reactors can be applied to the next generation of nuclear power systems. HTGR systems are being developed in South Africa, China, Japan, the United States, and Russia. Elements of the HTGR system evaluated included fuel demands on uranium ore mining and milling, conversion, enrichment services, and fuel fabrication; fuel management in-core; spent fuel characteristics affecting fuel recycling and refabrication, fuel handling, interim

  10. Medical Radioisotopes Production Without A Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Van der Keur, H.

    2010-05-15

    This report is answering the key question: Is it possible to ban the use of research reactors for the production of medical radioisotopes? Chapter 2 offers a summarized overview on the history of nuclear medicine. Chapter 3 gives an overview of the basic principles and understandings of nuclear medicine. The production of radioisotopes and its use in radiopharmaceuticals as a tracer for imaging particular parts of the inside of the human body (diagnosis) or as an agent in radiotherapy. Chapter 4 lists the use of popular medical radioisotopes used in nuclear imaging techniques and radiotherapy. Chapter 5 analyses reactor-based radioisotopes that can be produced by particle accelerators on commercial scale, other alternatives and the advantages of the cyclotron. Chapter 6 gives an overview of recent developments and prospects in worldwide radioisotopes production. Chapter 7 presents discussion, conclusions and recommendations, and is answering the abovementioned key question of this report: Is it possible to ban the use of a nuclear reactor for the production of radiopharmaceuticals? Is a safe and secure production of radioisotopes possible?.

  11. Risk perception & strategic decision making :general insights, a framework, and specific application to electricity generation using nuclear energy.

    Energy Technology Data Exchange (ETDEWEB)

    Brewer, Jeffrey D.

    2005-11-01

    The objective of this report is to promote increased understanding of decision making processes and hopefully to enable improved decision making regarding high-consequence, highly sophisticated technological systems. This report brings together insights regarding risk perception and decision making across domains ranging from nuclear power technology safety, cognitive psychology, economics, science education, public policy, and neural science (to name a few). It forms them into a unique, coherent, concise framework, and list of strategies to aid in decision making. It is suggested that all decision makers, whether ordinary citizens, academics, or political leaders, ought to cultivate their abilities to separate the wheat from the chaff in these types of decision making instances. The wheat includes proper data sources and helpful human decision making heuristics; these should be sought. The chaff includes ''unhelpful biases'' that hinder proper interpretation of available data and lead people unwittingly toward inappropriate decision making ''strategies''; obviously, these should be avoided. It is further proposed that successfully accomplishing the wheat vs. chaff separation is very difficult, yet tenable. This report hopes to expose and facilitate navigation away from decision-making traps which often ensnare the unwary. Furthermore, it is emphasized that one's personal decision making biases can be examined, and tools can be provided allowing better means to generate, evaluate, and select among decision options. Many examples in this report are tailored to the energy domain (esp. nuclear power for electricity generation). The decision making framework and approach presented here are applicable to any high-consequence, highly sophisticated technological system.

  12. Advanced Space Nuclear Reactors from Fiction to Reality

    Science.gov (United States)

    Popa-Simil, L.

    The advanced nuclear power sources are used in a large variety of science fiction movies and novels, but their practical development is, still, in its early conceptual stages, some of the ideas being confirmed by collateral experiments. The novel reactor concept uses the direct conversion of nuclear energy into electricity, has electronic control of reactivity, being surrounded by a transmutation blanket and very thin shielding being small and light that at its very limit may be suitable to power an autonomously flying car. It also provides an improved fuel cycle producing minimal negative impact to environment. The key elements started to lose the fiction attributes, becoming viable actual concepts and goals for the developments to come, and on the possibility to achieve these objectives started to become more real because the theory shows that using the novel nano-technologies this novel reactor might be achievable in less than a century.

  13. Wire core reactor for nuclear thermal propulsion

    Science.gov (United States)

    Harty, Richard B.; Brengle, Robert G.

    1993-01-01

    Studies have been performed of a compact high-performance nuclear rocket reactor that incorporates a tungsten alloy wire fuel element. This reactor, termed the wire core reactor, can deliver a specific impulse of 1,000 s using an expander cycle and a nozzle expansion ratio of 500 to 1. The core is constructed of layers of 0.8-mm-dia fueled tungsten wires wound over alternate layers of spacer wires, which forms a rugged annular lattice. Hydrogen flow in the core is annular, flowing from inside to outside. In addition to the concepts compact size and good heat transfer, the core has excellent power-flow matching features and can resist vibration and thermal stresses during star-up and shutdown.

  14. Source term and radiation dose estimates for postulated damage to the 102 Building at the General Electric Vallecitos Nuclear Center

    Energy Technology Data Exchange (ETDEWEB)

    Mishima, J.; McPherson, R.B.; Schwendiman, L.C.; Watson, E.C.; Ayer, J.E.

    1979-02-01

    Three scenarios representing significant levels of containment loss due to moderate, substantial, and major damage to the 102 Building at the Vallecitos Nuclear Center are postulated, and the potential radiation doses to the general population as a result of the airborne releases of radionuclides are estimated. The damage scenarios are not correlated to any specific level of seismic activity. The three scenarios are: (1) Moderate damage scenario--perforation of the enclosures in and the structure comprising the Plutonium Analytical Laboratory. (2) Substantial damage scenario--complete loss of containment of the Plutonium Analytical Laboratory and loss of the filters sealing the inlet to the Radioactive Materials Laboratory hot cells. (3) Major damage scenario--the damage outlined in (2) plus the perforation of enclosures holding significant inventories of dispersible plutonium in and the structure comprising the Advanced Fuels Laboratory.

  15. An overview of future sustainable nuclear power reactors

    OpenAIRE

    Andreas Poullikkas

    2013-01-01

    In this paper an overview of the current and future nuclear power reactor technologies is carried out. In particular, the nuclear technology is described and the classification of the current and future nuclear reactors according to their generation is provided. The analysis has shown that generation II reactors currently in operation all around the world lack significantly in safety precautions and are prone to loss of coolant accident (LOCA). In contrast, generation III reactors, which are ...

  16. Muon trackers for imaging a nuclear reactor

    Science.gov (United States)

    Kume, N.; Miyadera, H.; Morris, C. L.; Bacon, J.; Borozdin, K. N.; Durham, J. M.; Fuzita, K.; Guardincerri, E.; Izumi, M.; Nakayama, K.; Saltus, M.; Sugita, T.; Takakura, K.; Yoshioka, K.

    2016-09-01

    A detector system for assessing damage to the cores of the Fukushima Daiichi nuclear reactors by using cosmic-ray muon tomography was developed. The system consists of a pair of drift-tube tracking detectors of 7.2× 7.2-m2 area. Each muon tracker consists of 6 x-layer and 6 y-layer drift-tube detectors. Each tracker is capable of measuring muon tracks with 12 mrad angular resolutions, and is capable of operating under 50-μ Sv/h radiation environment by removing gamma induced background with a novel time-coincidence logic. An estimated resolution to observe nuclear fuel debris at Fukushima Daiichi is 0.3 m when the core is imaged from outside the reactor building.

  17. Nuclear modules for space electric propulsion

    Science.gov (United States)

    Difilippo, F. C.

    1998-01-01

    The analysis of interplanetary cargo and piloted missions requires the calculations of the performances and masses of subsystems to be integrated in a final design. In a preliminary and scoping stage the designer needs to evaluate options in an iterative way by using simulations that run fast on a computer. As a consequence of a collaborative agreement between the National Aeronautic and Space Administration (NASA) and the Oak Ridge National Laboratory (ORNL), ORNL has been involved in the development of models and calculational procedures for the analysis (neutronic and thermal hydraulic) of power sources for nuclear electric propulsion. The nuclear modules will be integrated into the whole simulation of the nuclear electric propulsion system. The vehicles use either a Brayton direct-conversion cycle, using the heated helium from a NERVA-type reactor, or a potassium Rankine cycle, with the working fluid heated on the secondary side of a heat exchanger and lithium on the primary side coming from a fast reactor. Given a set of input conditions, the codes calculate composition, dimensions, volumes, and masses of the core, reflector, control system, pressure vessel, neutron and gamma shields, as well as the thermal hydraulic conditions of the coolant, clad and fuel. Input conditions are power, core life, pressure and temperature of the coolant at the inlet of the core, either the temperature of the coolant at the outlet of the core or the coolant mass flow and the fluences and integrated doses at the cargo area. Using state-of-the-art neutron cross sections and transport codes, a database was created for the neutronic performance of both reactor designs. The free parameters of the models are the moderator/fuel mass ratio for the NERVA reactor and the enrichment and the pitch of the lattice for the fast reactor. Reactivity and energy balance equations are simultaneously solved to find the reactor design. Thermalhydraulic conditions are calculated by solving the one

  18. Reference Neutron Radiographs of Nuclear Reactor Fuel

    DEFF Research Database (Denmark)

    Domanus, Joseph Czeslaw

    1986-01-01

    Reference neutron radiographs of nuclear reactor fuel were produced by the Euraton Neutron Radiography Working Group and published in 1984 by the Reidel Publishing Company. In this collection a classification is given of the various neutron radiographic findings, that can occur in different parts...... of pelletized, annular and vibro-conpacted nuclear fuel pins. Those parts of the pins are shown where changes of appearance differ from those for the parts as fabricated. Also radiographs of those as fabricated parts are included. The collection contains 158 neutron radiographs, reproduced on photographic paper...

  19. Neutron spectrometer for fast nuclear reactors

    CERN Document Server

    Osipenko, M; Ricco, G; Caiffi, B; Pompili, F; Pillon, M; Angelone, M; Verona-Rinati, G; Cardarelli, R; Mila, G; Argiro, S

    2015-01-01

    In this paper we describe the development and first tests of a neutron spectrometer designed for high flux environments, such as the ones found in fast nuclear reactors. The spectrometer is based on the conversion of neutrons impinging on $^6$Li into $\\alpha$ and $t$ whose total energy comprises the initial neutron energy and the reaction $Q$-value. The $^6$LiF layer is sandwiched between two CVD diamond detectors, which measure the two reaction products in coincidence. The spectrometer was calibrated at two neutron energies in well known thermal and 3 MeV neutron fluxes. The measured neutron detection efficiency varies from 4.2$\\times 10^{-4}$ to 3.5$\\times 10^{-8}$ for thermal and 3 MeV neutrons, respectively. These values are in agreement with Geant4 simulations and close to simple estimates based on the knowledge of the $^6$Li(n,$\\alpha$)$t$ cross section. The energy resolution of the spectrometer was found to be better than 100 keV when using 5 m cables between the detector and the preamplifiers.

  20. Some views on nuclear reactor safety

    Energy Technology Data Exchange (ETDEWEB)

    Tanguy, P.Y. [Electricite de France, Paris (France)

    1995-04-01

    This document is the text of a speech given by Pierre Y. Tanguy (Electricite de France) at the 22nd Water Reactor Safety Meeting held in Bethesda, MD in 1994. He describes the EDF nuclear program in broad terms and proceeds to discuss operational safety results with EDF plants. The speaker also outlines actions to enhance safety planned for the future, and he briefly mentions French cooperation with the Chinese on the Daya Bay project.

  1. Oklo reactors and implications for nuclear science

    Science.gov (United States)

    Davis, E. D.; Gould, C. R.; Sharapov, E. I.

    2014-04-01

    We summarize the nuclear physics interests in the Oklo natural nuclear reactors, focusing particularly on developments over the past two decades. Modeling of the reactors has become increasingly sophisticated, employing Monte Carlo simulations with realistic geometries and materials that can generate both the thermal and epithermal fractions. The water content and the temperatures of the reactors have been uncertain parameters. We discuss recent work pointing to lower temperatures than earlier assumed. Nuclear cross-sections are input to all Oklo modeling and we discuss a parameter, the 175Lu ground state cross-section for thermal neutron capture leading to the isomer 176mLu, that warrants further investigation. Studies of the time dependence of dimensionless fundamental constants have been a driver for much of the recent work on Oklo. We critically review neutron resonance energy shifts and their dependence on the fine structure constant α and the ratio Xq = mq/Λ (where mq is the average of the u and d current quark masses and Λ is the mass scale of quantum chromodynamics (QCD)). We suggest a formula for the combined sensitivity to α and Xq that exhibits the dependence on proton number Z and mass number A, potentially allowing quantum electrodynamic (QED) and QCD effects to be disentangled if a broader range of isotopic abundance data becomes available.

  2. Nuclear vapor thermal reactor propulsion technology

    Science.gov (United States)

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

    1993-01-01

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

  3. Advanced burner test reactor preconceptual design report.

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y. I.; Finck, P. J.; Grandy, C.; Cahalan, J.; Deitrich, L.; Dunn, F.; Fallin, D.; Farmer, M.; Fanning, T.; Kim, T.; Krajtl, L.; Lomperski, S.; Moisseytsev, A.; Momozaki, Y.; Sienicki, J.; Park, Y.; Tang, Y.; Reed, C.; Tzanos, C; Wiedmeyer, S.; Yang, W.; Chikazawa, Y.; JAEA

    2008-12-16

    The goals of the Global Nuclear Energy Partnership (GNEP) are to expand the use of nuclear energy to meet increasing global energy demand, to address nuclear waste management concerns and to promote non-proliferation. Implementation of the GNEP requires development and demonstration of three major technologies: (1) Light water reactor (LWR) spent fuel separations technologies that will recover transuranics to be recycled for fuel but not separate plutonium from other transuranics, thereby providing proliferation-resistance; (2) Advanced Burner Reactors (ABRs) based on a fast spectrum that transmute the recycled transuranics to produce energy while also reducing the long term radiotoxicity and decay heat loading in the repository; and (3) Fast reactor fuel recycling technologies to recover and refabricate the transuranics for repeated recycling in the fast reactor system. The primary mission of the ABR Program is to demonstrate the transmutation of transuranics recovered from the LWR spent fuel, and hence the benefits of the fuel cycle closure to nuclear waste management. The transmutation, or burning of the transuranics is accomplished by fissioning and this is most effectively done in a fast spectrum. In the thermal spectrum of commercial LWRs, some transuranics capture neutrons and become even heavier transuranics rather than being fissioned. Even with repeated recycling, only about 30% can be transmuted, which is an intrinsic limitation of all thermal spectrum reactors. Only in a fast spectrum can all transuranics be effectively fissioned to eliminate their long-term radiotoxicity and decay heat. The Advanced Burner Test Reactor (ABTR) is the first step in demonstrating the transmutation technologies. It directly supports development of a prototype full-scale Advanced Burner Reactor, which would be followed by commercial deployment of ABRs. The primary objectives of the ABTR are: (1) To demonstrate reactor-based transmutation of transuranics as part of an

  4. High-intensity power-resolved radiation imaging of an operational nuclear reactor

    Science.gov (United States)

    Beaumont, Jonathan S.; Mellor, Matthew P.; Villa, Mario; Joyce, Malcolm J.

    2015-01-01

    Knowledge of the neutron distribution in a nuclear reactor is necessary to ensure the safe and efficient burnup of reactor fuel. Currently these measurements are performed by in-core systems in what are extremely hostile environments and in most reactor accident scenarios it is likely that these systems would be damaged. Here we present a compact and portable radiation imaging system with the ability to image high-intensity fast-neutron and gamma-ray fields simultaneously. This system has been deployed to image radiation fields emitted during the operation of a TRIGA test reactor allowing a spatial visualization of the internal reactor conditions to be obtained. The imaged flux in each case is found to scale linearly with reactor power indicating that this method may be used for power-resolved reactor monitoring and for the assay of ongoing nuclear criticalities in damaged nuclear reactors. PMID:26450669

  5. High-intensity power-resolved radiation imaging of an operational nuclear reactor

    Science.gov (United States)

    Beaumont, Jonathan S.; Mellor, Matthew P.; Villa, Mario; Joyce, Malcolm J.

    2015-10-01

    Knowledge of the neutron distribution in a nuclear reactor is necessary to ensure the safe and efficient burnup of reactor fuel. Currently these measurements are performed by in-core systems in what are extremely hostile environments and in most reactor accident scenarios it is likely that these systems would be damaged. Here we present a compact and portable radiation imaging system with the ability to image high-intensity fast-neutron and gamma-ray fields simultaneously. This system has been deployed to image radiation fields emitted during the operation of a TRIGA test reactor allowing a spatial visualization of the internal reactor conditions to be obtained. The imaged flux in each case is found to scale linearly with reactor power indicating that this method may be used for power-resolved reactor monitoring and for the assay of ongoing nuclear criticalities in damaged nuclear reactors.

  6. Expert system for online surveillance of nuclear reactor coolant pumps

    Energy Technology Data Exchange (ETDEWEB)

    Gross, K.C.; Singer, R.M.; Humenik, K.E.

    1992-12-31

    This report describes an expert system for online surveillance of nuclear reactor coolant pumps. This system provides a means for early detection of pump or sensor degradation. Degradation is determined through the use of a statistical analysis technique, sequential probability ratio test, applied to information from several sensors which are responsive to differing physical parameters. The results of sequential testing of the data provide the operator with an early warning of possible sensor or pump failure.

  7. Reactors

    CERN Document Server

    International Electrotechnical Commission. Geneva

    1988-01-01

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

  8. Immobilization of ion exchange radioactive resins of the TRIGA Mark III nuclear reactor; Inmovilizacion de resinas de intercambio ionico radiactivas del reactor nuclear Triga Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Garcia M, H.; Emeterio H, M.; Canizal S, C. [Instituto Nacional de Investigaciones Nucleares, A.P. 18-1027, C.P. 11801 Mexico D.F. (Mexico)

    2000-07-01

    This work has the objective to develop the process and to define the agglutinating material which allows the immobilization of the ion exchange radioactive resins coming from the TRIGA Mark III nuclear reactor contaminated with Ba-133, Co-60, Cs-137, Eu-152, and Mn-54 through the behavior analysis of different immobilization agents such as: bitumens, cement and polyester resin. According to the International Standardization the archetype samples were observed with the following tests: determination of free liquid, leaching, charge resistance, biodegradation, irradiation, thermal cycle, burned resistance. Generally all the tests were satisfactorily achieved, for each agent. Therefore, the polyester resin could be considered as the main immobilizing. (Author)

  9. Nuclear electric power sources

    Science.gov (United States)

    Singh, J. J.

    1978-01-01

    Measurements on radioactive commercial p-n junction silicon cells show that these units are capable of delivering several hundred microwatts per curie of Am-241 alpha source, indicating their usefulness in such electronic devices as hearing aids, heart pacemakers, electronic watches, delay timers and nuclear dosimeter chargers. It is concluded that the Am-241 sources are superior to the beta sources used previously, because of higher alpha specific ionization and simultaneous production of low energy photons which are easily converted into photoelectrons for additional power.

  10. A General Test of Association for Quantitative Traits in Nuclear Families

    National Research Council Canada - National Science Library

    Abecasis, G.R; Cardon, L.R; Cookson, W.O.C

    2000-01-01

    .... Appropriately designed tests of association that use family-based controls to account for population substructure can provide direct tests of linkage disequilibrium and efficient fine-mapping tools. T...

  11. Nuclear Electric looks to the private sector

    Energy Technology Data Exchange (ETDEWEB)

    Varley, James

    1995-03-01

    The state-owned utility Nuclear Electric, which is responsible for nuclear power generation in England and Wales, was created in 1990 following withdrawal of nuclear from electricity privatisation. Having successfully made itself much more commercial, Nuclear Electric would like the freedom of operating in the private sector. (author).

  12. Uncertainties in the Anti-neutrino Production at Nuclear Reactors

    OpenAIRE

    Djurcic, Z.(Argonne National Laboratory, Argonne, Illinois, 60439, U.S.A.); Detwiler, J. A.; Piepke, A.; Foster Jr., V. R.; Miller, L.; Gratta, G.

    2008-01-01

    Anti-neutrino emission rates from nuclear reactors are determined from thermal power measurements and fission rate calculations. The uncertainties in these quantities for commercial power plants and their impact on the calculated interaction rates in electron anti-neutrino detectors is examined. We discuss reactor-to-reactor correlations between the leading uncertainties and their relevance to reactor anti-neutrino experiments.

  13. iDREAM: an industrial detector for nuclear reactor monitoring

    Science.gov (United States)

    Gribov, I. V.; Gromov, M. B.; Lukjanchenko, G. A.; Novikova, G. J.; Obinyakov, B. A.; Oralbaev, A. Y.; Skorokhvatov, M. D.; Sukhotin, S. V.; Chepurnov, A. S.; Etenko, A. V.

    2016-02-01

    Prototype of industrial reactor antineutrino detector iDREAM is dedicated for an experiment to demonstrate the possibility of remote monitoring of PWR reactor operational modes by neutrino method in real-time in order to avoid undeclared exposure modes for nuclear fuel and unauthorized removal of isotopes. The prototype detector was started up in 2014. To test the detector elements and components of electronics distilled water has been used as a target, which enables the use of Cerenkov radiation from cosmic muons as a physical signal. Also parallel measuring of the long-term stability has been doing for samples of liquid organic scintillator doped with gadolinium and synthesized by different methods

  14. PITR: Princeton Ignition Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    1978-12-01

    The principal objectives of the PITR - Princeton Ignition Test Reactor - are to demonstrate the attainment of thermonuclear ignition in deuterium-tritium, and to develop optimal start-up techniques for plasma heating and current induction, in order to determine the most favorable means of reducing the size and cost of tokamak power reactors. This report describes the status of the plasma and engineering design features of the PITR. The PITR geometry is chosen to provide the highest MHD-stable values of beta in a D-shaped plasma, as well as ease of access for remote handling and neutral-beam injection.

  15. Nuclear electric ion propulsion for three deep space missions

    Science.gov (United States)

    Chiravalle, Vincent P.

    2008-03-01

    Nuclear electric ion propulsion is considered for three sample deep space missions starting from a 500 km low Earth orbit encompassing the transfer of a 100 MT payload into a 1500 km orbit around Mars, the rendezvous of a 10 MT payload with the Jovian moon Europa and the rendezvous of a similar payload with Saturn's moon Titan. Near term ion engine and space nuclear reactor technology are assumed. It is shown that nuclear electric ion propulsion offers more than twice the payload for the Mars mission relative to the case when a nuclear thermal rocket is used for the trans-Mars injection maneuver at Earth, and about the same payload advantage relative to the case when solar electric propulsion is used for the Mars heliocentric transfer. For missions to the outer planets nuclear electric ion propulsion increases the payload mass fraction by a factor of two or more compared with high thrust systems that utilize gravity assist trajectories.

  16. Application of Simulated Reactivity Feedback in Nonnuclear Testing of a Direct-Drive Gas-Cooled Reactor

    Science.gov (United States)

    Bragg-Sitton, S. M.; Webster, K. L.

    2007-01-01

    Nonnuclear testing can be a valuable tool in the development of an in-space nuclear power or propulsion system. In a nonnuclear test facility, electric heaters are used to simulate heat from nuclear fuel. Standard testing allows one to fully assess thermal, heat transfer, and stress related attributes of a given system but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. The integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and full nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response and response characteristics, and assess potential design improvements with a relatively small fiscal investment. Initial system dynamic response testing was demonstrated on the integrated SAFE 100a heat pipe cooled, electrically heated reactor and heat exchanger hardware. This Technical Memorandum discusses the status of the planned dynamic test methodology for implementation in the direct-drive gas-cooled reactor testing and assesses the additional instrumentation needed to implement high-fidelity dynamic testing.

  17. Designed porosity materials in nuclear reactor components

    Science.gov (United States)

    Yacout, A. M.; Pellin, Michael J.; Stan, Marius

    2016-09-06

    A nuclear fuel pellet with a porous substrate, such as a carbon or tungsten aerogel, on which at least one layer of a fuel containing material is deposited via atomic layer deposition, and wherein the layer deposition is controlled to prevent agglomeration of defects. Further, a method of fabricating a nuclear fuel pellet, wherein the method features the steps of selecting a porous substrate, depositing at least one layer of a fuel containing material, and terminating the deposition when the desired porosity is achieved. Also provided is a nuclear reactor fuel cladding made of a porous substrate, such as silicon carbide aerogel or silicon carbide cloth, upon which layers of silicon carbide are deposited.

  18. Designed porosity materials in nuclear reactor components

    Energy Technology Data Exchange (ETDEWEB)

    Yacout, A. M.; Pellin, Michael J.; Stan, Marius

    2016-09-06

    A nuclear fuel pellet with a porous substrate, such as a carbon or tungsten aerogel, on which at least one layer of a fuel containing material is deposited via atomic layer deposition, and wherein the layer deposition is controlled to prevent agglomeration of defects. Further, a method of fabricating a nuclear fuel pellet, wherein the method features the steps of selecting a porous substrate, depositing at least one layer of a fuel containing material, and terminating the deposition when the desired porosity is achieved. Also provided is a nuclear reactor fuel cladding made of a porous substrate, such as silicon carbide aerogel or silicon carbide cloth, upon which layers of silicon carbide are deposited.

  19. Advanced nuclear reactor public opinion project

    Energy Technology Data Exchange (ETDEWEB)

    Benson, B.

    1991-07-25

    This Interim Report summarizes the findings of our first twenty in-depth interviews in the Advanced Nuclear Reactor Public Opinion Project. We interviewed 6 industry trade association officials, 3 industry attorneys, 6 environmentalists/nuclear critics, 3 state officials, and 3 independent analysts. In addition, we have had numerous shorter discussions with various individuals concerned about nuclear power. The report is organized into the four categories proposed at our April, 1991, Advisory Group meeting: safety, cost-benefit analysis, science education, and communications. Within each category, some change of focus from that of the Advisory Group has been required, to reflect the findings of our interviews. This report limits itself to describing our findings. An accompanying memo draws some tentative conclusions.

  20. Thermophotovoltaic Energy Conversion in Space Nuclear Reactor Power Systems

    Science.gov (United States)

    2004-12-01

    contrasted with nuclear thermal rockets which use the heat from a nuclear fission reactor to heat propellant to provide rocket thrust and radioisotope...K. Note that the highest temperature (2550 K by the Pewee reactor) was for a nuclear thermal rocket application and has the shortest duration (40 min

  1. Recent irradiation tests for future nuclear system at HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Man Soon; Choo, Kee Nam; Yang, Seong Woo; Park, Sang Jun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-05-15

    The capsule at HANARO is a device that evaluates the irradiation effects of nuclear materials and fuels, which can reproduce the environment of nuclear power plants and accelerate to reach to the end of life condition. As the integrity assessment and the extension of lifetime of nuclear power plants are recently considered as important issues in Korea, the requirements for irradiation test are gradually being increased. The capacity and capability irradiation tests at HANARO are becoming important because Korea strives to develop SFR (Sodium-cooled Fast Reactor) and VHTR (Very High Temperature Reactor) among the future nuclear system and to export the research reactors and to develop the fusion reactor technology.

  2. Light Water Reactor Sustainability Program: Evaluation of Localized Cable Test Methods for Nuclear Power Plant Cable Aging Management Programs

    Energy Technology Data Exchange (ETDEWEB)

    Glass, Samuel W. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Fifield, Leonard S. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Hartman, Trenton S. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2016-05-30

    This Pacific Northwest National Laboratory (PNNL) milestone report describes progress to date on the investigation of nondestructive test (NDE) methods focusing particularly on local measurements that provide key indicators of cable aging and damage. The work includes a review of relevant literature as well as hands-on experimental verification of inspection capabilities. As NPPs consider applying for second, or subsequent, license renewal (SLR) to extend their operating period from 60 years to 80 years, it important to understand how the materials installed in plant systems and components will age during that time and develop aging management programs (AMPs) to assure continued safe operation under normal and design basis events (DBE). Normal component and system tests typically confirm the cables can perform their normal operational function. The focus of the cable test program is directed toward the more demanding challenge of assuring the cable function under accident or DBE. Most utilities already have a program associated with their first life extension from 40 to 60 years. Regrettably, there is neither a clear guideline nor a single NDE that can assure cable function and integrity for all cables. Thankfully, however, practical implementation of a broad range of tests allows utilities to develop a practical program that assures cable function to a high degree. The industry has adopted 50% elongation at break (EAB) relative to the un-aged cable condition as the acceptability standard. All tests are benchmarked against the cable EAB test. EAB is a destructive test so the test programs must apply an array of other NDE tests to assure or infer the overall set of cable’s system integrity. These cable NDE programs vary in rigor and methodology. As the industry gains experience with the efficacy of these programs, it is expected that implementation practice will converge to a more common approach. This report addresses the range of local NDE cable tests that are

  3. EXTENDING NUCLEAR ENERGY TO NON-ELECTRICAL APPLICATIONS

    Energy Technology Data Exchange (ETDEWEB)

    R. Boardman; M. McKellar; D. Ingersoll; Z. Houghton; , R. Bromm; C. Desportes

    2014-09-01

    Electricity represents less than half of all energy consumed in the United States and globally. Although a few commercial nuclear power plants world-wide provide energy to non-electrical applications such as district heating and water desalination, nuclear energy has been largely relegated to base-load electricity production. A new generation of smaller-sized nuclear power plants offers significant promise for extending nuclear energy to many non-electrical applications. The NuScale small modular reactor design is especially well suited for these non-traditional customers due to its small unit size, very robust reactor protection features and a highly flexible and scalable plant design. A series of technical and economic evaluation studies have been conducted to assess the practicality of using a NuScale plant to provide electricity and heat to a variety of non-electrical applications, including water desalination, oil refining, and hydrogen production. The studies serve to highlight the unique design features of the NuScale plant for these applications and provide encouraging conclusions regarding the technical and economic viability of extending clean nuclear energy to a broad range of non-electrical energy consumers.

  4. Preliminary materials selection issues for the next generation nuclear plant reactor pressure vessel.

    Energy Technology Data Exchange (ETDEWEB)

    Natesan, K.; Majumdar, S.; Shankar, P. S.; Shah, V. N.; Nuclear Engineering Division

    2007-03-21

    In the coming decades, the United States and the entire world will need energy supplies to meet the growing demands due to population increase and increase in consumption due to global industrialization. One of the reactor system concepts, the Very High Temperature Reactor (VHTR), with helium as the coolant, has been identified as uniquely suited for producing hydrogen without consumption of fossil fuels or the emission of greenhouse gases [Generation IV 2002]. The U.S. Department of Energy (DOE) has selected this system for the Next Generation Nuclear Plant (NGNP) Project, to demonstrate emissions-free nuclear-assisted electricity and hydrogen production within the next 15 years. The NGNP reference concepts are helium-cooled, graphite-moderated, thermal neutron spectrum reactors with a design goal outlet helium temperature of {approx}1000 C [MacDonald et al. 2004]. The reactor core could be either a prismatic graphite block type core or a pebble bed core. The use of molten salt coolant, especially for the transfer of heat to hydrogen production, is also being considered. The NGNP is expected to produce both electricity and hydrogen. The process heat for hydrogen production will be transferred to the hydrogen plant through an intermediate heat exchanger (IHX). The basic technology for the NGNP has been established in the former high temperature gas reactor (HTGR) and demonstration plants (DRAGON, Peach Bottom, AVR, Fort St. Vrain, and THTR). In addition, the technologies for the NGNP are being advanced in the Gas Turbine-Modular Helium Reactor (GT-MHR) project, and the South African state utility ESKOM-sponsored project to develop the Pebble Bed Modular Reactor (PBMR). Furthermore, the Japanese HTTR and Chinese HTR-10 test reactors are demonstrating the feasibility of some of the planned components and materials. The proposed high operating temperatures in the VHTR place significant constraints on the choice of material selected for the reactor pressure vessel for

  5. Solution of heat removal from nuclear reactors by natural convection

    Directory of Open Access Journals (Sweden)

    Zitek Pavel

    2014-03-01

    Full Text Available This paper summarizes the basis for the solution of heat removal by natural convection from both conventional nuclear reactors and reactors with fuel flowing coolant (such as reactors with molten fluoride salts MSR.The possibility of intensification of heat removal through gas lift is focused on. It might be used in an MSR (Molten Salt Reactor for cleaning the salt mixture of degassed fission products and therefore eliminating problems with iodine pitting. Heat removal by natural convection and its intensification increases significantly the safety of nuclear reactors. Simultaneously the heat removal also solves problems with lifetime of pumps in the primary circuit of high-temperature reactors.

  6. Collective control of a nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rognin, L.

    1995-06-01

    Nowadays, mainly related to the increasing complexity of working environments, working activities become more and collective. The present research on the paradoxical nature of working teams, considered from a reliability point of view. This document is composed of four Sections. The first Section introduces the context of the research, its objectives and the underlying assumptions. In the second Section, we describe a working situation, which is the control of a nuclear reactor. Relations between cooperative work and reliability are discussed in the third Section. Finally, in the fourth Section, a synthesis of the research and some perspectives are proposed. (authors). 7 refs.

  7. NaK flow control by electromagnetic pump of the SNAP-10A space nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Borges, Eduardo M.; Braz Filho, Francisco A.; Guimaraes, Lamartine N.F., E-mail: eduardo@ieav.cta.br, E-mail: fbraz@ieav.cta.br, E-mail: guimarae@ieav.cta.br [Instituto de Estudos Avancados (IEAv/DCTA), Sao Jose dos Campos, SP (Brazil)

    2013-07-01

    In the Institute for Advanced Studies (IEAv), were developed, successfully, the first two direct current electromagnetic pumps (DC EM pumps) of Brazil. The first was built with C-type magnet and coils; and the second, with Samarium-Cobalt permanent magnets, for magnetic field generation. Both were tested and performed quite satisfactory. The electromagnetic pump uses the Faraday principle, in which the interaction of the magnetic field and electric current generates the magneto-motive force, which produces the circulation of the fluid. This type of equipment may be used for controlling the liquid metal flow in nuclear space fast reactors. This paper shows the computer programs developed for design and evaluation of DC EM and electromagnetic thermoelectric (EMTE) pumps, the DC EM pump of Samarium-Cobalt magnets data to Mercury loop flow control, the EMTE pump of SNAP space nuclear reactor. It also compares the theoretical results to experimental data of NaK primary loop flow control by electromagnetic thermoelectric pump of the SNAP-10A space nuclear reactor, with satisfactory results, confirming the viability of the electromagnetic pumps evaluation scheme. (author)

  8. Activities in the field of small nuclear power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Baranaev, Yu.D.; Dolgov, V.V.; Sergeev, Yu.A. [Physics and Power Eng. Inst., Obninsk (Russian Federation). State Res. Centre

    1997-10-01

    Considerable efforts have been undertaken for development, design, construction and operation of small nuclear power plants (SNPP) in Russia. Systematic work in this area was started in the mid-1950s. The driving force for this activity was the awareness that the use of nuclear fuel would practically solve the problem of fuel transportation. As far as the remote northern regions are concerned, this provides the key advantage of nuclear over conventional energy sources. The activity in the field of SNPP has included pre-design analytical feasibility studies and experimental research including large-scale experiments on critical assemblies, thermal and hydraulic test facilities, research and development work, construction and operation of pilot and demonstration SNPPs, and finally, construction and more than 20 years of operation of the commercial SNPP, namely Bilibino nuclear co-generation plant (NCGP) located in Chukotka autonomous district, which is one of the most remote regions in the far north-east of Russia. In recent years, studies have been carried out on the development of several new SNPP designs using advanced reactors of the new generation. Among these are the second stage of Bilibino NCGP, floating NCGP VOLNOLOM-3, designated for siting in the Arctic sea coast area, and a nuclear district heating plant for the town of Apatity, in the Murmansk region. In this paper, the background and current status of the SNPPs are given, and the problems as well as prospects of small nuclear reactors development and implementation are considered. (orig.) 20 refs.

  9. Irradiation rigs in material testing reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rozenblum, F.; Gonnier, C.; Bignan, G. [CEA, Research Centers of Saclay and Cadarache (France)

    2011-07-01

    Osiris is a research reactor with a thermal power of 70 MW. It is a light-water reactor, open-core pool type, the principal aim of which is to carry out tests and irradiate structural materials and fuel elements of nuclear power plants under a high flux of neutrons, and to produce radioisotopes. Osiris operates around 200 days a year, in cycles of varying lengths from 3 to 4 weeks. A shutdown of about 10 days between two cycles allows reloading the core with fuel. Mainly 2 types of irradiation device are present: capsules for materials irradiation (CHOUCA and IRMA devices) and fuels irradiation loops (GRIFFONOS and ISABELLE). Although Osiris is still providing experiments of very good quality, it is facing obsolescence due to its ageing. Osiris is planned to be shut down during next decade. Consequently, it has been decided to launch the construction of the Jules Horowitz Reactor (JHR) in Cadarache. JHR is a water cooled reactor which provides the necessary flexibility and accessibility to manage several highly instrumented experiments, reproducing different reactor environments (water, gas or liquid metal loops), generating transient regimes (key for safety). The JHR facility includes the reactor building, including core, cooling system and the experimental bunkers connected to the core through pool wall penetrations and the auxiliary building, including pools and hot cells necessary for the experimental irradiation process. JHR core is optimised to produce high fast neutron flux to study structural material ageing and high thermal neutrons flux for fuel experiments. The conception of this first fleet of devices integrates the operational experience accumulated by the existing MTR and specifically the Osiris one

  10. Environmental concerns regarding a materials test reactor fuel fabrication facility at the Nuclear and Energy Research Institute - IPEN; Atomos para el desarrollo de Mexico

    Energy Technology Data Exchange (ETDEWEB)

    Santos, G. R. T.; Durazzo, M.; Carvalho, E. F. U. [IPEN, CNEN-SP, P.O. Box 11049, CEP 05422-970, Sao Paulo (Brazil); Riella, H. G. [Universidade Federal de Santa Catarina, Departamento de Engenharia Quimica, Campus Universitario, Florianopolis, CEP 88040-900 (Brazil)]. e-mail: grsantos@ipen.br

    2008-07-01

    The aim of the industrial activities success, front to a more and more informed and demanding society and to a more and more competitive market demands an environmental administration policy which doesn't limit itself to assist the legislation but anticipate and prevent, in a responsible way, possible damages to the environment. One of the maim programs of the Institute of Energetic and Nuclear Research of the national Commission of Nuclear Energy located in Brazil, through the Center of Nuclear Fuel - CCN - is to manufacture MTR-type fuel elements using low-enrichment uranium (20 wt% {sup 2}35U), to supply its IEA-RI research reactor. Integrated in this program, this work aims at well developing and assuring a methodology to implant an environment, health and safety policy, foreseeing its management with the use of detailed data reports and through the adoption of new tools for improving the management, in order to fulfil the applicable legislation and accomplish all the environmental, operational and works aspects. The applied methodology for the effluents management comprises different aspects, including the specific environmental legislation of a country, main available effluents treatment techniques, process flow analyses from raw materials and intakes to products, generated effluents, residuals and emissions. Data collections were accomplished for points gathering and tests characterization, classification and compatibility of the generated effluents and their eventual environmental impacts. This study aims to implant the Sustainable Concept in order to guarantee access to financial resources, allowing cost reduction, maximizing long-term profits, preventing and reducing environmental accident risks and stimulating both the attraction and the keeping of a motivated manpower. Work on this project has already started and, even though many technical actions have not still ended, the results have being extremely valuable. These results can already give to

  11. Request for Naval Reactors Comment on Proposed Prometheus Space Flight Nuclear Reactor High Tier Reactor Safety Requirements and for Naval Reactors Approval to Transmit These Requirements to JPL

    Energy Technology Data Exchange (ETDEWEB)

    D. Kokkinos

    2005-04-28

    The purpose of this letter is to request Naval Reactors comments on the nuclear reactor high tier requirements for the PROMETHEUS space flight reactor design, pre-launch operations, launch, ascent, operation, and disposal, and to request Naval Reactors approval to transmit these requirements to Jet Propulsion Laboratory to ensure consistency between the reactor safety requirements and the spacecraft safety requirements. The proposed PROMETHEUS nuclear reactor high tier safety requirements are consistent with the long standing safety culture of the Naval Reactors Program and its commitment to protecting the health and safety of the public and the environment. In addition, the philosophy on which these requirements are based is consistent with the Nuclear Safety Policy Working Group recommendations on space nuclear propulsion safety (Reference 1), DOE Nuclear Safety Criteria and Specifications for Space Nuclear Reactors (Reference 2), the Nuclear Space Power Safety and Facility Guidelines Study of the Applied Physics Laboratory.

  12. MIRENE, A Mini-Nuclear Reactor For Neutronography--Data And Applications

    Science.gov (United States)

    Houelle, M.; Gerberon, J. M.

    1983-08-01

    MIRENE is a MIni nuclear REactor for NEutronography. In the first part of this paper MIRENE is described and its characteristics are given. The core uses only 1 kg of enriched uranium in solution state. It works in a self-limited pulse mode. The neutron pulses are collimated in two beams which cross the concrete pro-tection walls surrounding the reactor. The main characteristics are : . peak power : 161 kW . exponential rise time : .87 sec . overall energy in a pulse : 2.9 MJ (6.8 x 1016 fissions) . axial beam : - exposure aera : 30 cm x 30 cm - useful fluence per pulse : thermal neutrons : 9 x 108 n/cm2 ; γ-rays : 22 rads (Cd ratio on gold detector : 2) . lateral beam : - exposure aera : 18 cm x 24 cm - useful fluence per pulse : thermal neutrons : 2.6 x 108 n/cm2 ; γ-rays : . 7 rad (Cd ratio : 9). In the second part of the paper, many applications of MIRENE in much different fields are indicated. The results we have obtained since MIRENE started to operate, in 1977, are shown : - In nuclear engineering :.testing of first neutron reactor fuel-pins .control of "neutrophage screens" used in transport and storage of nuclear-fuel materials to secure the criticity-safety .observation of irradiated-oxyde samples in order to determine the Equation of State of the fuel used in fast-neutron reactors .observation of UO2-H20 mixing conditions in the field of cri-ticity experiments - In engineering, MIRENE has a large field of applications, two examples are given : . the control of the sealing of an electric isolator . the visualization of the bonding layer between two high density metals - Finally we show an original application in agronomy which has given very good results : the observation of the in-situ-growth of a corn-root. All these results prove that MIRENE as well as similar reactors can bring about an important contribution as Non-Destructive-Testing stools in the most large field of applications. Their simplicity of design and working connected to their intrinsic

  13. Common-Cause Failure Analysis for Reactor Protection System Reliability Studies

    Energy Technology Data Exchange (ETDEWEB)

    Gentillon, C.; Rasmuson, D.; Eide, S.; Wierman, T.

    1999-08-01

    Analyses were performed of the safety-related performance of the reactor protection system (RPS) at U.S. Westinghouse and General Electric commercial reactors during the period 1984 through 1995. RPS operational data from these reactors were collected from the Nuclear Plant Reliability Data System (NPRDS) and Licensee Event Reports (LER). The common-cause failure (CCF) modeling in the fault trees developed for these studies and the analysis and use of common-cause failure data were sophisticated, state-of-the-art efforts. The overall CCF effort helped to test and expand the limits of the U.S. Nuclear Regulatory Commission's CCF methodology.

  14. High Fidelity Thermal Simulators for Non-Nuclear Testing: Analysis and Initial Results

    Science.gov (United States)

    Bragg-Sitton, Shannon M.; Dickens, Ricky; Dixon, David

    2007-01-01

    Non-nuclear testing can be a valuable tool in the development of a space nuclear power system, providing system characterization data and allowing one to work through various fabrication, assembly and integration issues without the cost and time associated with a full ground nuclear test. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Testing with non-optimized heater elements allows one to assess thermal, heat transfer, and stress related attributes of a given system, but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. High fidelity thermal simulators that match both the static and the dynamic fuel pin performance that would be observed in an operating, fueled nuclear reactor can vastly increase the value of non-nuclear test results. With optimized simulators, the integration of thermal hydraulic hardware tests with simulated neutronie response provides a bridge between electrically heated testing and fueled nuclear testing, providing a better assessment of system integration issues, characterization of integrated system response times and response characteristics, and assessment of potential design improvements' at a relatively small fiscal investment. Initial conceptual thermal simulator designs are determined by simple one-dimensional analysis at a single axial location and at steady state conditions; feasible concepts are then input into a detailed three-dimensional model for comparison to expected fuel pin performance. Static and dynamic fuel pin performance for a proposed reactor design is determined using SINDA/FLUINT thermal analysis software, and comparison is made between the expected nuclear performance and the performance of conceptual thermal simulator designs. Through a series of iterative analyses, a conceptual high fidelity design can developed. Test results presented in this paper correspond to a "first cut" simulator design for a potential

  15. Effects of Nuclear Energy on Sustainable Development and Energy Security: Sodium-Cooled Fast Reactor Case

    Directory of Open Access Journals (Sweden)

    Sungjoo Lee

    2016-09-01

    Full Text Available We propose a stepwise method of selecting appropriate indicators to measure effects of a specific nuclear energy option on sustainable development and energy security, and also to compare an energy option with another. Focusing on the sodium-cooled fast reactor, one of the highlighted Generation IV reactors, we measure and compare its effects with the standard pressurized water reactor-based nuclear power, and then with coal power. Collecting 36 indicators, five experts select seven key indicators to meet data availability, nuclear energy relevancy, comparability among energy options, and fit with Korean energy policy objectives. The results show that sodium-cooled fast reactors is a better alternative than existing nuclear power as well as coal electricity generation across social, economic and environmental dimensions. Our method makes comparison between energy alternatives easier, thereby clarifying consequences of different energy policy decisions.

  16. The reactor ALLEGRO and the sustainable nuclear energy in Central Europe

    Directory of Open Access Journals (Sweden)

    Gadó János

    2014-01-01

    Full Text Available The Visegrád-4 countries (CZ, HU, PL and SK would like to use nuclear energy on the long run. The construction of new Generation 3+ nuclear units probably belong in each country to this realm. These reactors will provide safe and cheap electric energy approximately until the end of the 21st century. In order to use nuclear energy in the 22nd century, sustainability of fuel supply shall be achieved by applying Generation 4 breeder reactors with fast spectrum. The corresponding research and development is organized now in the framework of the V4G4 Centre of Excellence establshed by the nuclear research institutes of the region with a strong technical support from the French CEA. The most important milestone of these efforts is the start-up of the ALLEGRO reactor that shall demonstrate the viability of the gas cooled fast reactor technology.

  17. Nuclear energy and opportunity to strengthen the sustainable electricity sector; Energia nuclear una oportunidad para fortalecer el sector electrico sustentable

    Energy Technology Data Exchange (ETDEWEB)

    Robles N, A. G. [Comision Federal de Electricidad, Direccion de Proyectos de Inversion Financiada, Gerencia de Proteccion Ambiental, Paseo de la Reforma No. 164, Col. Juarez, 06600 Ciudad de Mexico (Mexico)

    2016-09-15

    The beginning of electricity in Mexico was through the use and exploitation of natural resources; as the demand grew, more generation power plants were required with great capacity and at the same time the fuels used varied, although, oil continued to be the main fuel. At present, due to the effects of climate change, the Conference of the Parties has proposed to reduce the consumption of fossil fuels to give way to clean energy (wind, solar, geothermal, nuclear, etc.), which entails gradually modifying the energy matrix of the electricity sector. The National Development Plan and the National Electricity Sector Development Program, this coordinated by the Energy Secretariat in Mexico, establish policies to promote sustainable development, increasing electricity generation through clean energy sources, including nuclear energy. However, such plans are not accurate in the strategy to be followed to ensure compliance with the increased participation of nuclear energy. This article proposes a nuclear program for the Mexican electricity sector, under the terms of a State policy, aimed at crystallizing a sustainable electricity development 2015-2036; considering that the application to the electricity sector constitutes a representative and justified example of the incorporation of environmental aspects in decision processes for the preservation of the environment. In order to determine the quantity and type of reactors, as well as the number of nuclear power plants and increase of the installed capacity, the general planning scheme of the electric sector was used, taking as reference the modeling criteria of the WASP planning system. Finally, is concluded that the electricity generated by fission of radioactive elements is an opportunity to fulfill the commitments made by Mexico at COP 21 and to meet in an environmentally friendly way the energy requirement that our country needs. (Author)

  18. Georgia Tech Studies of Sub-Critical Advanced Burner Reactors with a D-T Fusion Tokamak Neutron Source for the Transmutation of Spent Nuclear Fuel

    Science.gov (United States)

    Stacey, W. M.

    2009-09-01

    The possibility that a tokamak D-T fusion neutron source, based on ITER physics and technology, could be used to drive sub-critical, fast-spectrum nuclear reactors fueled with the transuranics (TRU) in spent nuclear fuel discharged from conventional nuclear reactors has been investigated at Georgia Tech in a series of studies which are summarized in this paper. It is found that sub-critical operation of such fast transmutation reactors is advantageous in allowing longer fuel residence time, hence greater TRU burnup between fuel reprocessing stages, and in allowing higher TRU loading without compromising safety, relative to what could be achieved in a similar critical transmutation reactor. The required plasma and fusion technology operating parameter range of the fusion neutron source is generally within the anticipated operational range of ITER. The implications of these results for fusion development policy, if they hold up under more extensive and detailed analysis, is that a D-T fusion tokamak neutron source for a sub-critical transmutation reactor, built on the basis of the ITER operating experience, could possibly be a logical next step after ITER on the path to fusion electrical power reactors. At the same time, such an application would allow fusion to contribute to meeting the nation's energy needs at an earlier stage by helping to close the fission reactor nuclear fuel cycle.

  19. Electrical studies at the proposed Wahmonie and Calico Hills nuclear waste sites, Nevada Test Site, Nye County, Nevada

    Science.gov (United States)

    Hoover, D.B.; Chornack, Michael P.; Nervick, K.H.; Broker, M.M.

    1982-01-01

    Two sites in the southwest quadrant of the Nevada Test Site (NTS) were investigated as potential repositories for high-level nuclear waste. These are designated the Wahmonie and Calico Hills sites. The emplacement medium at both sites was to be an inferred intrusive body at shallow depth; the inference of the presence of the body was based on aeromagnetic and regional gravity data. This report summarizes results of Schlumberger VES, induced polarization dipole-dipole traverses and magnetotelluric soundings made in the vicinity of the sites in order to characterize the geoelectric section. At the Wahmonie site VES work identified a low resistivity unit at depth surrounding the inferred intrusive body. The low resistivity unit is believed to be either the argillite (Mississippian Eleana Formation) or a thick unit of altered volcanic rock (Tertiary). Good electrical contrast is provided between the low resistivity unit and a large volume of intermediate resistivity rock correlative with the aeromagnetic and gravity data. The intermediate resistivity unit (100-200 ohm-m) is believed to be the intrusive body. The resistivity values are very low for a fresh, tight intrusive and suggest significant fracturing, alteration and possible mineralization have occurred within the upper kilometer of rock. Induced polarization data supports the VES work, identifies a major fault on the northwest side of the inferred intrusive and significant potential for disseminated mineralization within the body. The mineralization potential is particularly significant because as late as 1928, a strike of high grade silver-gold ore was made at the site. The shallow electrical data at Calico Hills revealed no large volume high resistivity body that could be associated with a tight intrusive mass in the upper kilometer of section. A drill hole UE 25A-3 sunk to 762 m (2500 ft) at the site revealed only units of the Eleana argillite thermally metamorphosed below 396 m (1300 ft) and in part highly

  20. A combined gas cooled nuclear reactor and fuel cell cycle

    Science.gov (United States)

    Palmer, David J.

    Rising oil costs, global warming, national security concerns, economic concerns and escalating energy demands are forcing the engineering communities to explore methods to address these concerns. It is the intention of this thesis to offer a proposal for a novel design of a combined cycle, an advanced nuclear helium reactor/solid oxide fuel cell (SOFC) plant that will help to mitigate some of the above concerns. Moreover, the adoption of this proposal may help to reinvigorate the Nuclear Power industry while providing a practical method to foster the development of a hydrogen economy. Specifically, this thesis concentrates on the importance of the U.S. Nuclear Navy adopting this novel design for its nuclear electric vessels of the future with discussion on efficiency and thermodynamic performance characteristics related to the combined cycle. Thus, the goals and objectives are to develop an innovative combined cycle that provides a solution to the stated concerns and show that it provides superior performance. In order to show performance, it is necessary to develop a rigorous thermodynamic model and computer program to analyze the SOFC in relation with the overall cycle. A large increase in efficiency over the conventional pressurized water reactor cycle is realized. Both sides of the cycle achieve higher efficiencies at partial loads which is extremely important as most naval vessels operate at partial loads as well as the fact that traditional gas turbines operating alone have poor performance at reduced speeds. Furthermore, each side of the cycle provides important benefits to the other side. The high temperature exhaust from the overall exothermic reaction of the fuel cell provides heat for the reheater allowing for an overall increase in power on the nuclear side of the cycle. Likewise, the high temperature helium exiting the nuclear reactor provides a controllable method to stabilize the fuel cell at an optimal temperature band even during transients helping

  1. Investigation of Thermal Hydraulics of a Nuclear Reactor Moderator

    Science.gov (United States)

    Sarchami, Araz

    A three-dimensional numerical modeling of the thermo hydraulics of Canadian Deuterium Uranium (CANDU) nuclear reactor is conducted. The moderator tank is a Pressurized heavy water reactor which uses heavy water as moderator in a cylindrical tank. The main use of the tank is to bring the fast neutrons to the thermal neutron energy levels. The moderator tank compromises of several bundled tubes containing nuclear rods immersed inside the heavy water. It is important to keep the water temperature in the moderator at sub-cooled conditions, to prevent potential failure due to overheating of the tubes. Because of difficulties in measuring flow characteristics and temperature conditions inside a real reactor moderator, tests are conducted using a scaled moderator in moderator test facility (MTF) by Chalk River Laboratories of Atomic Energy of Canada Limited (CRL, AECL). MTF tests are conducted using heating elements to heat tube surfaces. This is different than the real reactor where nuclear radiation is the source of heating which results in a volumetric heating of the heavy water. The data recorded inside the MTF tank have shown levels of fluctuations in the moderator temperatures and requires in depth investigation of causes and effects. The purpose of the current investigation is to determine the causes for, and the nature of the moderator temperature fluctuations using three-dimensional simulation of MTF with both (surface heating and volumetric heating) modes. In addition, three dimensional simulation of full scale actual moderator tank with volumetric heating is conducted to investigate the effects of scaling on the temperature distribution. The numerical simulations are performed on a 24-processor cluster using parallel version of the FLUENT 12. During the transient simulation, 55 points of interest inside the tank are monitored for their temperature and velocity fluctuations with time.

  2. 77 FR 36014 - Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors

    Science.gov (United States)

    2012-06-15

    ... COMMISSION Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors AGENCY: Nuclear...-1277, ``Initial Test Program of Emergency Core Cooling Systems for Boiling- Water Reactors.'' This... testing features of emergency core cooling systems (ECCSs) for boiling-water reactors (BWRs). DATES...

  3. Generalized and Stability Rational Functions for Dynamic Systems of Reactor Kinetics

    Directory of Open Access Journals (Sweden)

    Ahmed E. Aboanber

    2013-01-01

    Full Text Available The base of reactor kinetics dynamic systems is a set of coupled stiff ordinary differential equations known as the point reactor kinetics equations. These equations which express the time dependence of the neutron density and the decay of the delayed neutron precursors within a reactor are first order nonlinear and essentially describe the change in neutron density within the reactor due to a change in reactivity. Outstanding the particular structure of the point kinetic matrix, a semianalytical inversion is performed and generalized for each elementary step resulting eventually in substantial time saving. Also, the factorization techniques based on using temporarily the complex plane with the analytical inversion is applied. The theory is of general validity and involves no approximations. In addition, the stability of rational function approximations is discussed and applied to the solution of the point kinetics equations of nuclear reactor with different types of reactivity. From the results of various benchmark tests with different types of reactivity insertions, the developed generalized Padé approximation (GPA method shows high accuracy, high efficiency, and stable character of the solution.

  4. ANALISIS TRANSIEN PADA FIXED BED NUCLEAR REACTOR

    Directory of Open Access Journals (Sweden)

    M. Rizaal

    2015-03-01

    Full Text Available Desain teras Fixed Bed Nuclear Reactor (FBNR yang modular memungkinkan pengendalian daya dapat dilakukan dengan mengatur ketinggian suspended core dan laju aliran massa pendingin. Tujuan penelitian ini adalah mempelajari perubahan daya termal teras sebagai akibat perubahan laju aliran massa pendingin yang masuk ke teras reaktor dan perubahan ketinggian suspended core serta mempelajari karakteristik keselamatan melekat yang dimiliki FBNR saat terjadi kegagalan pelepasan kalor (loss of heat sink. Keadaan neutronik teras dimodelkan pada kondisi tunak dengan menggunakan paket program Standard Reactor Analysis Code (SRAC untuk memperoleh data fluks neutron, konstanta grup, fraksi neutron kasip, konstanta peluruhan prekursor neutron kasip, dan beberapa parameter teras penting lainnya. Selanjutnya data tersebut digunakan pada perhitungan transien sebagai syarat awal. Analisis transien dilakukan pada tiga kondisi, yaitu saat terjadi penurunan laju aliran massa pendingin, saat terjadi penurunan ketinggian suspended core, dan saat terjadi kegagalan sistem pelepasan kalor. Hasil yang diperoleh dari penelitian ini menunjukkan bahwa penurunan laju aliran massa pendingin sebesar 50%, dari kondisi normal, menyebabkan daya termal teras turun 28% dibanding daya sebelumnya. Penurunan ketinggian suspended core sebesar 30% dari ketinggian normal menyebabkan daya termal teras turun 17% dibanding daya sebelumnya. Sementara untuk kondisi kegagalan sistem pelepasan kalor, daya termal teras mengalami penurunan sebesar 76%. Dengan demikian, pengendalian daya pada FBNR dapat dilakukan dengan mengatur laju aliran massa pendingin dan ketinggian suspended core, serta keselamatan melekat yang handal pada kondisi kegagalan sistem pelepasan kalor. Kata kunci: FBNR, transien, daya, laju aliran massa, suspended core Modular in design enables Fixed Bed Nuclear Reactor (FBNR power controlled by the adjustment of suspended core and coolant flow rate. The main purposes of this paper

  5. Multiscale Methods for Nuclear Reactor Analysis

    Science.gov (United States)

    Collins, Benjamin S.

    The ability to accurately predict local pin powers in nuclear reactors is necessary to understand the mechanisms that cause fuel pin failure during steady state and transient operation. In the research presented here, methods are developed to improve the local solution using high order methods with boundary conditions from a low order global solution. Several different core configurations were tested to determine the improvement in the local pin powers compared to the standard techniques, that use diffusion theory and pin power reconstruction (PPR). Two different multiscale methods were developed and analyzed; the post-refinement multiscale method and the embedded multiscale method. The post-refinement multiscale methods use the global solution to determine boundary conditions for the local solution. The local solution is solved using either a fixed boundary source or an albedo boundary condition; this solution is "post-refinement" and thus has no impact on the global solution. The embedded multiscale method allows the local solver to change the global solution to provide an improved global and local solution. The post-refinement multiscale method is assessed using three core designs. When the local solution has more energy groups, the fixed source method has some difficulties near the interface: however the albedo method works well for all cases. In order to remedy the issue with boundary condition errors for the fixed source method, a buffer region is used to act as a filter, which decreases the sensitivity of the solution to the boundary condition. Both the albedo and fixed source methods benefit from the use of a buffer region. Unlike the post-refinement method, the embedded multiscale method alters the global solution. The ability to change the global solution allows for refinement in areas where the errors in the few group nodal diffusion are typically large. The embedded method is shown to improve the global solution when it is applied to a MOX/LEU assembly

  6. Nuclear reactors built, being built, or planned 1993

    Energy Technology Data Exchange (ETDEWEB)

    1993-08-01

    Nuclear Reactors Built, Being Built, or Planned contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of December 31, 1993. The Office of Scientific and Technical Information, US Department of Energy, gathers this information annually from Washington headquarters and field offices of DOE; from the US Nuclear Regulatory Commission (NRC); from the US reactor manufacturers who are the principal nuclear embassies; and from foreign governmental nuclear departments. The book consists of three divisions, as follows: (1) a commercial reactor locator map and tables of the characteristic and statistical data that follow; a table of abbreviations; (2) tables of data for reactors operating, being built, or planned; and (3) tables of data for reactors that have been shut down permanently or dismantled. The reactors are subdivided into the following parts: civilian, production, military, export, and critical assembly.

  7. Nuclear waste disposal utilizing a gaseous core reactor

    Science.gov (United States)

    Paternoster, R. R.

    1975-01-01

    The feasibility of a gaseous core nuclear reactor designed to produce power to also reduce the national inventories of long-lived reactor waste products through nuclear transmutation was examined. Neutron-induced transmutation of radioactive wastes is shown to be an effective means of shortening the apparent half life.

  8. Visual interface for the automation of the instrumented pendulum of Charpy tests used in the surveillance program of reactors vessel of nuclear power plants; Interfase visual para la automatizacion del pendulo instrumentado de pruebas Charpy utilizado en el programa de vigilancia de la vasija de reactores de centrales nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Rojas S, A.S.; Sainz M, E.; Ruiz E, J.A. [ININ, Carretera Mexico-Toluca Km.36.5, Mpio. de Ocoyoacac, Estado de Mexico (Mexico)]. E-mail: asrs@nuclear.inin.mx; esm@nuclear.inin.mx; jare@nuclear.inin.mx

    2004-07-01

    Inside the Programs of Surveillance of the nuclear power stations periodic information is required on the state that keep the materials with those that builds the vessel of the reactor. This information is obtained through some samples or test tubes that are introduced inside the core of the reactor and it is observed if its physical characteristics remain after having been subjected to the radiation changes and temperature. The rehearsal with the instrumented Charpy pendulum offers information on the behavior of fracture dynamics of a material. In the National Institute of Nuclear Research (ININ) it has an instrumented Charpy pendulum. The operation of this instrument is manual, having inconveniences to carry out rehearsals with radioactive material, handling of high and low temperatures, to fulfill the normative ones for the realization of the rehearsals, etc. In this work the development of a computational program is presented (virtual instrument), for the automation of the instrumented pendulum. The system has modules like: Card of data acquisition, signal processing, positioning system, tempered system, pneumatic system, compute programs like it is the visual interface for the operation of the instrumented Charpy pendulum and the acquisition of impact signals. This system shows that given the characteristics of the nuclear industry with radioactive environments, the virtual instrumentation and the automation of processes can contribute to diminish the risks to the personnel occupationally exposed. (Author)

  9. Modular Lead-Bismuth Fast Reactors in Nuclear Power

    OpenAIRE

    Vladimir Petrochenko; Georgy Toshinsky

    2012-01-01

    On the basis of the unique experience of operating reactors with heavy liquid metal coolant–eutectic lead-bismuth alloy in nuclear submarines, the concept of modular small fast reactors SVBR-100 for civilian nuclear power has been developed and validated. The features of this innovative technology are as follows: a monoblock (integral) design of the reactor with fast neutron spectrum, which can operate using different types of fuel in various fuel cycles including MOX fuel in a self-providing...

  10. Spent nuclear fuel discharges from U.S. reactors 1994

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-02-01

    Spent Nuclear Fuel Discharges from US Reactors 1994 provides current statistical data on fuel assemblies irradiated at commercial nuclear reactors operating in the US. This year`s report provides data on the current inventories and storage capacities at these reactors. Detailed statistics on the data are presented in four chapters that highlight 1994 spent fuel discharges, storage capacities and inventories, canister and nonfuel component data, and assembly characteristics. Five appendices, a glossary, and bibliography are also included. 10 figs., 34 tabs.

  11. A generalized perturbation program for CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-12-31

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

  12. Uncertainties in the Anti-neutrino Production at Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Djurcic, Zelimir; Detwiler, Jason A.; Piepke, Andreas; Foster Jr., Vince R.; Miller, Lester; Gratta, Giorgio

    2008-08-06

    Anti-neutrino emission rates from nuclear reactors are determined from thermal power measurements and fission rate calculations. The uncertainties in these quantities for commercial power plants and their impact on the calculated interaction rates in {bar {nu}}{sub e} detectors is examined. We discuss reactor-to-reactor correlations between the leading uncertainties, and their relevance to reactor {bar {nu}}{sub e} experiments.

  13. Nuclear reactor materials at the atomic scale

    Directory of Open Access Journals (Sweden)

    Emmanuelle A. Marquis

    2009-11-01

    Full Text Available With the renewed interest in nuclear energy, developing new materials able to respond to the stringent requirements of the next-generation fission and future fusion reactors has become a priority. An efficient search for such materials requires detailed knowledge of material behaviour under irradiation, high temperatures and corrosive environments. Minimizing the rates of materials degradation will be possible only if the mechanisms by which it occurs are understood. Atomic-scale experimental probing as well as modelling can provide some answers and help predict in-service behaviour. This article illustrates how this approach has already improved our understanding of precipitation under irradiation, corrosion behaviour, and stress corrosion cracking. It is also now beginning to provide guidance for the development of new alloys.

  14. Neutron measurements at nuclear power reactors [55

    CERN Document Server

    Scherpelz, R I

    2002-01-01

    Staff from the Pacific Northwest National Laboratory (operated by Battelle Memorial Institute), have performed neutron measurements at a number of commercial nuclear power plants in the United States. Neutron radiation fields at light water reactor (LWR) power plants are typically characterized by low-energy distributions due to the presence of large amounts of scattering material such as water and concrete. These low-energy distributions make it difficult to accurately monitor personnel exposures, since most survey meters and dosimeters are calibrated to higher-energy fields such as those produced by bare or D sub 2 O-moderated sup 2 sup 5 sup 2 Cf sources. Commercial plants typically use thermoluminescent dosimeters in an albedo configuration for personnel dosimetry and survey meters based on a thermal-neutron detector inside a cylindrical or spherical moderator for dose rate assessment, so their methods of routine monitoring are highly dependent on the energy of the neutron fields. Battelle has participate...

  15. Meteodiffusive Characterization of Algiers' Nuclear Research Reactor

    Directory of Open Access Journals (Sweden)

    Mourad Messaci

    2007-01-01

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

  16. Nuclear safety in light water reactors severe accident phenomenology

    CERN Document Server

    Sehgal, Bal Raj

    2011-01-01

    This vital reference is the only one-stop resource on how to assess, prevent, and manage severe nuclear accidents in the light water reactors (LWRs) that pose the most risk to the public. LWRs are the predominant nuclear reactor in use around the world today, and they will continue to be the most frequently utilized in the near future. Therefore, accurate determination of the safety issues associated with such reactors is central to a consideration of the risks and benefits of nuclear power. This book emphasizes the prevention and management of severe accidents to teach nuclear professionals

  17. Construction and operation of parallel electric and magnetic field spectrometers for mass/energy resolved multi-ion charge exchange diagnostics on the Tokamak Fusion Test Reactor

    Science.gov (United States)

    Medley, S. S.; Roquemore, A. L.

    1998-07-01

    A novel charge exchange spectrometer using a dee-shaped region of parallel electric and magnetic fields was developed at the Princeton Plasma Physics Laboratory for neutral particle diagnostics on the Tokamak Fusion Test Reactor (TFTR). The E∥B spectrometer has an energy range of 0.5⩽A (amu)E (keV)⩽600 and provides mass-resolved energy spectra of H+, D+, and T+ (or 3He+) ion species simultaneously during a single discharge. The detector plane exhibits parallel rows of analyzed ions, each row containing the energy dispersed ions of a given mass-to-charge ratio. The detector consists of a large area microchannel plate (MCP) which is provided with three rectangular, semicontinuous active area strips, one coinciding with each of the mass rows for detection of H+, D+, and T+ (or 3He+) and each mass row has 75 energy channels. To suppress spurious signals attending operation of the plate in the magnetic fringe field of the spectrometer, the MCP was housed in a double-walled iron shield with a wire mesh ion entrance window. Using an accelerator neutron generator, the MCP neutron detection efficiency was measured to be 1.7×10-3 and 6.4×10-3 counts/neutron/cm2 for 2.5 MeV-DD and 14 MeV-DT neutrons, respectively. The design and calibration of the spectrometer are described in detail, including the effect of MCP exposure to tritium, and results obtained during high performance D-D operation on TFTR are presented to illustrate the performance of the E∥B spectrometer. The spectrometers were not used during D-T plasma operation due to the cost of providing the required radiation shielding.

  18. Space and Terrestrial Power System Integration Optimization Code BRMAPS for Gas Turbine Space Power Plants With Nuclear Reactor Heat Sources

    Science.gov (United States)

    Juhasz, Albert J.

    2007-01-01

    expected within the next 30 to 50 years, as predicted by the Hubbert model and confirmed by other global energy consumption prognoses. Having invested national resources into the development of NGNP, the technology and experience accumulated during the project needs to be documented clearly and in sufficient detail for young engineers coming on-board at both DOE and NASA to acquire it. Hands on training on reactor operation, test rigs of turbomachinery, and heat exchanger components, as well as computational tools will be needed. Senior scientist/engineers involved with the development of NGNP should also be encouraged to participate as lecturers, instructors, or adjunct professors at local universities having engineering (mechanical, electrical, nuclear/chemical, and/or materials) as one of their fields of study.

  19. Qualification of non-destructive examination for belgian nuclear reactor pressure vessel inspection

    Energy Technology Data Exchange (ETDEWEB)

    Couplet, D. [TRACTEBEL, Brussels (Belgium); Francoise, T. [Intercontrole, 94 - Rungis (France)

    2001-07-01

    In Service Inspection (ISI) participates to the assessment of Nuclear Reactor Pressure Vessel Integrity. The performance of Non Destructive Examination (NDE) techniques must be demonstrated according to predefined objectives. The qualification process is essential to trust the reliability of the information provided by NDE. In Belgian Nuclear Power Plants, the qualification was conducted through a collaboration between the vendor and a technical group from the Electricity Utility. The important facts of this qualification will be presented: - the detailed definition of the inspection and qualifications objectives, based on a combination of the ASME code and the European Methodology for Qualification; - the systematic verification of the NDE performance and limitations, for each ISI objective, through an adequate combination of tests on blocks and technical justification; - the continuous improvement of the NDE procedure; - the feedback and the lessons learnt from site experience; - the necessary multi-disciplinary approach (NDE, degradation mechanisms, structural integrity)

  20. Computation system for nuclear reactor core analysis. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.; Petrie, L.M.

    1977-04-01

    This report documents a system which contains computer codes as modules developed to evaluate nuclear reactor core performance. The diffusion theory approximation to neutron transport may be applied with the VENTURE code treating up to three dimensions. The effect of exposure may be determined with the BURNER code, allowing depletion calculations to be made. The features and requirements of the system are discussed and aspects common to the computational modules, but the latter are documented elsewhere. User input data requirements, data file management, control, and the modules which perform general functions are described. Continuing development and implementation effort is enhancing the analysis capability available locally and to other installations from remote terminals.

  1. Sustainable and safe nuclear fission energy technology and safety of fast and thermal nuclear reactors

    CERN Document Server

    Kessler, Günter

    2012-01-01

    Unlike existing books of nuclear reactor physics, nuclear engineering and nuclear chemical engineering this book covers a complete description and evaluation of nuclear fission power generation. It covers the whole nuclear fuel cycle, from the extraction of natural uranium from ore mines, uranium conversion and enrichment up to the fabrication of fuel elements for the cores of various types of fission reactors. This is followed by the description of the different fuel cycle options and the final storage in nuclear waste repositories. In addition the release of radioactivity under normal and possible accidental conditions is given for all parts of the nuclear fuel cycle and especially for the different fission reactor types.

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

    Energy Technology Data Exchange (ETDEWEB)

    J. K. Wright; R. N. Wright

    2010-07-01

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

  3. The RES Reactor. A test reactor for the French naval propulsion

    Energy Technology Data Exchange (ETDEWEB)

    Pivet, Sylvestre [CEA, Centre de Cadarache, F-13108 Saint Paul lez Durance (France); Minguet, Jean-Luc [AREVA-Technicatome, BP17, 91192 Gif-sur-Yvette (France)

    2006-07-01

    In the Cadarache nuclear research centre the French Atomic Energy Commission (CEA) operates, with the support of TECHNICATOME as nuclear operator, the experimental facilities which are necessary for the French naval propulsion program. Since the sixties these facilities have brought a large contribution to the development and to the technical support for the nuclear propulsion; they have been used also to train the French Navy operators. The last experimental reactor, the RNG, is now at the end of its life cycle after thirty years of a profitable operation. A replacement reactor is needed to sustain any evolution of the naval propulsion reactors as well as to guarantee a safe operation and a high level of availability of the existing onboard reactors. The aim of the RES program is namely to build such a test facility. Its construction program started in 2003. By the year 2009 the RES reactor will take over the mission of the RNG. We present hereafter: - A brief history of the French experimental reactors built in support to the naval propulsion, - The needs of the naval propulsion and the related objectives of the RES program, - The corresponding architecture and main characteristics of the RES facility, - The current status of the RES construction. The contents of the paper is as follows: 1. Introduction; 2. History of the French nuclear propulsion experimental reactors; 3. Needs of the naval propulsion and related objectives of the RES reactor; 4. RES architecture and main characteristics; 4.1. The pool module; 4.2. The reactor module; 4.3. The RES reactor, an innovative concept; 5. Realisation status; 6. Conclusion. To summarize, from the year 2009 the RES will be an efficient facility available for irradiation and qualification programs. Its large experimental capabilities will allow relevant fuel and core irradiations. This will give access to a real progress in the knowledge of fuel and core physics as well as in the related simulation tools. This reactor

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-09-26

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

  5. Design and Transient Analysis of Passive Safety Cooling Systems for Advanced Nuclear Reactors

    OpenAIRE

    Galvez, Cristhian

    2011-01-01

    The Pebble Bed Advanced High Temperature Reactor (PB-AHTR) is a pebble fueled, liquid salt cooled, high temperature nuclear reactor design that can be used for electricity generation or other applications requiring the availability of heat at elevated temperatures. A stage in the design evolution of this plant requires the analysis of the plant during a variety of potential transients to understand the primary and safety cooling system response. This study focuses on the performance of the pa...

  6. A brief history of design studies on innovative nuclear reactors

    Science.gov (United States)

    Sekimoto, Hiroshi

    2014-09-01

    In a short period after the success of CP1, many types of nuclear reactors were proposed and investigated. However, soon only a small number of reactors were selected for practical use. Around 1970, only LWRs with small number of CANDUs were operated in the western world, and FBRs were under development. It was about the time when Apollo moon landing was accomplished. However, at the same time, the future of human being was widely considered pessimistic and Limits to Growth was published. In the end of 1970's the TMI accident occurred and many nuclear reactor contracts were cancelled in USA and any more contracts had not been concluded until recent years. From the reflection of this accident, many Inherent Safe Reactors (ISRs) were proposed, though none of them were constructed. A common idea of ISRs is smallness of their size. Tokyo Institute of Technology (TokyoTech) held a symposium on small reactors, SR/TIT, in 1991, where many types of small ISRs were presented. Recently small reactors attract interest again. The most ideas employed in these reactors were the same discussed in SR/TIT. In 1980's the radioactive wastes from fuel cycle became a severe problem around the world. In TokyoTech, this issue was discussed mainly from the viewpoint of nuclear transmutations. The neutron economy became inevitable for these innovative nuclear reactors especially small long-life reactors and transmutation reactors.

  7. A brief history of design studies on innovative nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Sekimoto, Hiroshi, E-mail: hsekimot@gmail.com [Emeritus Professor, Tokyo Institute of Technology (Japan)

    2014-09-30

    In a short period after the success of CP1, many types of nuclear reactors were proposed and investigated. However, soon only a small number of reactors were selected for practical use. Around 1970, only LWRs with small number of CANDUs were operated in the western world, and FBRs were under development. It was about the time when Apollo moon landing was accomplished. However, at the same time, the future of human being was widely considered pessimistic and Limits to Growth was published. In the end of 1970’s the TMI accident occurred and many nuclear reactor contracts were cancelled in USA and any more contracts had not been concluded until recent years. From the reflection of this accident, many Inherent Safe Reactors (ISRs) were proposed, though none of them were constructed. A common idea of ISRs is smallness of their size. Tokyo Institute of Technology (TokyoTech) held a symposium on small reactors, SR/TIT, in 1991, where many types of small ISRs were presented. Recently small reactors attract interest again. The most ideas employed in these reactors were the same discussed in SR/TIT. In 1980’s the radioactive wastes from fuel cycle became a severe problem around the world. In TokyoTech, this issue was discussed mainly from the viewpoint of nuclear transmutations. The neutron economy became inevitable for these innovative nuclear reactors especially small long-life reactors and transmutation reactors.

  8. Nuclear reactors built, being built, or planned, 1991

    Energy Technology Data Exchange (ETDEWEB)

    Simpson, B.

    1992-07-01

    This document contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of December 31, 1991. The book is divided into three major sections: Section 1 consists of a reactor locator map and reactor tables; Section 2 includes nuclear reactors that are operating, being built, or planned; and Section 3 includes reactors that have been shut down permanently or dismantled. Sections 2 and 3 contain the following classification of reactors: Civilian, Production, Military, Export, and Critical Assembly. Export reactor refers to a reactor for which the principal nuclear contractor is an American company -- working either independently or in cooperation with a foreign company (Part 4, in each section). Critical assembly refers to an assembly of fuel and assembly of fuel and moderator that requires an external source of neutrons to initiate and maintain fission. A critical assembly is used for experimental measurements (Part 5).

  9. 78 FR 66785 - Korea Hydro and Nuclear Power Co., Ltd., and Korea Electric Power Corporation

    Science.gov (United States)

    2013-11-06

    ... APR1400 Standard Plant Design submitted by Korea Hydro and Nuclear Power Co., Ltd. (KHNP) and Korea... COMMISSION Korea Hydro and Nuclear Power Co., Ltd., and Korea Electric Power Corporation AGENCY: Nuclear..., construction, operation and maintenance of the Optimized Power Reactor 1000 (OPR1000), the APR1400...

  10. Theory of neutron slowing down in nuclear reactors

    CERN Document Server

    Ferziger, Joel H; Dunworth, J V

    2013-01-01

    The Theory of Neutron Slowing Down in Nuclear Reactors focuses on one facet of nuclear reactor design: the slowing down (or moderation) of neutrons from the high energies with which they are born in fission to the energies at which they are ultimately absorbed. In conjunction with the study of neutron moderation, calculations of reactor criticality are presented. A mathematical description of the slowing-down process is given, with particular emphasis on the problems encountered in the design of thermal reactors. This volume is comprised of four chapters and begins by considering the problems

  11. C Reactor overbore test facility review

    Energy Technology Data Exchange (ETDEWEB)

    Carlson, P.A.; Nilson, R.

    1964-04-24

    In 1961, large-size, smooth-bore, Zircaloy process tubes were installed in C-Reactor graphite channels that had been enlarged to 2.275 inches. These tubes were installed to provide a test and demonstration facility for the concept of overboring as a means of securing significant improvement in the production capability of the reactors, After two years of facility operation, it is now appropriate to consider the extent to which original objectives have been achieved, to re-examine the original objectives, and to consider the best future use of this unique facility. This report presents the general results of such a review and re-examination in more detail.

  12. Future electricity production methods. Part 1: Nuclear energy

    Science.gov (United States)

    Nifenecker, Hervé

    2011-02-01

    The global warming challenge aims at stabilizing the concentrations of Green House Gas (GHG) in the atmosphere. Carbon dioxide is the most effective of the anthropogenic GHG and is essentially produced by consumption of fossil fuels. Electricity production is the dominant cause of CO2 emissions. It is, therefore, crucial that the share of 'carbon less' electricity production techniques increases at a fast pace. This is the more so, that 'clean' electricity would be useful to displace 'dirty' techniques in other fields such as heat production and transportation. Here we examine the extent to which nuclear energy could be operational in providing 'clean' electricity. A nuclear intensive scenario is shown to give the possibility to divide CO2 emissions by a factor of 2 worldwide, within 50 years. However, the corresponding sharp increase in nuclear power will put a heavy burden on uranium reserves and will necessitate the development of breeding reactors as soon as possible. A review of present and future reactors is given with special attention to the safety issues. The delicate question of nuclear fuel cycle is discussed concerning uranium reserves and management of used fuels. It is shown that dealing with nuclear wastes is more a socio-political problem than a technical one. The third difficult question associated with the development of nuclear energy is the proliferation risk. It is advocated that, while this is, indeed, a very important question, it is only weakly related to nuclear power development. Finally, the possibilities of nuclear fusion are discussed and it is asserted that, under no circumstances, could nuclear fusion give a significant contribution to the solution of the energy problem before 50 years, too late for dealing with the global warming challenge.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-12-01

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

  14. Evaluation of a hydrogen sensor for nuclear reactor containment monitoring

    Energy Technology Data Exchange (ETDEWEB)

    Hoffheins, B.S.; McKnight, T.E.; Lauf, R.J.; Smith, R.R. [Oak Ridge National Lab., TN (United States); James, R.E. [Electric Power Research Inst., Palo Alto, CA (United States)

    1997-02-01

    Measurement of hydrogen concentration in containment atmospheres in nuclear plants is a key safety capability. Current technologies require extensive sampling systems and subsequent maintenance and calibration costs can be very expensive. A new hydrogen sensor has been developed that is small and potentially inexpensive to install and maintain. Its size and low power requirement make it suitable in distributed systems for pinpointing hydrogen buildup. This paper will address the first phase of a testing program conducted to evaluate this sensor for operation in reactor containments.

  15. The role of actinide burning and the Integral Fast Reactor in the future of nuclear power

    Energy Technology Data Exchange (ETDEWEB)

    Hollaway, W.R.; Lidsky, L.M.; Miller, M.M.

    1990-12-01

    A preliminary assessment is made of the potential role of actinide burning and the Integral Fast Reactor (IFR) in the future of nuclear power. The development of a usable actinide burning strategy could be an important factor in the acceptance and implementation of a next generation of nuclear power. First, the need for nuclear generating capacity is established through the analysis of energy and electricity demand forecasting models which cover the spectrum of bias from anti-nuclear to pro-nuclear. The analyses take into account the issues of global warming and the potential for technological advances in energy efficiency. We conclude, as do many others, that there will almost certainly be a need for substantial nuclear power capacity in the 2000--2030 time frame. We point out also that any reprocessing scheme will open up proliferation-related questions which can only be assessed in very specific contexts. The focus of this report is on the fuel cycle impacts of actinide burning. Scenarios are developed for the deployment of future nuclear generating capacity which exploit the advantages of actinide partitioning and actinide burning. Three alternative reactor designs are utilized in these future scenarios: The Light Water Reactor (LWR); the Modular Gas-Cooled Reactor (MGR); and the Integral Fast Reactor (FR). Each of these alternative reactor designs is described in some detail, with specific emphasis on their spent fuel streams and the back-end of the nuclear fuel cycle. Four separation and partitioning processes are utilized in building the future nuclear power scenarios: Thermal reactor spent fuel preprocessing to reduce the ceramic oxide spent fuel to metallic form, the conventional PUREX process, the TRUEX process, and pyrometallurgical reprocessing.

  16. Summary of space nuclear reactor power systems, 1983--1992

    Energy Technology Data Exchange (ETDEWEB)

    Buden, D.

    1993-08-11

    This report summarizes major developments in the last ten years which have greatly expanded the space nuclear reactor power systems technology base. In the SP-100 program, after a competition between liquid-metal, gas-cooled, thermionic, and heat pipe reactors integrated with various combinations of thermoelectric thermionic, Brayton, Rankine, and Stirling energy conversion systems, three concepts:were selected for further evaluation. In 1985, the high-temperature (1,350 K), lithium-cooled reactor with thermoelectric conversion was selected for full scale development. Since then, significant progress has been achieved including the demonstration of a 7-y-life uranium nitride fuel pin. Progress on the lithium-cooled reactor with thermoelectrics has progressed from a concept, through a generic flight system design, to the design, development, and testing of specific components. Meanwhile, the USSR in 1987--88 orbited a new generation of nuclear power systems beyond the, thermoelectric plants on the RORSAT satellites. The US has continued to advance its own thermionic fuel element development, concentrating on a multicell fuel element configuration. Experimental work has demonstrated a single cell operating time of about 1 1/2-y. Technology advances have also been made in the Stirling engine; an advanced engine that operates at 1,050 K is ready for testing. Additional concepts have been studied and experiments have been performed on a variety of systems to meet changing needs; such as powers of tens-to-hundreds of megawatts and highly survivable systems of tens-of-kilowatts power.

  17. Final Physics Report for the Engineering Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wolfe, I. B. [Savannah River Site (SRS), Aiken, SC (United States)

    1956-06-25

    This report is a summary of the physics design work performed on the Engineering Test Reactor. The ETR presents computational difficulties not found in other reactors because of the large number of experimental holes in the core. The physics of the ETR depends strongly upon the contents of the in-core experimental facilities. In order to properly evaluate the reactor; taking into account the experiments in the core, multi-region, two-dimensional calculations are required. These calculations require .the use of a large computer such as the Remington Rand Univac and are complex and expensive enough to warrant a five-stage program: 1. In the early stages of design, only preliminary two-dimensional calculations were performed .in order to obtain a rough idea of the general behavior of the reactor and its critical mass with tentative experiments in place. 2. A large amount of work was carried out in which the reactor was approximated as one with a uniform homogeneous core. With this model, detailed studies were carried out to investigate the feasibility and to obtain general design data on such points as the design and properties of the gray and black-control rods, the design of the beryllium reflector, gamma and neutron heating, the use of burnable poisons, etc. In performing these calculations, use was made of the IBM 650 PROD code obtained from KAPL. 3. With stages 1 and 2 carried out, two-dimensional calculations of the core at start-up conditions were performed on the Univac computer. 4. Detailed two-dimensional calculations of the properties of the ETR with a proposed first set of experiments in place were carried out. 5. A series of nuclear tests were performed at the reactivity measurements facility at the MTR site in order to confirm the validity of the analytical techniques in physics analysis. In performing the two-dimensional Univac calculations, the MUG code developed by KAPL and the Cuthill code developed at the David Taylor Model Basin were utilized. In

  18. The necessity of nuclear reactors for targeted radionuclide therapies.

    Science.gov (United States)

    Krijger, Gerard C; Ponsard, Bernard; Harfensteller, Mark; Wolterbeek, Hubert T; Nijsen, Johannes W F

    2013-07-01

    Nuclear medicine has been contributing towards personalized therapies. Nuclear reactors are required for the working horses of both diagnosis and treatment, i.e., Tc-99m and I-131. In fact, reactors will remain necessary to fulfill the demand for a variety of radionuclides and are essential in the expanding field of targeted radionuclide therapies for cancer. However, the main reactors involved in the global supply are ageing and expected to shut down before 2025. Therefore, the fields of (nuclear) medicine, nuclear industry and politics share a global responsibility, faced with the task to secure future access to suitable nuclear reactors. At the same time, alternative production routes should be industrialized. For this, a coordinating entity should be put into place.

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

    Energy Technology Data Exchange (ETDEWEB)

    J. K. Wright; R. N. Wright

    2008-04-01

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

  20. Nuclear Systems Enhanced Performance Program, Maintenance Cycle Extension in Advanced Light Water Reactor Design

    Energy Technology Data Exchange (ETDEWEB)

    Professor Neill Todreas

    2001-10-01

    A renewed interest in new nuclear power generation in the US has spurred interest in developing advanced reactors with features which will address the public's concerns regarding nuclear generation. However, it is economic performance which will dictate whether any new orders for these plants will materialize. Economic performance is, to a great extent, improved by maximizing the time that the plant is on-line generating electricity relative to the time spent off-line conducting maintenance and refueling. Indeed, the strategy for the advanced light water reactor plant IRIS (International Reactor, Innovative and Secure) is to utilize an eight year operating cycle. This report describes a formalized strategy to address, during the design phase, the maintenance-related barriers to an extended operating cycle. The top-level objective of this investigation was to develop a methodology for injecting component and system maintainability issues into the reactor plant design process to overcome these barriers. A primary goal was to demonstrate the applicability and utility of the methodology in the context of the IRIS design. The first step in meeting the top-level objective was to determine the types of operating cycle length barriers that the IRIS design team is likely to face. Evaluation of previously identified regulatory and investment protection surveillance program barriers preventing a candidate operating PWR from achieving an extended (48 month) cycle was conducted in the context of the IRIS design. From this analysis, 54 known IRIS operating cycle length barriers were identified. The resolution methodology was applied to each of these barriers to generate design solution alternatives for consideration in the IRIS design. The methodology developed has been demonstrated to narrow the design space to feasible design solutions which enable a desired operating cycle length, yet is general enough to have broad applicability. Feedback from the IRIS design team

  1. Decision-support tool for assessing future nuclear reactor generation portfolios

    NARCIS (Netherlands)

    S. Jain (Shashi); F Roelofs; C.W. Oosterlee (Cornelis)

    2014-01-01

    htmlabstractCapital costs, fuel, operation and maintenance (O&M) costs, and electricity prices play a key role in the economics of nuclear power plants. Often standardized reactor designs are required to be locally adapted, which often impacts the project plans and the supply chain. It then becomes

  2. Decision-support tool for assessing future nuclear reactor generation portfolios

    NARCIS (Netherlands)

    Jain, S.; Roelofs, F; Oosterlee, C.W.

    2014-01-01

    Capital costs, fuel, operation and maintenance (O&M) costs, and electricity prices play a key role in the economics of nuclear power plants. Often standardized reactor designs are required to be locally adapted, which often impacts the project plans and the supply chain. It then becomes difficult to

  3. Applications, progress, and the business of small, mini, and modular nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rudin, F. [Hyperion Power Generation, Santa Fe, NM (United States)

    2010-07-01

    This presentation discussed the activities of Hyperion Power Generation, a privately-owned company that is currently commercialized a small civilian nuclear reactor developed in the Los Alamos National Laboratory. The company is developing small, mini, and modular nuclear reactors ranging in cost from $75 million to $500 million. Nuclear power currently accounts for 18 percent of the total electricity produced by the United States, and large-scale nuclear power plants (NPP) typically cost between $6 billion to $9 billion. Smaller-scale nuclear plants can be used with smaller electricity grids and can be added as demand for electricity increases. The average cost per kWh for a mini-NPP is $0.04487 compared with $0.05072 for a large-scale NPP. The widespread use of smaller and modular reactors will lead to increased employment. The reactors have been designed to ensure a high level of safety and security. Issues related to training, operations, and maintenance were also reviewed. tabs., figs.

  4. Controls in new construction reactors-factory testing of the non-safety portion of the Lungmen nuclear power plant distributed control system

    Energy Technology Data Exchange (ETDEWEB)

    Wu, Y. S. [Taiwan Power Company, 242, Roosevelt Road, Sec. 3, Taipei 100, Taiwan (China); Dick, J. W. [Invensys System Inc., 33 Commercial St., Foxboro, MA 02035 (United States); Tetirick, C. W. [GE Energy, 1989 Little Orchard Street, San Jose, CA 95125-1030 (United States)

    2006-07-01

    The construction permit for Taipower's Lungmen Nuclear Units 1 and 2, two ABWR plants, was issued on March 17, 1999[1], The construction of these units is progressing actively at site. The digital I and C system supplied by GE, which is designated as the Distributed Control and Information System (DCIS) in this project, is being implemented primarily at one vendor facility. In order to ensure the reliability, safety and availability of the DCIS, it is required to comprehensively test the whole DCIS in factory. This article describes the test requirements and acceptance criteria for functional testing of the Non-Safety Distributed Control and Information system (DCIS) for Taiwan Power's Lungmen Units 1 and 2 GE selected Invensys as the equipment supplier for this Non-Safety portion of DCIS. The DCIS system of the Lungmen Units is a physically distributed control system. Field transmitters are connected to hard I/O terminal inputs on the Invensys I/A system. Once the signal is digitized on FBMs (Field Bus Modules) in Remote Multiplexing Units (RMUs), the signal is passed into an integrated control software environment. Control is based on the concept of compounds and blocks where each compound is a logical collection of blocks that performs a control function. Each point identified by control compound and block can be individually used throughout the DCIS system by referencing its unique name. In the Lungmen Project control logic and HSI (Human System Interface) requirements are divided into individual process systems called MPLs (Master Parts List). Higher-level Plant Computer System (PCS) algorithms access control compounds and blocks in these MPLs to develop functions. The test requirements and acceptance criteria for the DCIS system of the Lungmen Project are divided into three general categories (see 1,2,3 below) of verification, which in turn are divided into several specific tests: 1. DCIS System Physical Checks a) RMU Test - To confirm that the hard

  5. Selection of nuclear reactors through the hierarchic analysis process: the Mexican case; Seleccion de reactores nucleares mediante el proceso de analisis jerarquico: el caso Mexicano

    Energy Technology Data Exchange (ETDEWEB)

    Martin del Campo, C.; Nelson, P.F.; Francois, J.L. [Departamento de Sistemas Energeticos, Facultad de Ingenieria, UNAM, Paseo Cuauhnahuac 8532, Jiutepec, 62550 Morelos (Mexico)]. e-mail: cmcm@fi-b.unam.mx

    2008-07-01

    In this work the decision making method known as hierarchical analysis process for the selection of a new reactor in Mexico was applied. The main objective of the process it is to select the nuclear reactor technology more appropriate for Mexico, to begin the bid process inside one or two years to begin their operation in 2016. The options were restricted to four reactors that fulfill the following ones approaches: 1) its are advanced reactors, from the technological point of view, with regard to the reactors that at the moment operate in the Laguna Verde Power Station, 2) its are reactors that have the totally finished design, 3) its are reactors that already have the certification on the part of the regulator organism of the origin country or that they are in an advanced state of the certification process and 4) its are reactors offered by the companies that they have designed and built the greater number of reactors that are at the moment in operation at world level. Taking into account these restrictions it was decided to consider as alternative at the reactors: Advanced Boiling Water Reactor (A BWR), European Reactor of Pressurized Water (EPR), Water at Pressure reactor (AP1000) and Simplified Economic Reactor of Boiling Water (ESBWR). The evaluation approaches include economic and of safety indicators, qualitative some of them and other quantitative ones. Another grade of complexity in the solution of the problem is that there are actors that can be involved in the definition of the evaluation approaches and in the definition of the relative importance among them, according to each actor's interests. To simplify the problem its were only considered two actors or groups of interest that can influence in more significant way and that are the Federal Commission of Electricity and the National Commission of Nuclear Safety and Safeguards. The qualifications for each reactor in function of the evaluation approaches were obtained, being the A BWR the best

  6. Computer simulation of Angra-2 PWR nuclear reactor core using MCNPX code

    Energy Technology Data Exchange (ETDEWEB)

    Medeiros, Marcos P.C. de; Rebello, Wilson F., E-mail: eng.cavaliere@ime.eb.br, E-mail: rebello@ime.eb.br [Instituto Militar de Engenharia - Secao de Engenharia Nuclear, Rio de Janeiro, RJ (Brazil); Oliveira, Claudio L. [Universidade Gama Filho, Departamento de Matematica, Rio de Janeiro, RJ (Brazil); Vellozo, Sergio O., E-mail: vellozo@cbpf.br [Centro Tecnologico do Exercito. Divisao de Defesa Quimica, Biologica e Nuclear, Rio de Janeiro, RJ (Brazil); Silva, Ademir X. da, E-mail: ademir@nuclear.ufrj.br [Coordenacao dos Programas de Pos Gaduacao de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil)

    2011-07-01

    In this work the MCNPX (Monte Carlo N-Particle Transport Code) code was used to develop a computerized model of the core of Angra 2 PWR (Pressurized Water Reactor) nuclear reactor. The model was created without any kind of homogenization, but using real geometric information and material composition of that reactor, obtained from the FSAR (Final Safety Analysis Report). The model is still being improved and the version presented in this work is validated by comparing values calculated by MCNPX with results calculated by others means and presented on FSAR. This paper shows the results already obtained to K{sub eff} and K{infinity}, general parameters of the core, considering the reactor operating under stationary conditions of initial testing and operation. Other stationary operation conditions have been simulated and, in all tested cases, there was a close agreement between values calculated computationally through this model and data presented on the FSAR, which were obtained by other codes. This model is expected to become a valuable tool for many future applications. (author)

  7. Sequential probability ratio tests for reactor signal validation and sensor surveillance applications

    Energy Technology Data Exchange (ETDEWEB)

    Humenik, K. (Maryland Univ., Baltimore, MD (USA)); Gross, K.C. (Argonne National Lab., IL (USA))

    1989-11-09

    This paper examines the properties of sequential probability ratio tests (SPRT's) and the application of these tests to nuclear power reactor operation. Recently SPRT's have been applied to delayed-neutron (DN) signal data analysis using actual reactor data from the Experimental Breeder Reactor-II, which is operated by Argonne National Laboratory. The implementation of this research as part of an expert system is described. Mathematical properties of the SPRT are investigated, and theoretical results are validated with tests that use DN-signal data taken from the EBR-II in Idaho. Variations of the basic SPRT and applications to general signal validation are also explored. 16 refs., 3 figs.

  8. Nuclear reactors built, being built, or planned 1996

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-08-01

    This publication contains unclassified information about facilities, built, being built, or planned in the United States for domestic use or export as of December 31, 1996. The Office of Scientific and Technical Information, U.S. Department of Energy, gathers this information annually from Washington headquarters, and field offices of DOE; from the U.S. Nuclear Regulatory Commission (NRC); from the U. S. reactor manufacturers who are the principal nuclear contractors for foreign reactor locations; from U.S. and foreign embassies; and from foreign governmental nuclear departments. The book consists of three divisions, as follows: (1) a commercial reactor locator map and tables of the characteristic and statistical data that follow; a table of abbreviations; (2) tables of data for reactors operating, being built, or planned; and (3) tables of data for reactors that have been shut down permanently or dismantled.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-10-01

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

  10. Nuclear reactor core modelling in multifunctional simulators

    Energy Technology Data Exchange (ETDEWEB)

    Puska, E.K. [VTT Energy, Nuclear Energy, Espoo (Finland)

    1999-06-01

    The thesis concentrates on the development of nuclear reactor core models for the APROS multifunctional simulation environment and the use of the core models in various kinds of applications. The work was started in 1986 as a part of the development of the entire APROS simulation system. The aim was to create core models that would serve in a reliable manner in an interactive, modular and multifunctional simulator/plant analyser environment. One-dimensional and three-dimensional core neutronics models have been developed. Both models have two energy groups and six delayed neutron groups. The three-dimensional finite difference type core model is able to describe both BWR- and PWR-type cores with quadratic fuel assemblies and VVER-type cores with hexagonal fuel assemblies. The one- and three-dimensional core neutronics models can be connected with the homogeneous, the five-equation or the six-equation thermal hydraulic models of APROS. The key feature of APROS is that the same physical models can be used in various applications. The nuclear reactor core models of APROS have been built in such a manner that the same models can be used in simulator and plant analyser applications, as well as in safety analysis. In the APROS environment the user can select the number of flow channels in the three-dimensional reactor core and either the homogeneous, the five- or the six-equation thermal hydraulic model for these channels. The thermal hydraulic model and the number of flow channels have a decisive effect on the calculation time of the three-dimensional core model and thus, at present, these particular selections make the major difference between a safety analysis core model and a training simulator core model. The emphasis on this thesis is on the three-dimensional core model and its capability to analyse symmetric and asymmetric events in the core. The factors affecting the calculation times of various three-dimensional BWR, PWR and WWER-type APROS core models have been

  11. Design and Test of Advanced Thermal Simulators for an Alkali Metal-Cooled Reactor Simulator

    Science.gov (United States)

    Garber, Anne E.; Dickens, Ricky E.

    2011-01-01

    The Early Flight Fission Test Facility (EFF-TF) at NASA Marshall Space Flight Center (MSFC) has as one of its primary missions the development and testing of fission reactor simulators for space applications. A key component in these simulated reactors is the thermal simulator, designed to closely mimic the form and function of a nuclear fuel pin using electric heating. Continuing effort has been made to design simple, robust, inexpensive thermal simulators that closely match the steady-state and transient performance of a nuclear fuel pin. A series of these simulators have been designed, developed, fabricated and tested individually and in a number of simulated reactor systems at the EFF-TF. The purpose of the thermal simulators developed under the Fission Surface Power (FSP) task is to ensure that non-nuclear testing can be performed at sufficiently high fidelity to allow a cost-effective qualification and acceptance strategy to be used. Prototype thermal simulator design is founded on the baseline Fission Surface Power reactor design. Recent efforts have been focused on the design, fabrication and test of a prototype thermal simulator appropriate for use in the Technology Demonstration Unit (TDU). While designing the thermal simulators described in this paper, effort were made to improve the axial power profile matching of the thermal simulators. Simultaneously, a search was conducted for graphite materials with higher resistivities than had been employed in the past. The combination of these two efforts resulted in the creation of thermal simulators with power capacities of 2300-3300 W per unit. Six of these elements were installed in a simulated core and tested in the alkali metal-cooled Fission Surface Power Primary Test Circuit (FSP-PTC) at a variety of liquid metal flow rates and temperatures. This paper documents the design of the thermal simulators, test program, and test results.

  12. Thermodynamic exergy analysis for small modular reactor in nuclear hybrid energy system

    OpenAIRE

    Boldon Lauren; Sabharwall Piyush; Rabiti Cristian; Bragg-Sitton Shannon M.; Liu Li

    2016-01-01

    Small modular reactors (SMRs) provide a unique opportunity for future nuclear development with reduced financial risks, allowing the United States to meet growing energy demands through safe, reliable, clean air electricity generation while reducing greenhouse gas emissions and the reliance on unstable fossil fuel prices. A nuclear power plant is comprised of several complex subsystems which utilize materials from other subsystems and their surroundings. The economic utility of resources, or ...

  13. Reactor coolant pump testing using motor current signatures analysis

    Energy Technology Data Exchange (ETDEWEB)

    Burstein, N.; Bellamy, J.

    1996-12-01

    This paper describes reactor coolant pump motor testing carried out at Florida Power Corporation`s Crystal River plant using Framatome Technologies` new EMPATH (Electric Motor Performance Analysis and Trending Hardware) system. EMPATH{trademark} uses an improved form of Motor Current Signature Analysis (MCSA), technology, originally developed at Oak Ridge National Laboratories, for detecting deterioration in the rotors of AC induction motors. Motor Current Signature Analysis (MCSA) is a monitoring tool for motor driven equipment that provides a non-intrusive means for detecting the presence of mechanical and electrical abnormalities in the motor and the driven equipment. The base technology was developed at the Oak Ridge National Laboratory as a means for determining the affects of aging and service wear specifically on motor-operated valves used in nuclear power plant safety systems, but it is applicable to a broad range of electric machinery. MCSA is based on the recognition that an electric motor (ac or dc) driving a mechanical load acts as an efficient and permanently available transducer by sensing mechanical load variations, large and small, long-term and rapid, and converting them into variations in the induced current generated in the motor windings. The motor current variations, resulting from changes in load caused by gears, pulleys, friction, bearings, and other conditions that may change over the life of the motor, are carried by the electrical cables powering the motor and are extracted at any convenient location along the motor lead. These variations modulate the 60 Hz carrier frequency and appear as sidebands in the spectral plot.

  14. Review of nuclear data improvement needs for nuclear radiation measurement techniques used at the CEA experimental reactor facilities

    Science.gov (United States)

    Destouches, Christophe

    2016-03-01

    The constant improvement of the neutron and gamma calculation codes used in experimental nuclear reactors goes hand in hand with that of the associated nuclear data libraries. The validation of these calculation schemes always requires the confrontation with integral experiments performed in experimental reactors to be completed. Nuclear data of interest, straight as cross sections, or elaborated ones such as reactivity, are always derived from a reaction rate measurement which is the only measurable parameter in a nuclear sensor. So, in order to derive physical parameters from the electric signal of the sensor, one needs specific nuclear data libraries. This paper presents successively the main features of the measurement techniques used in the CEA experimental reactor facilities for the on-line and offline neutron/gamma flux characterizations: reactor dosimetry, neutron flux measurements with miniature fission chambers and Self Power Neutron Detector (SPND) and gamma flux measurements with chamber ionization and TLD. For each technique, the nuclear data necessary for their interpretation will be presented, the main identified needs for improvement identified and an analysis of their impact on the quality of the measurement. Finally, a synthesis of the study will be done.

  15. Review of nuclear data improvement needs for nuclear radiation measurement techniques used at the CEA experimental reactor facilities

    Directory of Open Access Journals (Sweden)

    Destouches Christophe

    2016-01-01

    Full Text Available The constant improvement of the neutron and gamma calculation codes used in experimental nuclear reactors goes hand in hand with that of the associated nuclear data libraries. The validation of these calculation schemes always requires the confrontation with integral experiments performed in experimental reactors to be completed. Nuclear data of interest, straight as cross sections, or elaborated ones such as reactivity, are always derived from a reaction rate measurement which is the only measurable parameter in a nuclear sensor. So, in order to derive physical parameters from the electric signal of the sensor, one needs specific nuclear data libraries. This paper presents successively the main features of the measurement techniques used in the CEA experimental reactor facilities for the on-line and offline neutron/gamma flux characterizations: reactor dosimetry, neutron flux measurements with miniature fission chambers and Self Power Neutron Detector (SPND and gamma flux measurements with chamber ionization and TLD. For each technique, the nuclear data necessary for their interpretation will be presented, the main identified needs for improvement identified and an analysis of their impact on the quality of the measurement. Finally, a synthesis of the study will be done.

  16. Small Fast Spectrum Reactor Designs Suitable for Direct Nuclear Thermal Propulsion

    Science.gov (United States)

    Schnitzler, Bruce G.; Borowski, Stanley K.

    2012-01-01

    Advancement of U.S. scientific, security, and economic interests through a robust space exploration program requires high performance propulsion systems to support a variety of robotic and crewed missions beyond low Earth orbit. Past studies, in particular those in support of the Space Exploration Initiative (SEI), have shown nuclear thermal propulsion systems provide superior performance for high mass high propulsive delta-V missions. The recent NASA Design Reference Architecture (DRA) 5.0 Study re-examined mission, payload, and transportation system requirements for a human Mars landing mission in the post-2030 timeframe. Nuclear thermal propulsion was again identified as the preferred in-space transportation system. A common nuclear thermal propulsion stage with three 25,000-lbf thrust engines was used for all primary mission maneuvers. Moderately lower thrust engines may also have important roles. In particular, lower thrust engine designs demonstrating the critical technologies that are directly extensible to other thrust levels are attractive from a ground testing perspective. An extensive nuclear thermal rocket technology development effort was conducted from 1955-1973 under the Rover/NERVA Program. Both graphite and refractory metal alloy fuel types were pursued. Reactors and engines employing graphite based fuels were designed, built and ground tested. A number of fast spectrum reactor and engine designs employing refractory metal alloy fuel types were proposed and designed, but none were built. The Small Nuclear Rocket Engine (SNRE) was the last engine design studied by the Los Alamos National Laboratory during the program. At the time, this engine was a state-of-the-art graphite based fuel design incorporating lessons learned from the very successful technology development program. The SNRE was a nominal 16,000-lbf thrust engine originally intended for unmanned applications with relatively short engine operations and the engine and stage design were

  17. Small Fast Spectrum Reactor Designs Suitable for Direct Nuclear Thermal Propulsion

    Energy Technology Data Exchange (ETDEWEB)

    Bruce G. Schnitzler; Stanley K. Borowski

    2012-07-01

    Advancement of U.S. scientific, security, and economic interests through a robust space exploration program requires high performance propulsion systems to support a variety of robotic and crewed missions beyond low Earth orbit. Past studies, in particular those in support of both the Strategic Defense Initiative (SDI) and Space Exploration Initiative (SEI), have shown nuclear thermal propulsion systems provide superior performance for high mass high propulsive delta-V missions. The recent NASA Design Reference Architecture (DRA) 5.0 Study re-examined mission, payload, and transportation system requirements for a human Mars landing mission in the post-2030 timeframe. Nuclear thermal propulsion was again identified as the preferred in-space transportation system. A common nuclear thermal propulsion stage with three 25,000-lbf thrust engines was used for all primary mission maneuvers. Moderately lower thrust engines may also have important roles. In particular, lower thrust engine designs demonstrating the critical technologies that are directly extensible to other thrust levels are attractive from a ground testing perspective. An extensive nuclear thermal rocket technology development effort was conducted from 1955-1973 under the Rover/NERVA Program. Both graphite and refractory metal alloy fuel types were pursued. Reactors and engines employing graphite based fuels were designed, built and ground tested. A number of fast spectrum reactor and engine designs employing refractory metal alloy fuel types were proposed and designed, but none were built. The Small Nuclear Rocket Engine (SNRE) was the last engine design studied by the Los Alamos National Laboratory during the program. At the time, this engine was a state-of-the-art graphite based fuel design incorporating lessons learned from the very successful technology development program. The SNRE was a nominal 16,000-lbf thrust engine originally intended for unmanned applications with relatively short engine

  18. Solid0Core Heat-Pipe Nuclear Batterly Type Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ehud Greenspan

    2008-09-30

    This project was devoted to a preliminary assessment of the feasibility of designing an Encapsulated Nuclear Heat Source (ENHS) reactor to have a solid core from which heat is removed by liquid-metal heat pipes (HP).

  19. Nuclear reactors built, being built, or planned, 1994

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-07-01

    This document contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of December 31, 1994. The Office of Scientific and Technical Information, US Department of Energy, gathers this information annually from Washington headquarters and field offices of DOE; from the US Nuclear Regulatory Commission (NRC); from the US reactor manufacturers who are the principal nuclear contractors for foreign reactor locations; from US and foreign embassies; and from foreign governmental nuclear departments. The book consists of three divisions, as follows: a commercial reactor locator map and tables of the characteristic and statistical data that follow; a table of abbreviations; tables of data for reactors operating, being built, or planned; and tables of data for reactors that have been shut down permanently or dismantled. The reactors are subdivided into the following parts: Civilian, Production, Military, Export, and Critical Assembly. Export reactor refers to a reactor for which the principal nuclear contractor is a US company -- working either independently or in cooperation with a foreign company (Part 4). Critical assembly refers to an assembly of fuel and moderator that requires an external source of neutrons to initiate and maintain fission. A critical assembly is used for experimental measurements (Part 5).

  20. Nuclear reactors built, being built, or planned: 1995

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-08-01

    This report contains unclassified information about facilities built, being built, or planned in the US for domestic use or export as of December 31, 1995. The Office of Scientific and Technical Information, US Department of Energy, gathers this information annually from Washington headquarters and field offices of DOE; from the US Nuclear Regulatory Commission (NRC); from the US reactor manufacturers who are the principal nuclear contractors for foreign reactor locations; from US and foreign embassies; and from foreign governmental nuclear departments. The book consists of three divisions, as follows: (1) a commercial reactor locator map and tables of the characteristic and statistical data that follow; a table of abbreviations; (2) tables of data for reactors operating, being built, or planned; and (3) tables of data for reactors that have been shut down permanently or dismantled. The reactors are subdivided into the following parts: Civilian, Production, Military, Export, and Critical Assembly. Export reactor refers to a reactor for which the principal nuclear contractor is a US company--working either independently or in cooperation with a foreign company (Part 4). Critical assembly refers to an assembly of fuel and moderator that requires an external source of neutrons to initiate and maintain fission. A critical assembly is used for experimental measurements (Part 5).

  1. External fuel thermionic reactor system.

    Science.gov (United States)

    Mondt, J. F.; Peelgren, M. L.

    1971-01-01

    Thermionic reactors are prime candidates for nuclear electric propulsion. The national thermionic reactor effort is concentrated on the flashlight concept with the external-fuel concept as the backup. The external-fuel concept is very adaptable to a completely modular power subsystem which is attractive for highly reliable long-life applications. The 20- to 25-cm long, externally-fueled converters have been designed, fabricated, and successfully tested with many thermal cycles by electrical heating. However, difficulties have been encountered during encapsulation for nuclear heated tests and none have been started to date. These nuclear tests are required to demonstrate the concept feasibility.

  2. Grey Rod Test in HANARO Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Choo, K. N.; Kim, B. G.; Kang, Y. H. (and others)

    2008-08-15

    Westinghouse/KAERI/KNF agreed to perform an irradiation test in the HANARO reactor to obtain irradiation data on the new grey rods that will be part of an AP1000 system. As a preliminary test, two samples containing pure Ag (Reference) and Ag-In-Cd materials provided by Westinghouse Electric Company (WEC) were inserted in a KNF irradiation capsule of 07M-13N. The specimens were irradiated for 95.19days (4 cycles) in the CT test hole of the HANARO of a 30MW thermal output to have a fast neutron fluence of 1.11x10{sup 21}(n/cm{sup 2}) (E>1.0MeV). This report provides all the test conditions and data obtained during the irradiation test of the grey rods in HANARO requested by Westinghouse. The test was prepared according to the meeting minutes (June 26, 2007) and the on-going subject test was stopped midway by the request of Westinghouse.

  3. Test Results From a Direct Drive Gas Reactor Simulator Coupled to a Brayton Power Conversion Unit

    Science.gov (United States)

    Hervol, David S.; Briggs, Maxwell H.; Owen, Albert K.; Bragg-Sitton, Shannon M.

    2009-01-01

    The Brayton Power Conversion Unit (BPCU) located at NASA Glenn Research Center (GRC) in Cleveland, OH is a closed cycle system incorporating a turboaltemator, recuperator, and gas cooler connected by gas ducts to an external gas heater. For this series of tests, the BPCU was modified by replacing the gas heater with the Direct Drive Gas heater or DOG. The DOG uses electric resistance heaters to simulate a fast spectrum nuclear reactor similar to those proposed for space power applications. The combined system thermal transient behavior was the focus of these tests. The BPCU was operated at various steady state points. At each point it was subjected to transient changes involving shaft rotational speed or DOG electrical input. This paper outlines the changes made to the test unit and describes the testing that took place along with the test results.

  4. Supplying the nuclear arsenal: Production reactor technology, management, and policy, 1942--1992

    Energy Technology Data Exchange (ETDEWEB)

    Carlisle, R.P.; Zenzen, J.M.

    1994-01-01

    This book focuses on the lineage of America`s production reactors, those three at Hanford and their descendants, the reactors behind America`s nuclear weapons. The work will take only occasional sideways glances at the collateral lines of descent, the reactor cousins designed for experimental purposes, ship propulsion, and electric power generation. Over the decades from 1942 through 1992, fourteen American production reactors made enough plutonium to fuel a formidable arsenal of more than twenty thousand weapons. In the last years of that period, planners, nuclear engineers, and managers struggled over designs for the next generation of production reactors. The story of fourteen individual machines and of the planning effort to replace them might appear relatively narrow. Yet these machines lay at the heart of the nation`s nuclear weapons complex. The story of these machines is the story of arming the winning weapon, supplying the nuclear arms race. This book is intended to capture the history of the first fourteen production reactors, and associated design work, in the face of the end of the Cold War.

  5. High Fidelity, Fuel-Like Thermal Simulators for Non-Nuclear Testing: Analysis and Initial Test Results

    Science.gov (United States)

    Bragg-Sitton, Shannon M.; Dickens, Ricky; Dixon, David; Kapernick, Richard

    2007-01-01

    Non-nuclear testing can be a valuable tool in the development of a space nuclear power system, providing system characterization data and allowing one to work through various fabrication, assembly and integration issues without the cost and time associated with a full ground nuclear test. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Testing with non-optimized heater elements allows one to assess thermal, heat transfer. and stress related attributes of a given system, but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. High fidelity thermal simulators that match both the static and the dynamic fuel pin performance that would be observed in an operating, fueled nuclear reactor can vastly increase the value of non-nuclear test results. With optimized simulators, the integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and fueled nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response characteristics and assess potential design improvements at relatively small fiscal investment. Initial conceptual thermal simulator designs are determined by simple one-dimensional analysis at a single axial location and at steady state conditions; feasible concepts are then input into a detailed three-dimensional model for comparison to expected fuel pin performance. Static and dynamic fuel pin performance for a proposed reactor design is determined using SINDA/FLUINT thermal analysis software, and comparison is made between the expected nuclear performance and the performance of conceptual thermal simulator designs. Through a series of iterative analyses, a conceptual high fidelity design is developed

  6. Flow-induced vibration for light water reactors. Final progress report, July 1981-September 1981

    Energy Technology Data Exchange (ETDEWEB)

    Torres, M.R.

    1981-11-01

    Flow-Induced Vibration for Light Water Reactors (FIV for LWRs) is a program designed to improve the FIV performance of light water reactors through the development of design criteria, analytical models for predicting behavior of components, and general scaling laws to improve the accuracy of reduced-scale tests, and through the identification of high FIV risk areas. The program is managed by the General Electric Nuclear Power Systems Engineering Department and has three major contributors: General Electric Nuclear Power Systems Engineering Department (NPSED), General Electric Corporate Research and Development (CR and D) and Argonne National Laboratory (ANL). The program commenced December 1, 1976. This progress report summarizes the accomplishments achieved during the final period from July 1981 to September 1981. This is the last quarterly progress report to be issued for this program.

  7. Testing commercial catalysts in recycle reactors

    Energy Technology Data Exchange (ETDEWEB)

    Berty, J.M.

    1979-01-01

    Recycle reactors for quality control of catalyst production and for testing new catalysts for known or new processes have the following advantages over tubular reactors: they can reproduce the physical and chemical regime which surrounds the catalyst in a commercial reactor; they can achieve high mass and heat transfer; they exhibit uniform coke deposit; and they provide independence of mass velocity and space velocity. Their disadvantage is the unconventional specification of experiments in terms of discharge concentration which derives from the implicit nature of the basic mathematical relationships. Recycle reactor test methods are outlined for quality control and for testing catalysts, e.g., supported nickel from different manufacturers, for processes whose chemistry is well known. Approaches for testing catalysts for new processes are discussed. The standard recycle reactor developed at Union Carbide Corp. and manufactured by Autoclave Engineers, and several of its modifications are described.

  8. Characterization of nuclear reactor containment penetrations. Preliminary report

    Energy Technology Data Exchange (ETDEWEB)

    Bump, T.R.; Seidensticker, R.W.; Shackelford, M.A.; Gambhir, V.K.; McLennan, G.L.

    1984-06-01

    This report summarizes the survey work conducted by Argonne National Laboratory on the design and details of major penetrations in 22 nuclear power plants. The survey includes all containment types and materials in current use. It also includes details of all types of penetrations (except for electrical penetration assemblies and valves) and the seals and gaskets used in them. The report provides a test matrix for testing major penetrations and for testing seals and gaskets in order to evaluate their leakage potential under severe accident conditions.

  9. An introduction to the engineering of fast nuclear reactors

    CERN Document Server

    Judd, Anthony M

    2014-01-01

    An invaluable resource for both graduate-level engineering students and practising nuclear engineers who want to expand their knowledge of fast nuclear reactors, the reactors of the future! This book is a concise yet comprehensive introduction to all aspects of fast reactor engineering. It covers topics including neutron physics; neutron flux spectra; flux distribution; Doppler and coolant temperature coefficients; the performance of ceramic and metal fuels under irradiation, structural changes, and fission-product migration; the effects of irradiation and corrosion on structural materials, irradiation swelling; heat transfer in the reactor core and its effect on core design; coolants including sodium and lead-bismuth alloy; coolant circuits; pumps; heat exchangers and steam generators; and plant control. The book includes new discussions on lead-alloy and gas coolants, metal fuel, the use of reactors to consume radioactive waste, and accelerator-driven subcritical systems.

  10. Physics of nuclear reactors; La physique des reacteurs nucleaires

    Energy Technology Data Exchange (ETDEWEB)

    Marguet, S. [Ecole Nationale Superieure de Risques Industriels de Bourges, 18 (France); Institut de Transfert de Technologie d' EDF, 92 - Clamart (France)

    2011-07-01

    This manual covers all the aspects of the science of neutron transport in nuclear reactors and can be used with great advantage by students, engineers or even reactor experts. It is composed of 18 chapters: 1) basis of nuclear physics, 2) the interactions of neutrons with matter, 3) the interactions of electromagnetic radiations and charged-particles with matter, 4) neutron slowing-down, 5) resonant absorption, 6) Doppler effect, 7) neutron thermalization, 8) Boltzmann equation, 9) calculation methods in neutron transport theory, 10) neutron scattering, 11) reactor reactivity, 12) theory of the critical homogenous pile, 13) the neutron reflector, 14) the heterogeneous reactor, 15) the equations of the fuel cycle, 16) neutron counter-reactions, 17) reactor kinetics, and 18) calculation methods in neutron scattering

  11. Nuclear data and reactor physics activities in Indonesia

    Energy Technology Data Exchange (ETDEWEB)

    Liem, P.H. [National Atomic Energy Agency, Tangerang (Indonesia). Center for Multipurpose Reactor

    1998-03-01

    The nuclear data and reactor physics activities in Indonesia, especially, in the National Atomic Energy Agency are presented. In the nuclear data field, the Agency is now taking the position of a user of the main nuclear data libraries such as JENDL and ENDF/B. These nuclear data libraries become the main sources for producing problem dependent cross section sets that are needed by cell calculation codes or transport codes for design, analysis and safety evaluation of research reactors. In the reactor physics field, besides utilising the existing core analysis codes obtained from bilateral and international co-operation, the Agency is putting much effort to self-develop Batan`s codes for reactor physics calculations, in particular, for research reactor and high temperature reactor design, analysis and fuel management. Under the collaboration with JAERI, Monte Carlo criticality calculations on the first criticality of RSG GAS (MPR-30) first core were done using JAERI continuous energy, vectorized Monte Carlo code, MVP, with JENDL-3.1 and JENDL-3.2 nuclear data libraries. The results were then compared with the experiment data collected during the commissioning phase. Monte Carlo calculations with both JENDL-3.1 and -3.2 libraries produced k{sub eff} values with excellent agreement with experiment data, however, systematically, JENDL-3.2 library showed slightly higher k{sub eff} values than JENDL-3.1 library. (author)

  12. Nuclear Hybrid Energy System Model Stability Testing

    Energy Technology Data Exchange (ETDEWEB)

    Greenwood, Michael Scott [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Cetiner, Sacit M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Fugate, David W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-04-01

    A Nuclear Hybrid Energy System (NHES) uses a nuclear reactor as the basic power generation unit, and the power generated is used by multiple customers as combinations of thermal power or electrical power. The definition and architecture of a particular NHES can be adapted based on the needs and opportunities of different localities and markets. For example, locations in need of potable water may be best served by coupling a desalination plant to the NHES. Similarly, a location near oil refineries may have a need for emission-free hydrogen production. Using the flexible, multi-domain capabilities of Modelica, Argonne National Laboratory, Idaho National Laboratory, and Oak Ridge National Laboratory are investigating the dynamics (e.g., thermal hydraulics and electrical generation/consumption) and cost of a hybrid system. This paper examines the NHES work underway, emphasizing the control system developed for individual subsystems and the overall supervisory control system.

  13. A novel scheme for making cheap electricity with nuclear energy

    Energy Technology Data Exchange (ETDEWEB)

    Pettibone, J.S.

    1991-04-01

    Nuclear fuels should produce cheaper electricity than coal, considering their high specific energy and low cost. To exploit these properties, the scheme proposed here replaces the expensive reactor/steam-turbine system with an engine in which the expansion of a gas heated by a nuclear explosion raises a mass of liquid, thereby producing stored hydraulic energy. This energy could be converted to electricity by hydroelectric generation with water as the working fluid or by magnetohydrodynamic (MHD) generation with molten metal. A rough cost analysis suggests the hydroelectric system could reduce the present cost of electricity by two-thirds, and the MHD system by even more. Such cheap power would make feasible large-scale electrolysis to produce hydrogen and other fuels and chemical raw materials. 2 refs., 1 fig.

  14. Novel scheme for making cheap electricity with nuclear energy

    Energy Technology Data Exchange (ETDEWEB)

    Pettibone, J.A.

    1979-08-24

    Nuclear fuels should produce cheaper electricity than coal, considering their high specific energy and low cost. To exploit these properties, the scheme proposed here replaces the expensive reactor/steam-turbine system with an engine in which the expansion of a gas heated by a nuclear explosion raises a mass of liquid, thereby producing stored hydraulic energy. This energy could be converted to electricity by hydroelectric generation with water as the working fluid or by magnetohydrodynamic (MHD) generation with molten metal. A rough cost analysis suggests the hydroelectric system could reduce the present cost of electricity by two-thirds, and the MHD system by even more. Such cheap power would make feasible large-scale electrolysis to produce hydrogen and other fuels and chemical raw materials.

  15. Nuclear Technology Series. Course 8: Reactor Safety.

    Science.gov (United States)

    Center for Occupational Research and Development, Inc., Waco, TX.

    This technical specialty course is one of thirty-five courses designed for use by two-year postsecondary institutians in five nuclear technician curriculum areas: (1) radiation protection technician, (2) nuclear instrumentation and control technician, (3) nuclear materials processing technician, (4) nuclear quality-assurance/quality-control…

  16. Nuclear Technology Series. Course 12: Reactor Physics.

    Science.gov (United States)

    Center for Occupational Research and Development, Inc., Waco, TX.

    This technical specialty course is one of thirty-five courses designed for use by two-year postsecondary institutions in five nuclear technician curriculum areas: (1) radiation protection technician, (2) nuclear instrumentation and control technician, (3) nuclear materials processing technician, (4) nuclear quality-assurance/quality-control…

  17. Radioisotope Electric Propulsion (REP): A Near-Term Approach to Nuclear Propulsion

    Science.gov (United States)

    Schmidt, George R.; Manzella, David H.; Kamhawi, Hani; Kremic, Tibor; Oleson, Steven R.; Dankanich, John W.; Dudzinski, Leonard A.

    2009-01-01

    Studies over the last decade have shown radioisotope-based nuclear electric propulsion to be enhancing and, in some cases, enabling for many potential robotic science missions. Also known as radioisotope electric propulsion (REP), the technology offers the performance advantages of traditional reactor-powered electric propulsion (i.e., high specific impulse propulsion at large distances from the Sun), but with much smaller, affordable spacecraft. Future use of REP requires development of radioisotope power sources with system specific powers well above that of current systems. The US Department of Energy and NASA have developed an advanced Stirling radioisotope generator (ASRG) engineering unit, which was subjected to rigorous flight qualification-level tests in 2008, and began extended lifetime testing later that year. This advancement, along with recent work on small ion thrusters and life extension technology for Hall thrusters, could enable missions using REP sometime during the next decade.

  18. Monitoring Akkuyu Nuclear Reactor Using Anti-Neutrino Flux Measurement

    CERN Document Server

    Ozturk, Sertac; Ozcan, V Erkcan; Unel, Gokhan

    2016-01-01

    We present a simulation based study for monitoring Akkuyu Nuclear Power Plant's activity using anti-neutrino flux originating from the reactor core. A water Cherenkov detector has been designed and optimization studies have been performed using Geant4 simulation toolkit. A first study for the design of a monitoring detector facility for Akkuyu Nuclear Power Plant has been discussed in this paper.

  19. Economics and utilization of thorium in nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    1978-05-01

    Information on thorium utilization in power reactors is presented concerning the potential demand for nuclear power, the potential supply for nuclear power, economic performance of thorium under different recycle policies, ease of commercialization of the economically preferred cases, policy options to overcome institutional barriers, and policy options to overcome technological and regulatory barriers.

  20. Fuel element concept for long life high power nuclear reactors

    Science.gov (United States)

    Mcdonald, G. E.; Rom, F. E.

    1969-01-01

    Nuclear reactor fuel elements have burnups that are an order of magnitude higher than can currently be achieved by conventional design practice. Elements have greater time integrated power producing capacity per unit volume. Element design concept capitalizes on known design principles and observed behavior of nuclear fuel.

  1. 77 FR 3009 - Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors

    Science.gov (United States)

    2012-01-20

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors..., ``Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors.''...

  2. 76 FR 14437 - Economic Simplified Boiling Water Reactor Standard Design: GE Hitachi Nuclear Energy; Issuance of...

    Science.gov (United States)

    2011-03-16

    ... From the Federal Register Online via the Government Publishing Office ] NUCLEAR REGULATORY COMMISSION Economic Simplified Boiling Water Reactor Standard Design: GE Hitachi Nuclear Energy; Issuance of... GE Hitachi Nuclear Energy (GEH) for the economic simplified boiling water reactor (ESBWR) standard...

  3. A Spouted Bed Reactor Monitoring System for Particulate Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    D. S. Wendt; R. L. Bewley; W. E. Windes

    2007-06-01

    Conversion and coating of particle nuclear fuel is performed in spouted (fluidized) bed reactors. The reactor must be capable of operating at temperatures up to 2000°C in inert, flammable, and coating gas environments. The spouted bed reactor geometry is defined by a graphite retort with a 2.5 inch inside diameter, conical section with a 60° included angle, and a 4 mm gas inlet orifice diameter through which particles are removed from the reactor at the completion of each run. The particles may range from 200 µm to 2 mm in diameter. Maintaining optimal gas flow rates slightly above the minimum spouting velocity throughout the duration of each run is complicated by the variation of particle size and density as conversion and/or coating reactions proceed in addition to gas composition and temperature variations. In order to achieve uniform particle coating, prevent agglomeration of the particle bed, and monitor the reaction progress, a spouted bed monitoring system was developed. The monitoring system includes a high-sensitivity, low-response time differential pressure transducer paired with a signal processing, data acquisition, and process control unit which allows for real-time monitoring and control of the spouted bed reactor. The pressure transducer is mounted upstream of the spouted bed reactor gas inlet. The gas flow into the reactor induces motion of the particles in the bed and prevents the particles from draining from the reactor due to gravitational forces. Pressure fluctuations in the gas inlet stream are generated as the particles in the bed interact with the entering gas stream. The pressure fluctuations are produced by bulk movement of the bed, generation and movement of gas bubbles through the bed, and the individual motion of particles and particle subsets in the bed. The pressure fluctuations propagate upstream to the pressure transducer where they can be monitored. Pressure fluctuation, mean differential pressure, gas flow rate, reactor

  4. Modular Nuclear Testing Concept

    Energy Technology Data Exchange (ETDEWEB)

    Wouters, L. F.

    1964-07-01

    The continuing concern with efficient utilization of manpower at the Nevada Test Site (NTS) and the seemingly high cost of individual nuclear shots, together with the recent evolution of the L-12 scope, generated some fresh thoughts concerning more efficient procedures for the `average` test. Every time anyone looks at the problem., they tend to analyze the existing conventional approach and try to find the one or two most expensive and `lossy` factors in the can of worms. Usually this turns out to be a problem within the realm of specialization of the particular analyst! People not so directly concerned with the program tend to look for, or wish for, or even `invent` miracles`. Our present techniques appear to be the sum (and possibly even the product) of many small contributions which have all been beaten down to the same level of importance. Such a situation in any systemic problem is usually symptomatic of the need for fairly violent departures in the aver-all system approach, at least in thinking. This report proposes and details a modular nuclear testing concept.

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

    Energy Technology Data Exchange (ETDEWEB)

    Wittenbrock, N. G.

    1982-01-01

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

  6. Designing a mini subcritical nuclear reactor; Diseno de un mini reactor nuclear subcritico

    Energy Technology Data Exchange (ETDEWEB)

    Escobedo G, C. R.; Vega C, H. R.; Davila H, V. M., E-mail: rafelaescobedo@hotmail.com [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Jardin Juarez 147, Col. Centro, 98000 Zacatecas, Zac. (Mexico)

    2015-10-15

    In this work the design of a mini subcritical nuclear reactor formed by means of light water moderator, uranium as fuel, and isotopic neutron source of {sup 239}PuBe was carried out. The design was done by Monte Carlo methods with the code MCNP5 in which uranium was modeled in an array of concentric holes cylinders of 8.5, 14.5, 20.5, 26.5, 32.5 cm of internal radius and 3 cm of thickness, 36 cm of height. Different models were made from a single fuel cylinder (natural uranium) to five. The neutron source of {sup 239}PuBe was situated in the center of the mini reactor; in each arrangement was used water as moderator. Cross sections libraries Endf/Vi were used and the number of stories was large enough to ensure less uncertainty than 3%. For each case the effective multiplication factor k{sub e}-f{sub f}, the amplification factor and the power was calculated. Outside the mini reactor the ambient dose equivalent H (10) was calculated for different cases. The value of k{sub eff}, the amplification factor and power are directly related to the number of cylinders of uranium as fuel. Although the average energy of the neutrons {sup 239}PuBe is between 4.5 and 5 MeV in the case of the mini reactor for a cylinder, in the neutron spectrum the presence of thermal neutrons does not exist, so that produced fissions are generated with fast neutrons, and in designs of two and three rings the neutron spectra shows the presence of thermal neutrons, however the fissions are being generated with fast neutrons. Finally in the four and five cases the amount of moderator is enough to thermalized the neutrons and thereby produce the fission. The maximum value for k{sub eff} was 0.82; this value is very close to the assembly of Universidad Autonoma de Zacatecas generating a k{sub eff} of 0.86. According to the safety and radiation protection standards for the design of mini reactor of one, two and three cylinders they comply with the established safety, while designs of four and five

  7. Spent nuclear fuel discharges from US reactors 1993

    Energy Technology Data Exchange (ETDEWEB)

    1995-02-01

    The Energy Information Administration (EIA) of the U.S. Department of Energy (DOE) administers the Nuclear Fuel Data Survey, Form RW-859. This form is used to collect data on fuel assemblies irradiated at commercial nuclear reactors operating in the United States, and the current inventories and storage capacities of those reactors. These data are important to the design and operation of the equipment and facilities that DOE will use for the future acceptance, transportation, and disposal of spent fuels. The data collected and presented identifies trends in burnup, enrichment, and spent nuclear fuel discharged form commercial light-water reactor as of December 31, 1993. The document covers not only spent nuclear fuel discharges; but also site capacities and inventories; canisters and nonfuel components; and assembly type characteristics.

  8. SNIF: A Futuristic Neutrino Probe for Undeclared Nuclear Fission Reactors

    CERN Document Server

    Lasserre, Thierry; Mention, Guillaume; Reboulleau, Romain; Cribier, Michel; Letourneau, Alain; Lhuillier, David

    2010-01-01

    Today reactor neutrino experiments are at the cutting edge of fundamental research in particle physics. Understanding the neutrino is far from complete, but thanks to the impressive progress in this field over the last 15 years, a few research groups are seriously considering that neutrinos could be useful for society. The International Atomic Energy Agency (IAEA) works with its Member States to promote safe, secure and peaceful nuclear technologies. In a context of international tension and nuclear renaissance, neutrino detectors could help IAEA to enforce the Treaty on the Non-Proliferation of Nuclear Weapons (NPT). In this article we discuss a futuristic neutrino application to detect and localize an undeclared nuclear reactor from across borders. The SNIF (Secret Neutrino Interactions Finder) concept proposes to use a few hundred thousand tons neutrino detectors to unveil clandestine fission reactors. Beyond previous studies we provide estimates of all known background sources as a function of the detecto...

  9. Radioactive fallout from Chinese nuclear weapons test

    Energy Technology Data Exchange (ETDEWEB)

    Thomas, C.W.; Soldat, J.K.; Silker, W.B.; Perkins, R.W.

    1976-09-26

    Radioactive fallout from this Chinese nuclear test resulted in measurable deposition of short-lived debris over much of the United States. The fallout levels varied by more than 1000-fold and showed significant temporary or spatial fractionation with higher levels of deposition being associated with rain. The particle size with which the airborne debris was associated decreased continuously with time following detonation and a substantial fraction of the {sup 131}I was associated with inorganic and organic gases. The potential radiation dose to an infant consuming milk produced at the location of the highest concentration of {sup 131}I measured on grass was estimated to be {approximately}l rem. This dose is about 50 times the annual dose received in the vicinity of a power reactor operating under the existing US Nuclear Regulatory Commission design guides. The potential upper limit thyroid dose for the population of 17 eastern seaboard states from this single test was estimated to be about 2.4 {times} 10{sup 6} man-thyroid-rem under the assumption that all dairy cows remained on fresh pasture throughout the month following the initial decomposition of fallout debris. This dose is about 200 times the estimated dose from currently operating nuclear power reactors and about 50 times the annual US population thyroid dose that would be received from 500 GWe of nuclear power reactors in the year 2000.

  10. General Relativity in Electrical Engineering

    OpenAIRE

    Leonhardt, Ulf; Philbin, Thomas G.

    2006-01-01

    In electrical engineering metamaterials have been developed that offer unprecedented control over electromagnetic fields. Here we show that general relativity lends the theoretical tools for designing devices made of such versatile materials. Given a desired device function, the theory describes the electromagnetic properties that turn this function into fact. We consider media that facilitate space-time transformations and include negative refraction. Our theory unifies the concepts operatin...

  11. Nuclear Electric Propulsion for Outer Space Missions

    Science.gov (United States)

    Barret, Chris

    2003-01-01

    Today we know of 66 moons in our very own Solar System, and many of these have atmospheres and oceans. In addition, the Hubble (optical) Space Telescope has helped us to discover a total of 100 extra-solar planets, i.e., planets going around other suns, including several solar systems. The Chandra (X-ray) Space Telescope has helped us to discover 33 Black Holes. There are some extremely fascinating things out there in our Universe to explore. In order to travel greater distances into our Universe, and to reach planetary bodies in our Solar System in much less time, new and innovative space propulsion systems must be developed. To this end NASA has created the Prometheus Program. When one considers space missions to the outer edges of our Solar System and far beyond, our Sun cannot be relied on to produce the required spacecraft (s/c) power. Solar energy diminishes as the square of the distance from the Sun. At Mars it is only 43% of that at Earth. At Jupiter, it falls off to only 3.6% of Earth's. By the time we get out to Pluto, solar energy is only .066% what it is on Earth. Therefore, beyond the orbit of Mars, it is not practical to depend on solar power for a s/c. However, the farther out we go the more power we need to heat the s/c and to transmit data back to Earth over the long distances. On Earth, knowledge is power. In the outer Solar System, power is knowledge. It is important that the public be made aware of the tremendous space benefits offered by Nuclear Electric Propulsion (NEP) and the minimal risk it poses to our environment. This paper presents an overview of the reasons for NEP systems, along with their basic components including the reactor, power conversion units (both static and dynamic), electric thrusters, and the launch safety of the NEP system.

  12. Nuclear spin squeezing via electric quadrupole interaction

    Science.gov (United States)

    Aksu Korkmaz, Yaǧmur; Bulutay, Ceyhun

    2016-01-01

    Control over nuclear-spin fluctuations is essential for processes that rely on preserving the quantum state of an embedded system. For this purpose, squeezing is a viable alternative, so far that has not been properly exploited for the nuclear spins. Of particular relevance in solids is the electric quadrupole interaction (QI), which operates on nuclei having spin higher than 1/2. In its general form, QI involves an electric-field gradient (EFG) biaxiality term. Here, we show that as this EFG biaxiality increases, it enables continuous tuning of single-particle squeezing from the one-axis twisting to the two-axis countertwisting limits. A detailed analysis of QI squeezing is provided, exhibiting the intricate consequences of EFG biaxiality. The initial states over the Bloch sphere are mapped out to identify those favorable for fast initial squeezing, or for prolonged squeezings. Furthermore, the evolution of squeezing in the presence of a phase-damping channel and an external magnetic field are investigated. We observe that dephasing drives toward an antisqueezed terminal state, the degree of which increases with the spin angular momentum. Finally, QI squeezing in the limiting case of a two-dimensional EFG with a perpendicular magnetic field is discussed, which is of importance for two-dimensional materials, and the associated beat patterns in squeezing are revealed.

  13. Fuel supply of nuclear power industry with the introduction of fast reactors

    Science.gov (United States)

    Muraviev, E. V.

    2014-12-01

    The results of studies conducted for the validation of the updated development strategy for nuclear power industry in Russia in the 21st century are presented. Scenarios with different options for the reprocessing of spent fuel of thermal reactors and large-scale growth of nuclear power industry based on fast reactors of inherent safety with a breeding ratio of ˜1 in a closed nuclear fuel cycle are considered. The possibility of enhanced fuel breeding in fast reactors is also taken into account in the analysis. The potential to establish a large-scale nuclear power industry that covers 100% of the increase in electric power requirements in Russia is demonstrated. This power industry may be built by the end of the century through the introduction of fast reactors (replacing thermal ones) with a gross uranium consumption of up to ˜1 million t and the termination of uranium mining even if the reprocessing of spent fuel of thermal reactors is stopped or suffers a long-term delay.

  14. Radionuclide inventories for short run-time space nuclear reactor systems

    Science.gov (United States)

    Coats, Richard L.

    1993-01-01

    Space Nuclear Reactor Systems, especially those used for propulsion, often have expected operation run times much shorter than those for land-based nuclear power plants. This produces substantially different radionuclide inventories to be considered in the safety analyses of space nuclear systems. This presentation describes an analysis utilizing ORIGEN2 and DKPOWER to provide comparisons among representative land-based and space systems. These comparisons enable early, conceptual considerations of safety issues and features in the preliminary design phases of operational systems, test facilities, and operations by identifying differences between the requirements for space systems and the established practice for land-based power systems. Early indications are that separation distance is much more effective as a safety measure for space nuclear systems than for power reactors because greater decay of the radionuclide activity occurs during the time to transport the inventory a given distance. In addition, the inventories of long-lived actinides are very low for space reactor systems.

  15. Advanced Test Reactor National Scientific User Facility

    Energy Technology Data Exchange (ETDEWEB)

    Frances M. Marshall; Jeff Benson; Mary Catherine Thelen

    2011-08-01

    The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is a large test reactor for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The ATR is a pressurized, light-water, high flux test reactor with a maximum operating power of 250 MWth. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material irradiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research by a broader user community. This paper highlights the ATR NSUF research program and the associated educational initiatives.

  16. Dynamic model of Fast Breeder Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Vaidyanathan, G., E-mail: vaidya@igcar.gov.i [Fast Reactor Technology Group, Indira Gandhi Center for Atomic Research, Kalpakkam (India); Kasinathan, N.; Velusamy, K. [Fast Reactor Technology Group, Indira Gandhi Center for Atomic Research, Kalpakkam (India)

    2010-04-15

    Fast Breeder Test Reactor (FBTR) is a 40 M Wt/13.2 MWe sodium cooled reactor operating since 1985. It is a loop type reactor. As part of the safety analysis the response of the plant to various transients is needed. In this connection a computer code named DYNAM was developed to model the reactor core, the intermediate heat exchanger, steam generator, piping, etc. This paper deals with the mathematical model of the various components of FBTR, the numerical techniques to solve the model, and comparison of the predictions of the code with plant measurements. Also presented is the benign response of the plant to a station blackout condition, which brings out the role of the various reactivity feedback mechanisms combined with a gradual coast down of reactor sodium flow.

  17. Heterogeneous Nuclear Reactor Models for Optimal Xenon Control.

    Science.gov (United States)

    Gondal, Ishtiaq Ahmad

    Nuclear reactors are generally modeled as homogeneous mixtures of fuel, control, and other materials while in reality they are heterogeneous-homogeneous configurations comprised of fuel and control rods along with other materials. Similarly, for space-time studies of a nuclear reactor, homogeneous, usually one-group diffusion theory, models are used, and the system equations are solved by either nodal or modal expansion approximations. Study of xenon-induced problems has also been carried out using similar models and with the help of dynamic programming or classical calculus of variations or the minimum principle. In this study a thermal nuclear reactor is modeled as a two-dimensional lattice of fuel and control rods placed in an infinite-moderator in plane geometry. The two-group diffusion theory approximation is used for neutron transport. Space -time neutron balance equations are written for two groups and reduced to one space-time algebraic equation by using the two-dimensional Fourier transform. This equation is written at all fuel and control rod locations. Iodine -xenon and promethium-samarium dynamic equations are also written at fuel rod locations only. These equations are then linearized about an equilibrium point which is determined from the steady-state form of the original nonlinear system equations. After studying poisonless criticality, with and without control, and the stability of the open-loop system and after checking its controllability, a performance criterion is defined for the xenon-induced spatial flux oscillation problem in the form of a functional to be minimized. Linear -quadratic optimal control theory is then applied to solve the problem. To perform a variety of different additional useful studies, this formulation has potential for various extensions and variations; for example, different geometry of the problem, with possible extension to three dimensions, heterogeneous -homogeneous formulation to include, for example, homogeneously

  18. Linear stability analysis of a nuclear reactor using the lumped model

    Directory of Open Access Journals (Sweden)

    Kale Vivek A.

    2016-01-01

    Full Text Available The stability analysis of a nuclear reactor is an important aspect in the design and operation of the reactor. A stable neutronic response to perturbations is essential from the safety point of view. In this paper, a general methodology has been developed for the linear stability analysis of nuclear reactors using the lumped reactor model. The reactor kinetics has been modelled using the point kinetics equations and the reactivity feedbacks from fuel, coolant and xenon have been modelled through the appropriate time dependent equations. These governing equations are linearized considering small perturbations in the reactor state around a steady operating point. The characteristic equation of the system is used to establish the stability zone of the reactor considering the reactivity coefficients as parameters. This methodology has been used to identify the stability region of a typical pressurized heavy water reactor. It is shown that the positive reactivity feedback from xenon narrows down the stability region. Further, it is observed that the neutron kinetics parameters (such as the number of delayed neutron precursor groups considered, the neutron generation time, the delayed neutron fractions, etc. do not have a significant influence on the location of the stability boundary. The stability boundary is largely influenced by the parameters governing the evolution of the fuel and coolant temperature and xenon concentration.

  19. Argonne Liquid-Metal Advanced Burner Reactor : components and in-vessel system thermal-hydraulic research and testing experience - pathway forward.

    Energy Technology Data Exchange (ETDEWEB)

    Kasza, K.; Grandy, C.; Chang, Y.; Khalil, H.; Nuclear Engineering Division

    2007-06-30

    This white paper provides an overview and status report of the thermal-hydraulic nuclear research and development, both experimental and computational, conducted predominantly at Argonne National Laboratory. Argonne from the early 1970s through the early 1990s was the Department of Energy's (DOE's) lead lab for thermal-hydraulic development of Liquid Metal Reactors (LMRs). During the 1970s and into the mid-1980s, Argonne conducted thermal-hydraulic studies and experiments on individual reactor components supporting the Experimental Breeder Reactor-II (EBR-II), Fast Flux Test Facility (FFTF), and the Clinch River Breeder Reactor (CRBR). From the mid-1980s and into the early 1990s, Argonne conducted studies on phenomena related to forced- and natural-convection thermal buoyancy in complete in-vessel models of the General Electric (GE) Prototype Reactor Inherently Safe Module (PRISM) and Rockwell International (RI) Sodium Advanced Fast Reactor (SAFR). These two reactor initiatives involved Argonne working closely with U.S. industry and DOE. This paper describes the very important impact of thermal hydraulics dominated by thermal buoyancy forces on reactor global operation and on the behavior/performance of individual components during postulated off-normal accident events with low flow. Utilizing Argonne's LMR expertise and design knowledge is vital to the further development of safe, reliable, and high-performance LMRs. Argonne believes there remains an important need for continued research and development on thermal-hydraulic design in support of DOE's and the international community's renewed thrust for developing and demonstrating the Global Nuclear Energy Partnership (GNEP) reactor(s) and the associated Argonne Liquid Metal-Advanced Burner Reactor (LM-ABR). This white paper highlights that further understanding is needed regarding reactor design under coolant low-flow events. These safety-related events are associated with the transition

  20. Role of Halden Reactor Project for world-wide nuclear energy development

    Energy Technology Data Exchange (ETDEWEB)

    McGrath, M.A.; Volkov, B.

    2011-07-01

    The great interest for utilization of nuclear materials to produce energy in the middle of last century needed special investigations using first class research facilities. Common problems in the area of nuclear fuel development motivated the establishment of joint research efforts. The OECD Halden Reactor Project (HRP) is a good example of such a cooperative research effort, which has been performing for more than 50 years. During that time, the Halden Reactor evolved from a prototype heavy water reactor envisaged as a power source for different applications to a research reactor that is able to simulate in-core conditions of modern commercial power reactors. The adaptability of the Halden Reactor enables the HRP to be an important international test facility for nuclear fuels and materials development. The long-term international cooperation is based on the flexible HRP organizational structure which also provides the continued success. [1,2] This paper gives a brief history of the Halden Reactor Project and its contribution to world-wide nuclear energy development. Recent expansion of the Project to the East and Asian countries may also assist and stimulate the development of a nuclear industry within these countries. The achievements of the HRP rely on the versatility of the research carried out in the reactor with reliable testing techniques and in-pile instrumentation. Diversification of scientific activity in the areas of development of alternative energy resources and man-machine technology also provide the HRP with a stable position as one of the leaders in the world scientific community. All of these aspects are described in this paper together with current experimental works, including the investigation of ULBA (Kazakhstan) production fuel in comparison with other world fuel suppliers, as well as other future and prospective plans of the Project.(Author)

  1. Fractional calculus with applications for nuclear reactor dynamics

    CERN Document Server

    Ray, Santanu Saha

    2015-01-01

    Introduces Novel Applications for Solving Neutron Transport EquationsWhile deemed nonessential in the past, fractional calculus is now gaining momentum in the science and engineering community. Various disciplines have discovered that realistic models of physical phenomenon can be achieved with fractional calculus and are using them in numerous ways. Since fractional calculus represents a reactor more closely than classical integer order calculus, Fractional Calculus with Applications for Nuclear Reactor Dynamics focuses on the application of fractional calculus to describe the physical behavi

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-06-15

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

  3. Target-fueled nuclear reactor for medical isotope production

    Science.gov (United States)

    Coats, Richard L.; Parma, Edward J.

    2017-06-27

    A small, low-enriched, passively safe, low-power nuclear reactor comprises a core of target and fuel pins that can be processed to produce the medical isotope .sup.99Mo and other fission product isotopes. The fuel for the reactor and the targets for the .sup.99Mo production are the same. The fuel can be low enriched uranium oxide, enriched to less than 20% .sup.235U. The reactor power level can be 1 to 2 MW. The reactor is passively safe and maintains negative reactivity coefficients. The total radionuclide inventory in the reactor core is minimized since the fuel/target pins are removed and processed after 7 to 21 days.

  4. Evaluation of two processes of hydrogen production starting from energy generated by high temperature nuclear reactors; Evaluacion de dos procesos de produccion de hidrogeno a partir de energia generada por reactores nucleares de alta temperatura

    Energy Technology Data Exchange (ETDEWEB)

    Valle H, J., E-mail: jvalle@upmh.edu.mx [Universidad Politecnica Metropolitana de Hidalgo, Boulevard Acceso a Tolcayuca 1009, Ex-Hacienda San Javier, 43860 Tolcayuca, Hidalgo (Mexico)

    2013-10-15

    In this work an evaluation to two processes of hydrogen production using energy generated starting from high temperature nuclear reactors (HTGR's) was realized. The evaluated processes are the electrolysis of high temperature and the thermo-chemistry cycle Iodine-Sulfur. The electrolysis of high temperature, contrary to the conventional electrolysis, allows reaching efficiencies of up to 60% because when increasing the temperature of the water, giving thermal energy, diminishes the electric power demand required to separate the molecule of the water. However, to obtain these efficiencies is necessary to have water vapor overheated to more than 850 grades C, temperatures that can be reached by the HTGR. On the other hand the thermo-chemistry cycle Iodine-Sulfur, developed by General Atomics in the 1970 decade, requires two thermal levels basically, the great of them to 850 grades C for decomposition of H{sub 2}SO{sub 4} and another minor to 360 grades C approximately for decomposition of H I, a high temperature nuclear reactor can give the thermal energy required for the process whose products would be only hydrogen and oxygen. In this work these two processes are described, complete models are developed and analyzed thermodynamically that allow to couple each hydrogen generation process to a reactor HTGR that will be implemented later on for their dynamic simulation. The obtained results are presented in form of comparative data table of each process, and with them the obtained net efficiencies. (author)

  5. Spectral structure of electron antineutrinos from nuclear reactors.

    Science.gov (United States)

    Dwyer, D A; Langford, T J

    2015-01-01

    Recent measurements of the positron energy spectrum obtained from inverse beta decay interactions of reactor electron antineutrinos show an excess in the 4 to 6 MeV region relative to current predictions. First-principles calculations of fission and beta decay processes within a typical pressurized water reactor core identify prominent fission daughter isotopes as a possible origin for this excess. These calculations also predict percent-level substructures in the antineutrino spectrum due to Coulomb effects in beta decay. Precise measurement of these substructures can elucidate the nuclear processes occurring within reactors. These substructures can be a systematic issue for measurements utilizing the detailed spectral shape.

  6. Spectral Structure of Electron Antineutrinos from Nuclear Reactors

    CERN Document Server

    Dwyer, D A

    2014-01-01

    Recent measurements of the positron energy spectrum obtained from inverse beta decay interactions of reactor electron antineutrinos show an excess in the 4 to 6 MeV region relative to current predictions. First-principle calculations of fission and beta decay processes within a typical pressurized water reactor core identify prominent fission daughter isotopes as a possible origin for this excess. These calculations also predict percent-level substructure in the antineutrino spectrum due to Coulomb effects in beta decay. Precise measurement of this substructure can constrain nuclear reactor physics. The substructure can be a systematic uncertainty for measurements utilizing the detailed spectral shape.

  7. Nuclear Safeguards Considerations For The Pebble Bed Modular Reactor (PBMR)

    Energy Technology Data Exchange (ETDEWEB)

    Phillip Casey Durst; David Beddingfield; Brian Boyer; Robert Bean; Michael Collins; Michael Ehinger; David Hanks; David L. Moses; Lee Refalo

    2009-10-01

    High temperature reactors (HTRs) have been considered since the 1940s, and have been constructed and demonstrated in the United Kingdom (Dragon), United States (Peach Bottom and Fort Saint Vrain), Japan (HTTR), Germany (AVR and THTR-300), and have been the subject of conceptual studies in Russia (VGM). The attraction to these reactors is that they can use a variety of reactor fuels, including abundant thorium, which upon reprocessing of the spent fuel can produce fissile U-233. Hence, they could extend the stocks of available uranium, provided the fuel is reprocessed. Another attractive attribute is that HTRs typically operate at a much higher temperature than conventional light water reactors (LWRs), because of the use of pyrolytic carbon and silicon carbide coated (TRISO) fuel particles embedded in ceramic graphite. Rather than simply discharge most of the unused heat from the working fluid in the power plant to the environment, engineers have been designing reactors for 40 years to recover this heat and make it available for district heating or chemical conversion plants. Demonstrating high-temperature nuclear energy conversion was the purpose behind Fort Saint Vrain in the United States, THTR-300 in Germany, HTTR in Japan, and HTR-10 and HTR-PM, being built in China. This resulted in nuclear reactors at least 30% or more thermodynamically efficient than conventional LWRs, especially if the waste heat can be effectively utilized in chemical processing plants. A modern variant of high temperature reactors is the Pebble Bed Modular Reactor (PBMR). Originally developed in the United States and Germany, it is now being redesigned and marketed by the Republic of South Africa and China. The team examined historical high temperature and high temperature gas reactors (HTR and HTGR) and reviewed safeguards considerations for this reactor. The following is a preliminary report on this topic prepared under the ASA-100 Advanced Safeguards Project in support of the NNSA Next

  8. Nuclear Analyses of Indian LLCB Test Blanket System in ITER

    Science.gov (United States)

    Swami, H. L.; Shaw, A. K.; Danani, C.; Chaudhuri, Paritosh

    2017-04-01

    Heading towards the Nuclear Fusion Reactor Program, India is developing Lead Lithium Ceramic Breeder (LLCB) tritium breeding blanket for its future fusion Reactor. A mock-up of the LLCB blanket is proposed to be tested in ITER equatorial port no.2, to ensure the overall performance of blanket in reactor relevant nuclear fusion environment. Nuclear analyses play an important role in LLCB Test Blanket System design & development. It is required for tritium breeding estimation, thermal-hydraulic design, coolants process design, radioactive waste management, equipment maintenance & replacement strategies and nuclear safety. The nuclear behaviour of LLCB test blanket module in ITER is predicated in terms of nuclear responses such as tritium production, nuclear heating, neutron fluxes and radiation damages. Radiation shielding capability of LLCB TBS inside and outside bio-shield was also assessed to fulfill ITER shielding requirements. In order to supports the rad-waste and safety assessment, nuclear activation analyses were carried out and radioactivity data were generated for LLCB TBS components. Nuclear analyses of LLCB TBS are performed using ITER recommended nuclear analyses codes (i.e. MCNP, EASY), nuclear cross section data libraries (i.e. FENDL 2.1, EAF) and neutronic model (ITER C-lite v.l). The paper describes a comprehensive nuclear performance of LLCB TBS in ITER.

  9. Development of Improved Burnable Poisons for Commercial Nuclear Power Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Renier, J.A.

    2002-04-17

    water reactor fuel core was chosen for the study, and state-of-the-art neutronic reactor core computer codes were used for analysis. Power distribution, fuel burnup, reactivity due to burnable poisons and other fission products, spectrum shift, core reactivity, moderator void coefficients, as well as other parameters were calculated as a function of time and fuel burnup. The results not only showed advantages of separation of burnable poison isotopes but revealed benefits to be achieved by careful selection of the configuration of even naturally occurring elements used as burnable poisons. The savings in terms of additional days of operation is shown in Figure 1, where the savings is plotted for each of six favorable isotopes in the four configurations. The benefit of isotope separation is most dramatic for dysprosium, but even the time savings in the case of gadolinium is several days. For a modern nuclear plant, one day's worth of electricity is worth about one million dollars, so the resulting savings of only a few days is considerable. It is also apparent that the amount of savings depends upon the configuration of the burnable poison.

  10. Nuclear instrumentation system operating experience and nuclear instrument testing in the EBR-II

    Energy Technology Data Exchange (ETDEWEB)

    Yingling, G. E.; Curran, R. N.

    1980-01-01

    In March of 1972 three wide range nuclear channels were purchased from Gulf Atomics Corporation and installed in EBR-II as a test. The three channels were operated as a test until April 1975 when they became a permanent part of the reactor shutdown system. Also described are the activities involved in evaluating and qualifying neutron detectors for LMFBR applications. Included are descriptions of the ANL Components Technology Division Test Program and the EBR-II Nuclear Instrument Test Facilities (NITF) used for the in-reactor testing and a summary of program test results from EBR-II.

  11. Selection of power plant elements for future reactor space electric power systems

    Energy Technology Data Exchange (ETDEWEB)

    Buden, D.; Bennett, G.A.; Copper, K.

    1979-09-01

    Various types of reactor designs, electric power conversion equipment, and reject-heat systems to be used in nuclear reactor power plants for future space missions were studied. The designs included gas-cooled, liquid-cooled, and heat-pipe reactors. For the power converters, passive types such as thermoelectric and thermionic converters and dynamic types such as Brayton, potassium Rankine, and Stirling cycles were considered. For the radiators, heat pipes for transfer and radiating surface, pumped fluid for heat transfer with fins as the radiating surface, and pumped fluid for heat transfer with heat pipes as the radiating surface were considered. After careful consideration of weights, sizes, reliabilities, safety, and development cost and time, a heat-pipe reactor design, thermoelectric converters, and a heat-pipe radiator for an experimental program were selected.

  12. Nuclear Power Plants. Revised.

    Science.gov (United States)

    Lyerly, Ray L.; Mitchell, Walter, III

    This publication is one of a series of information booklets for the general public published by the United States Atomic Energy Commission. Among the topics discussed are: Why Use Nuclear Power?; From Atoms to Electricity; Reactor Types; Typical Plant Design Features; The Cost of Nuclear Power; Plants in the United States; Developments in Foreign…

  13. Reactor Fuel Isotopics and Code Validation for Nuclear Applications

    Energy Technology Data Exchange (ETDEWEB)

    Francis, Matthew W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Weber, Charles F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Pigni, Marco T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gauld, Ian C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-02-01

    Experimentally measured isotopic concentrations of well characterized spent nuclear fuel (SNF) samples have been collected and analyzed by previous researchers. These sets of experimental data have been used extensively to validate the accuracy of depletion code predictions for given sets of burnups, initial enrichments, and varying power histories for different reactor types. The purpose of this report is to present the diversity of data in a concise manner and summarize the current accuracy of depletion modeling. All calculations performed for this report were done using the Oak Ridge Isotope GENeration (ORIGEN) code, an internationally used irradiation and decay code solver within the SCALE comprehensive modeling and simulation code. The diversity of data given in this report includes key actinides, stable fission products, and radioactive fission products. In general, when using the current ENDF/B-VII.0 nuclear data libraries in SCALE, the major actinides are predicted to within 5% of the measured values. Large improvements were seen for several of the curium isotopes when using improved cross section data found in evaluated nuclear data file ENDF/B-VII.0 as compared to ENDF/B-V-based results. The impact of the flux spectrum on the plutonium isotope concentrations as a function of burnup was also shown. The general accuracy noted for the actinide samples for reactor types with burnups greater than 5,000 MWd/MTU was not observed for the low-burnup Hanford B samples. More work is needed in understanding these large discrepancies. The stable neodymium and samarium isotopes were predicted to within a few percent of the measured values. Large improvements were seen in prediction for a few of the samarium isotopes when using the ENDF/B-VII.0 libraries compared to results obtained with ENDF/B-V libraries. Very accurate predictions were obtained for 133Cs and 153Eu. However, the predicted values for the stable ruthenium and rhodium isotopes varied

  14. Desalination plan with nuclear reactors as part of a sustainable development program in Mexico; Plan de desalinizacion con reactores nucleares como parte de un programa de desarrollo sustentable en Mexico

    Energy Technology Data Exchange (ETDEWEB)

    Rojas A, O; Calleros M, G., E-mail: oziel.rojas.siimisa@gmail.com [Soluciones en Instrumentacion Integral y Mantenimiento Industrial y Servicios, S. A. de C. V. (Mexico)

    2016-09-15

    This paper presents a project for the desalination of seawater with nuclear reactors, in order to supply fresh water to the populations near to the nuclear power plant. A case study is proposed with the nuclear power plant of Laguna Verde, implementing a system that allows taking advantage of the residual heat of the seawater condensate stage and with this, to supply drinking water to the surrounding localities where the vital liquid is scarce. In addition, legislation is proposed to allow some of the thermal energy generated by reactors producing electrical energy in Mexico to be used for the desalination of seawater and/or hydrogen production. (Author)

  15. University Reactor Conversion Lessons Learned Workshop for Texas A&M University Nuclear Science Center Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Eric C. Woolstenhulme; Dana M. Meyer

    2007-04-01

    The objectives of this meeting were to capture the observations, insights, issues, concerns, and ideas of those involved in the Texas A&M University Nuclear Science Center (TAMU NSC) TRIGA Reactor Conversion so that future efforts can be conducted with greater effectiveness, efficiency, and with fewer challenges. This workshop was held in conjunction with a similar workshop for the University of Florida Reactor Conversion. Some of the generic lessons from that workshop are included in this report for completeness.

  16. Production capabilities in US nuclear reactors for medical radioisotopes

    Energy Technology Data Exchange (ETDEWEB)

    Mirzadeh, S.; Callahan, A.P.; Knapp, F.F. Jr. [Oak Ridge National Lab., TN (United States); Schenter, R.E. [Westinghouse Hanford Co., Richland, WA (United States)

    1992-11-01

    The availability of reactor-produced radioisotopes in the United States for use in medical research and nuclear medicine has traditionally depended on facilities which are an integral part of the US national laboratories and a few reactors at universities. One exception is the reactor in Sterling Forest, New York, originally operated as part of the Cintichem (Union Carbide) system, which is currently in the process of permanent shutdown. Since there are no industry-run reactors in the US, the national laboratories and universities thus play a critical role in providing reactor-produced radioisotopes for medical research and clinical use. The goal of this survey is to provide a comprehensive summary of these production capabilities. With the temporary shutdown of the Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) in November 1986, the radioisotopes required for DOE-supported radionuclide generators were made available at the Brookhaven National Laboratory (BNL) High Flux Beam Reactor (HFBR). In March 1988, however, the HFBR was temporarily shut down which forced investigators to look at other reactors for production of the radioisotopes. During this period the Missouri University Research Reactor (MURR) played an important role in providing these services. The HFIR resumed routine operation in July 1990 at 85 MW power, and the HFBR resumed operation in June 1991, at 30 MW power. At the time of the HFBR shutdown, there was no available comprehensive overview which could provide information on status of the reactors operating in the US and their capabilities for radioisotope production. The obvious need for a useful overview was thus the impetus for preparing this survey, which would provide an up-to-date summary of those reactors available in the US at both the DOE-funded national laboratories and at US universities where service irradiations are currently or expected to be conducted.

  17. Non destructive testing of irradiated fuel assemblies at the IEA-R1 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Jose Eduardo Rosa da; Terremoto, Luis Antonio Albiac; Castanheira, Myrthes; Teodoro, Celso Antonio; Silva, Antonio Teixeira e; Damy, Margaret de Almeida; Lucki, Georgi [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)]. E-mails: jersilva@ipen.br; laaterre@ipen.br; myrthes@ipen.br; cteodoro@ipen.br; teixeira@ipen.br; madamy@ipen.br; glucki@ipen.br

    2007-07-01

    Fuel performance and nuclear fuel qualification require a post-irradiation analysis. Non-destructive methods are utilised both in irradiated fuel storage pools and in hot-cells laboratories. As Brazil does not have hot-cells facilities for post-irradiation analysis, a qualification program for the Material Testing Reactor (MTR) fuel elements made at IPEN/CNEN-SP was adopted, based on non-destructive tests. The IPEN Fuel Engineering Group - CENC developed basic facilities for fuels post-irradiated analysis inside the reactor pool, which gives indications of: general state, by visual inspection; the integrity of the irradiated fuel cladding, by sipping tests; thickness measurements of the fuel miniplates during the irradiation time, for swelling evaluation; and, local burn-up evaluation by gamma spectrometry along the active area of the fuel element. This work describes that facilities, equipment and examples of some irradiated fuels analysis performed. (author)

  18. Global risk of radioactive fallout after major nuclear reactor accidents

    Science.gov (United States)

    Lelieveld, J.; Kunkel, D.; Lawrence, M. G.

    2012-05-01

    Major reactor accidents of nuclear power plants are rare, yet the consequences are catastrophic. But what is meant by "rare"? And what can be learned from the Chernobyl and Fukushima incidents? Here we assess the cumulative, global risk of exposure to radioactivity due to atmospheric dispersion of gases and particles following severe nuclear accidents (the most severe ones on the International Nuclear Event Scale, INES 7), using particulate 137Cs and gaseous 131I as proxies for the fallout. Our results indicate that previously the occurrence of INES 7 major accidents and the risks of radioactive contamination have been underestimated. Using a global model of the atmosphere we compute that on average, in the event of a major reactor accident of any nuclear power plant worldwide, more than 90% of emitted 137Cs would be transported beyond 50 km and about 50% beyond 1000 km distance before being deposited. This corroborates that such accidents have large-scale and trans-boundary impacts. Although the emission strengths and atmospheric removal processes of 137Cs and 131I are quite different, the radioactive contamination patterns over land and the human exposure due to deposition are computed to be similar. High human exposure risks occur around reactors in densely populated regions, notably in West Europe and South Asia, where a major reactor accident can subject around 30 million people to radioactive contamination. The recent decision by Germany to phase out its nuclear reactors will reduce the national risk, though a large risk will still remain from the reactors in neighbouring countries.

  19. Global risk of radioactive fallout after major nuclear reactor accidents

    Directory of Open Access Journals (Sweden)

    J. Lelieveld

    2012-05-01

    Full Text Available Major reactor accidents of nuclear power plants are rare, yet the consequences are catastrophic. But what is meant by "rare"? And what can be learned from the Chernobyl and Fukushima incidents? Here we assess the cumulative, global risk of exposure to radioactivity due to atmospheric dispersion of gases and particles following severe nuclear accidents (the most severe ones on the International Nuclear Event Scale, INES 7, using particulate 137Cs and gaseous 131I as proxies for the fallout. Our results indicate that previously the occurrence of INES 7 major accidents and the risks of radioactive contamination have been underestimated. Using a global model of the atmosphere we compute that on average, in the event of a major reactor accident of any nuclear power plant worldwide, more than 90% of emitted 137Cs would be transported beyond 50 km and about 50% beyond 1000 km distance before being deposited. This corroborates that such accidents have large-scale and trans-boundary impacts. Although the emission strengths and atmospheric removal processes of 137Cs and 131I are quite different, the radioactive contamination patterns over land and the human exposure due to deposition are computed to be similar. High human exposure risks occur around reactors in densely populated regions, notably in West Europe and South Asia, where a major reactor accident can subject around 30 million people to radioactive contamination. The recent decision by Germany to phase out its nuclear reactors will reduce the national risk, though a large risk will still remain from the reactors in neighbouring countries.

  20. Application of a Russian nuclear reactor simulator VVER-1000; Aplicacion de un simulador de reactor nuclear ruso VVER-1000

    Energy Technology Data Exchange (ETDEWEB)

    Lopez-Peniche S, A. [UNAM, Facultad de Ingenieria, Circuito Interior, Ciudad Universitaria, 04360 Mexico D. F. (Mexico); Salazar S, E., E-mail: alpsordo@hotmail.com [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Laboratorio de Analisis en Ingenieria de Reactores Nucleares, 62250 Jiutepec, Morelos (Mexico)

    2012-10-15

    The objective of the present work is to give to know the most important characteristics in the Russian nuclear reactor of pressurized light water VVER-1000, doing emphasis in the differences that has with the western equivalent the reactor PWR in the design and the safety systems. Therefore, a description of the computerized simulation of the reactor VVER-1000 developed by the company Eniko TSO that the International Atomic of Energy Agency distributes to the states members with academic purposes will take place. The simulator includes mathematical models that represent to the essential systems in the real nuclear power plant, for what is possible to reproduce common faults and transitory characteristic of the nuclear industry with a behavior sufficiently attached to the reality. In this work is analyzed the response of the system before a turbine shot. After the accident in the nuclear power plant of Three Mile Island (US) they have been carried out improvements in the design of the reactor PWR and their safety systems. To know the reach and the limitations of the program, the events that gave place to this accident will be reproduced in the simulator VVER-1000. With base to the results of the simulation we will conclude that so reliable is the response of the safety system of this reactor. (Author)

  1. Sustainability and the Fixed Bed Nuclear Reactor (FBNR

    Directory of Open Access Journals (Sweden)

    Farhang Sefidvash

    2012-08-01

    Full Text Available Sustainability as a multifaceted and holistic concept is analyzed. Sustainability involves human relationship with elements such as natural environment, economy, power, governance, education and technology with the ultimate purpose of carrying forward an ever-advancing civilization. The Fixed Bed Nuclear Reactor (FBNR is an innovative, small, simple in design, inherently safe, non-proliferating, and environmentally friendly concept that its deployment can generate energy in a sustainable manner contributing to the prosperity of humanity. The development of FBNR will provide electricity as well as desalinated water through a simple but advanced technology for the developing, as well as developed countries. FBNR is environmentally friendly due to its inherent safety and the convenience of using its spent fuel as the source of radiation for irradiation purposes in agriculture, industry, and medicine. Politically, if a ping pong game brought peace between China and USA, a program of development of FBNR supported by the peace loving international community can become a more mature means to bring peace among certain apparently hostile nations who crave sustainable energy, desalinated water and simple advanced technology.

  2. Handbook of nuclear engineering: vol 1: nuclear engineering fundamentals; vol 2: reactor design; vol 3: reactor analysis; vol 4: reactors of waste disposal and safeguards

    CERN Document Server

    2013-01-01

    The Handbook of Nuclear Engineering is an authoritative compilation of information regarding methods and data used in all phases of nuclear engineering. Addressing nuclear engineers and scientists at all academic levels, this five volume set provides the latest findings in nuclear data and experimental techniques, reactor physics, kinetics, dynamics and control. Readers will also find a detailed description of data assimilation, model validation and calibration, sensitivity and uncertainty analysis, fuel management and cycles, nuclear reactor types and radiation shielding. A discussion of radioactive waste disposal, safeguards and non-proliferation, and fuel processing with partitioning and transmutation is also included. As nuclear technology becomes an important resource of non-polluting sustainable energy in the future, The Handbook of Nuclear Engineering is an excellent reference for practicing engineers, researchers and professionals.

  3. Reliability and safety of the electrical power supply complex of the Hanford production reactors

    Energy Technology Data Exchange (ETDEWEB)

    Robbins, F.D.

    1960-09-15

    Safety has been and must continue to be the inviolable modulus by which the operation of a nuclear reactor must be judged. A malfunction in any reactor may well result in a release of fission products which may dissipate over a wide geographical area. Such dissipation may place the health, happiness and even the lives of the people in the region in serious jeopardy. As a result, the property damage and liability cost may reach astronomical values in the order of magnitude of billions of dollars. Reliability of the electrical network is an indispensable factor in attaining a high order of safety assurance. Progress in the peaceful use of atomic energy may take the form of electrical power generation using the nuclear reactor as a source of thermal energy. In view of these factors it seems appropriate and profitable that a critical engineering study be made of the safety and reliability of the Hanford reactors without regard to cost economics. This individual and independent technical engineering analysis was made without regard to Hanford traditional engineering and administration assignments. The main objective has been to focus attention on areas which seem to merit further detailed study on conditions which seem to need adjustment but most of all on those changes which will improve reactor safety. This report is the result of such a study.

  4. Annual report of Power Reactor and Nuclear Fuel Development Corporation, fiscal year 1994

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-09-01

    This was the Annual Report of the Power Reactor and Nuclear Fuel Development Corporation, Fiscal Year of 1994. In this report, the following 12 items are described: (1) Development of the fast breeding reactor; (a) operation of the fast experimental reactor, `Joyo`, (b) construction and trial operation of the fast breeding prototype reactor, `Monju`, and (c) R and D of FBR; (2) Development of the new type conversion reactor; (a) operation of prototype reactor, `Fugen`, and (b) R and D of ATR; (3) Development of uranium mining and conversion; (4) Development of uranium concentration technology; (5) Development of plutonium fuel; (a) preparation of the MOX fuel, (b) preparation facility construction of the MOX fuel, (c) R and D of plutonium fuel. and (d) technical development of plutonium mixing and conversion; (6) Reprocessing of spent fuel; (7) Environmental technology development of radioactive waste; (8) Creative and innovative R and D; (9) Management and nuclear non-proliferation countermeasure of nuclear matter; (10) Safety management and safety study; (11) Related common business; and (12) General management business. (G.K.)

  5. Nuclear electric propulsion options for piloted Mars missions

    Science.gov (United States)

    George, Jeffrey A.

    Three nuclear electric propulsion (NEP) systems are discussed. The three systems are as follows: a system based on current SP-100 technology; a potassium Rankine-cycle based power conversion system, and an argon ion thruster system. The system will be researched for implementation in several possible vehicle configurations. The following are among the possible Mars vehicle configurations: a piloted 15 MWe multi-reactor vehicle; a piloted 10 MWe vehicle with ECCV; a piloted 10 MWe modular vehicle; piloted 10 and 15 MWe vehicles with ECCV and MEV; a piloted 5 MWe vehicle with ECCV; a 5 MWe cargo vehicle with 2 MEV's; and a 2.5 MWe vehicle with MEV.

  6. Contribution of prototypic material tests on the Plinius platform to the study of nuclear reactor severe accident; Contribution des essais en materiaux prototypiques sur la plate-forme Plinius a l'etude des accidents graves de reacteurs nucleaires

    Energy Technology Data Exchange (ETDEWEB)

    Journeau, Ch

    2008-01-15

    The PLINIUS experimental platform at CEA Cadarache is dedicated to the experimental study of nuclear reactor severe accidents thanks to experiments between 2000 and 3500 K with prototypic corium. Corium is the mixture that would be formed by an hypothetical core melting and its mixing with structural materials. Prototypical corium has the same chemical composition as the corium corresponding to a given accident scenario but has a different isotopic composition (use of depleted uranium,...). Research programs and test series have been performed to study corium thermophysical properties, fission product behaviour, corium spreading, solidification and interaction with concrete as well as its coolability. It was the frame of research training of many students and was realized within national, European and international collaborations. (author)

  7. Neutron Capture and the Antineutrino Yield from Nuclear Reactors.

    Science.gov (United States)

    Huber, Patrick; Jaffke, Patrick

    2016-03-25

    We identify a new, flux-dependent correction to the antineutrino spectrum as produced in nuclear reactors. The abundance of certain nuclides, whose decay chains produce antineutrinos above the threshold for inverse beta decay, has a nonlinear dependence on the neutron flux, unlike the vast majority of antineutrino producing nuclides, whose decay rate is directly related to the fission rate. We have identified four of these so-called nonlinear nuclides and determined that they result in an antineutrino excess at low energies below 3.2 MeV, dependent on the reactor thermal neutron flux. We develop an analytic model for the size of the correction and compare it to the results of detailed reactor simulations for various real existing reactors, spanning 3 orders of magnitude in neutron flux. In a typical pressurized water reactor the resulting correction can reach ∼0.9% of the low energy flux which is comparable in size to other, known low-energy corrections from spent nuclear fuel and the nonequilibrium correction. For naval reactors the nonlinear correction may reach the 5% level by the end of cycle.

  8. Neutron Capture and the Antineutrino Yield from Nuclear Reactors

    Science.gov (United States)

    Huber, Patrick; Jaffke, Patrick

    2016-03-01

    We identify a new, flux-dependent correction to the antineutrino spectrum as produced in nuclear reactors. The abundance of certain nuclides, whose decay chains produce antineutrinos above the threshold for inverse beta decay, has a nonlinear dependence on the neutron flux, unlike the vast majority of antineutrino producing nuclides, whose decay rate is directly related to the fission rate. We have identified four of these so-called nonlinear nuclides and determined that they result in an antineutrino excess at low energies below 3.2 MeV, dependent on the reactor thermal neutron flux. We develop an analytic model for the size of the correction and compare it to the results of detailed reactor simulations for various real existing reactors, spanning 3 orders of magnitude in neutron flux. In a typical pressurized water reactor the resulting correction can reach ˜0.9 % of the low energy flux which is comparable in size to other, known low-energy corrections from spent nuclear fuel and the nonequilibrium correction. For naval reactors the nonlinear correction may reach the 5% level by the end of cycle.

  9. THE COMPONENT TEST FACILITY – A NATIONAL USER FACILITY FOR TESTING OF HIGH TEMPERATURE GAS-COOLED REACTOR (HTGR) COMPONENTS AND SYSTEMS

    Energy Technology Data Exchange (ETDEWEB)

    David S. Duncan; Vondell J. Balls; Stephanie L. Austad

    2008-09-01

    The Next Generation Nuclear Plant (NGNP) and other High-Temperature Gas-cooled Reactor (HTGR) Projects require research, development, design, construction, and operation of a nuclear plant intended for both high-efficiency electricity production and high-temperature industrial applications, including hydrogen production. During the life cycle stages of an HTGR, plant systems, structures and components (SSCs) will be developed to support this reactor technology. To mitigate technical, schedule, and project risk associated with development of these SSCs, a large-scale test facility is required to support design verification and qualification prior to operational implementation. As a full-scale helium test facility, the Component Test facility (CTF) will provide prototype testing and qualification of heat transfer system components (e.g., Intermediate Heat Exchanger, valves, hot gas ducts), reactor internals, and hydrogen generation processing. It will perform confirmation tests for large-scale effects, validate component performance requirements, perform transient effects tests, and provide production demonstration of hydrogen and other high-temperature applications. Sponsored wholly or in part by the U.S. Department of Energy, the CTF will support NGNP and will also act as a National User Facility to support worldwide development of High-Temperature Gas-cooled Reactor technologies.

  10. Desalination of seawater with nuclear power reactors in cogeneration; Desalacion de agua de mar con reactores nucleares de potencia en cogeneracion

    Energy Technology Data Exchange (ETDEWEB)

    Flores E, R.M

    2004-07-01

    The growing demand for energy and hydraulic resources for satisfy the domestic, industrial, agricultural activities, etc. has wakened up the interest to carry out concerning investigations to study the diverse technologies guided to increase the available hydraulic resources, as well as to the search of alternatives of electric power generation, economic and socially profitable. In this sense the possible use of the nuclear energy is examined in cogeneration to obtain electricity and drinkable water for desalination of seawater. The technologies are analysed involved in the nuclear cogeneration (desalination technology, nuclear and desalination-nuclear joining) available in the world. At the same time it is exemplified the coupling of a nuclear reactor and a process of hybrid desalination that today in day the adult offers and economic advantages. Finally, the nuclear desalination is presented as a technical and economically viable solution in regions where necessities of drinkable water are had for the urban, agricultural consumption and industrial in great scale and that for local situations it is possible to satisfy it desalinating seawater. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-03-01

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

  12. Alloying of steel and graphite by hydrogen in nuclear reactor

    Science.gov (United States)

    Krasikov, E.

    2017-02-01

    In traditional power engineering hydrogen may be one of the first primary source of equipment damage. This problem has high actuality for both nuclear and thermonuclear power engineering. Study of radiation-hydrogen embrittlement of the steel raises the question concerning the unknown source of hydrogen in reactors. Later unexpectedly high hydrogen concentrations were detected in irradiated graphite. It is necessary to look for this source of hydrogen especially because hydrogen flakes were detected in reactor vessels of Belgian NPPs. As a possible initial hypothesis about the enigmatical source of hydrogen one can propose protons generation during beta-decay of free neutrons поскольку inasmuch as protons detected by researches at nuclear reactors as witness of beta-decay of free neutrons.

  13. Optimizing Nuclear Reactor Operation Using Soft Computing Techniques

    NARCIS (Netherlands)

    Entzinger, J.O.; Ruan, D.; Kahraman, Cengiz

    2006-01-01

    The strict safety regulations for nuclear reactor control make it di±cult to implement new control techniques such as fuzzy logic control (FLC). FLC however, can provide very desirable advantages over classical control, like robustness, adaptation and the capability to include human experience into

  14. Method of controlling crystallite size in nuclear-reactor fuels

    Science.gov (United States)

    Lloyd, M.H.; Collins, J.L.; Shell, S.E.

    Improved spherules for making enhanced forms of nuclear-reactor fuels are prepared by internal gelation procedures within a sol-gel operation and are accomplished by first boiling the concentrated HMTA-urea feed solution before engaging in the spherule-forming operation thereby effectively controlling crystallite size in the product spherules.

  15. Global risk of radioactive fallout after nuclear reactor accidents

    Science.gov (United States)

    Kunkel, D.; Lelieveld, J.; Lawrence, M. G.

    2012-04-01

    Reactor core meltdowns of nuclear power plants are rare, yet the consequences are catastrophic. But what is meant by "rare"? And what can be learned from the Chernobyl and Fukushima incidents? Here we assess the risk of exposure to radioactivity due to atmospheric dispersion of gases and particles following severe nuclear accidents, using particulate 137Cs and gaseous 131I as proxies for the fallout. It appears that previously the occurrence of major accidents and the risks of radioactive contamination have been underestimated. Using a global model of the atmosphere we compute that on average, in the event of a core melt of any nuclear power plant worldwide, more than 90 % of emitted 137Cs would be transported beyond 50 km and about 50 % beyond 1000 km distance. This corroborates that such accidents have large-scale and trans-boundary impacts. Although the emission strengths and atmospheric removal processes of 137Cs and 131I are quite different, the radioactive contamination patterns over land and the human deposition exposure are computed to be similar. High human exposure risks occur around reactors in densely populated regions, notably in southern Asia where a core melt can subject 55 million people to radioactive contamination. The recent decision by Germany to phase out its nuclear reactors will reduce the national risk, though a large risk will still remain from the reactors in neighbouring countries.

  16. Use of hafnium in control bars of nuclear reactors; Uso de hafnio en barras de control de reactores nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez S, J.R.; Alonso V, G. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: jrrs@nuclear.inin-mx

    2003-07-01

    Recently the use of hafnium as neutron absorber material in nuclear reactors has been reason of investigation by virtue of that this material has nuclear properties as to the neutrons absorption and structural that can prolong the useful life of the control mechanisms of the nuclear reactors. In this work some of those more significant hafnium properties are presented like nuclear material. Also there are presented calculations carried out with the HELIOS code for fuel cells of uranium oxide and of uranium and plutonium mixed oxides under controlled conditions with conventional bars of boron carbide and also with similar bars to which are substituted the absorbent material by metallic hafnium, the results are presented in this work. (Author)

  17. Nuclear reactors situation in Japan after the major earthquake of March 11, 2011. March 12, 2011, 8:00 PM status; Situation des reacteurs nucleaires au Japon suite au seisme majeur servenu le 11 mars 2011. Point de situation du 12 mars 2011 a 20 heures

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2011-07-01

    This situation note is established according to the information gained on March 12, 2011, at 8:00 PM, by the crisis centre of the French institute of radiation protection and nuclear safety (IRSN). The situation of the reactors No. 1, 2 and 3 of the Fukushima I site is briefly presented with the progress of the accident management actions. The operation principles of a BWR-type reactor and of a PWR-type reactor are presented in appendix as well as the confinement principle specific to Mark I-type BWR reactors designed by General Electric. The meteorological forecasts of the day are presented in a figure. (J.S.)

  18. Sipping test update device for fuel elements cladding inspections in IPR-r1 TRIGA reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rodrigues, R.R.; Mesquita, A.Z.; Andrade, E.P.D.; Gual, Maritza R., E-mail: rrr@cdtn.br, E-mail: amir@cdtn.br, E-mail: edson@cdtn.br, E-mail: maritzargual@gmail.com [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    It is in progress at the Centro de Desenvolvimento da Tecnologia Nuclear - CDTN (Nuclear Technology Development Center), a research project that aims to investigate possible leaks in the fuel elements of the TRIGA reactor, located in this research center. This paper presents the final form of sipping test device for TRIGA reactor, and results of the first experiments setup. Mechanical support strength tests were made by knotting device on the crane, charged with water from the conventional water supply, and tests outside the reactor pool with the use of new non-irradiated fuel elements encapsulated in stainless steel, and available safe stored in this unit. It is expected that tests with graphite elements from reactor pool are done soon after and also the test experiment with the first fuel elements in service positioned in the B ring (central ring) of the reactor core in the coming months. (author)

  19. 77 FR 39521 - Application for a License To Export Nuclear Reactor Major Components and Equipment

    Science.gov (United States)

    2012-07-03

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Application for a License To Export Nuclear Reactor Major Components and Equipment Pursuant to 10... Reactor internals, Components and For use in Braka nuclear power Company LLC reactor coolant equipment...

  20. Heat Pipe Reactor Dynamic Response Tests: SAFE-100 Reactor Core Prototype

    Science.gov (United States)

    Bragg-Sitton, Shannon M.

    2005-01-01

    The SAFE-I00a test article at the NASA Marshall Space Flight Center was used to simulate a variety of potential reactor transients; the SAFEl00a is a resistively heated, stainless-steel heat-pipe (HP)-reactor core segment, coupled to a gas-flow heat exchanger (HX). For these transients the core power was controlled by a point kinetics model with reactivity feedback based on core average temperature; the neutron generation time and the temperature feedback coefficient are provided as model inputs. This type of non-nuclear test is expected to provide reasonable approximation of reactor transient behavior because reactivity feedback is very simple in a compact fast reactor (simple, negative, and relatively monotonic temperature feedback, caused mostly by thermal expansion) and calculations show there are no significant reactivity effects associated with fluid in the HP (the worth of the entire inventory of Na in the core is .tests, the point kinetics model was based on core thermal expansion via deflection measurements. It was found that core deflection was a strung function of how the SAFE-100 modules were fabricated and assembled (in terms of straightness, gaps, and other tolerances). To remove the added variable of how this particular core expands as compared to a different concept, it was decided to use a temperature based feedback model (based on several thermocouples placed throughout the core).

  1. Modeling of Flow in Nuclear Reactor Fuel Cell Outlet

    Directory of Open Access Journals (Sweden)

    František URBAN

    2010-12-01

    Full Text Available Safe and effective load of nuclear reactor fuel cells demands qualitative and quantitative analysis of relations between coolant temperature in fuel cell outlet temperature measured by thermocouple and middle temperature of coolant in thermocouple plane position. In laboratory at Insitute of thermal power engineering of the Slovak University of Technology in Bratislava was installed an experimental physical fuel cell model of VVER 440 nuclear power plant with V 213 nuclear reactors. Objective of measurements on physical model was temperature and velocity profiles analysis in the fuel cell outlet. In this paper the measured temperature and velocity profiles are compared with the results of CFD simulation of fuel cell physical model coolant flow.

  2. Advanced methods for nuclear reactor gas laser coupling

    Energy Technology Data Exchange (ETDEWEB)

    Miley, G.H.; Verdeyen, J.T.

    1978-06-01

    Research is described that led to the discovery of three nuclear-pumped lasers (NPLs) using mixtures of Ne--N/sub 2/, He--Hg, and He or Ne with CO or CO/sub 2/. The Ne--N/sub 2/ NPL was the first laser obtained with modest neutron fluxes from a TRIGA reactor (vs fast burst reactors used elsewhere in such work), the He--Hg NPL was the first visible nuclear-pumped laser, while the Ne--CO and He--CO/sub 2/ lasers are the first to provide energy storage on a millisecond time scale. Important potential applications of NPLs include coupling and power transmission from remote power stations such as nuclear plants in satellites and neutron-feedback operation of inertial confinement fusion plants.

  3. MEGAHIT Roadmap: Applications for Nuclear Electric Propulsion

    OpenAIRE

    Jansen, Frank; Semenkin, Alexander; Bauer, Waldemar; WORMS, Jean-Claude; Detsis, Emmanouil; CLIQUET-MORENO, Elisa; Masson, Frederic; Ruault, Jean-Marc; Gaia, Enrico; Cristina, T.M.; Tinsley, Tim; Hodgson, Zara

    2014-01-01

    The paper introduces the three EC funded nuclear electric propulsion funded projects DiPoP, MEGAHIT and DEMOCRITOS. It describes in detail the European-Russian MEGAHIT project - the study outputs, the proposal for a key technology plan, a plan for a political and public supportable reference space mission. Moreover the content of the MEGAHIT global roadmap for international realization of the INPPS (International Nuclear Power and Propulsion System) is sketched.

  4. Emission and transmission tomography systems to be developed for the future needs of Jules Horowitz material testing reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kotiluoto, Petri [VTT Technical Research Centre of Finland, P.O.Box 1000, FI-02044 VTT (Finland)], E-mail: petri.kotiluoto@vtt.fi; Wasastjerna, Frej; Kekki, Tommi [VTT Technical Research Centre of Finland, P.O.Box 1000, FI-02044 VTT (Finland); Sipilae, Heikki; Banzuzi, Kukka [Oxford Instruments Analytical Oy, Nihtisillankuja 5, P.O.Box 85, FI-02631 Espoo (Finland); Kinnunen, Petri; Heikinheimo, Liisa [VTT Technical Research Centre of Finland, P.O.Box 1000, FI-02044 VTT (Finland)

    2009-08-01

    The new 100 MW Jules Horowitz material testing reactor will be built in Cadarache, France. It will support, for instance, research on new types of innovative nuclear fuel. As a Finnish in-kind contribution, 3D emission and transmission tomography equipment will be delivered for both the reactor and the active component storage pool. The image reconstruction of activities inside the used nuclear fuel will be based on gamma spectrometry measurements. A new type of underwater digital X-ray linear detector array is under development for transmission imaging, based on GaAs and direct conversion of X-rays into an electrical signal. A shared collimator will be used for both emission and transmission measurements. Some preliminary design has been performed. For the current design, the expected gamma spectrometric response of a typical high-purity germanium detector has been simulated with MCNP for minimum and maximum source activities (specified by CEA) to be measured in future.

  5. Neutron capture and the antineutrino yield from nuclear reactors

    CERN Document Server

    Huber, Patrick

    2015-01-01

    We identify a new, flux-dependent correction to the antineutrino spectrum as produced in nuclear reactors. The abundance of certain nuclides, whose decay chains produce antineutrinos above the threshold for inverse beta decay, has a nonlinear dependence on the neutron flux, unlike the vast majority of antineutrino producing nuclides, whose decay rate is directly related to the fission rate. We have identified four of these so-called nonlinear nuclides and determined that they result in an antineutrino excess at low-energies below 3.2MeV, dependent on the reactor thermal neutron flux. We develop an analytic model for the size of the correction and compare it to the results of detailed reactor simulations for various real existing reactors, spanning 3 orders of magnitude in neutron flux. In a typical pressurized water reactor the resulting correction can reach 0.9% of the low energy flux which is comparable in size to other, known low-energy corrections from spent nuclear fuel and the non-equilibrium correction...

  6. Nuclear power plants for mobile applications

    Science.gov (United States)

    Anderson, J. L.

    1972-01-01

    Mobile nuclear powerplants for applications other than large ships and submarines will require compact, lightweight reactors with especially stringent impact-safety design. The technical and economic feasibility that the broadening role of civilian nuclear power, in general, (land-based nuclear electric generating plants and nuclear ships) can extend to lightweight, safe mobile nuclear powerplants are examined. The paper discusses technical experience, identifies potential sources of technology for advanced concepts, cites the results of economic studies of mobile nuclear powerplants, and surveys future technical capabilities needed by examining the current use and projected needs for vehicles, machines, and habitats that could effectively use mobile nuclear reactor powerplants.

  7. Molten-salt reactors for efficient nuclear fuel utilization without plutonium separation

    Energy Technology Data Exchange (ETDEWEB)

    Engel, J.R.; Grimes, W.R.; Rhoades, W.A.; Dearing, J.F.

    1978-08-01

    Research and development studies of molten-salt reactors (MSRs) for special purposes have been under way since 1947 and for possible application as possible commercial nuclear electric power generators since 1956. For the latter, the previous emphasis has been on breeding performance and low fissile inventory to help limit the demand on nonrenewable natural resources (uranium) in an expanding nuclear economy; little or no thought has been given to alternative uses of nuclear fuels such as proliferation of nuclear explosives. As a consequence, the conceptual designs that evolved (e.g., the ORNL reference design MSBR) all favored enriched /sup 233/U as fuel with an on-site chemical processing facility from which portions of that fuel could be diverted fairly easily. With the current interest in limiting the proliferation potential of nuclear electric power systems, a redirected study of MSRs was undertaken in an effort to identify conceptual systems that would be attractive in this situation. It appears that practical proliferation-resistant MSRs could be designed and built, and the report describes a particularly attractive break-even breeder that includes an on-site chemical reprocessing facility within the reactor primary containment.

  8. Fundamental aspects of nuclear reactor fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Olander, D.R.

    1976-01-01

    The book presented is designed to function both as a text for first-year graduate courses in nuclear materials and as a reference for workers involved in the materials design and performance aspects of nuclear power plants. The contents are arranged under the following chapter headings: statistical thermodynamics, thermal properties of solids, crystal structures, cohesive energy of solids, chemical equilibrium, point defects in solids, diffusion in solids, dislocations and grain boundaries, equation of state of UO/sub 2/, fuel element thermal performance, fuel chemistry, behavior of solid fission products in oxide fuel elements, swelling due to fission gases, pore migration and fuel restructuring kinetics, fission gas release, mechanical properties of UO/sub 2/, radiation damage, radiation effects in metals, interaction of sodium and stainless steel, modeling of the structural behavior of fuel elements and assemblies. (DG)

  9. Evaluating the consequences of loss of flow accident for a typical VVER-1000 nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mirvakili, S.M.; Safaei, S. [Shiraz Univ., Shiraz (Iran, Islamic Republic of). Dept. of Nuclear Engineering, School of Mechanical Engineering; Faghihi, F. [Shiraz Univ., Shiraz (Iran, Islamic Republic of). Safety Research Center

    2010-07-01

    The loss of coolant flow in a nuclear reactor can result from a mechanical or electrical failure of the coolant pump. If the reactor is not tripped promptly, the immediate effect is a rapid increase in coolant temperature, decrease in minimum departure from nucleate boiling ratio (DNBR) and fuel damage. This study evaluated the shaft seizure of a reactor coolant pump in a VVER-1000 nuclear reactor. The locked rotor results in rapid reduction of flow through the affected reactor coolant loop and in turn leads to an increase in the primary coolant temperature and pressure. The analysis was conducted with regard for superimposing loss of power to the power plant at the initial accident moment. The required transient functions of flow, pressure and power were obtained using system transient calculations applied in COBRA-EN computer code in order to calculate the overall core thermal-hydraulic parameters such as temperature, critical heat flux and DNBR. The study showed that the critical period for the locked rotor accident is the first few seconds during which the maximum values of pressure and temperature are reached. 10 refs., 1 tab., 3 figs.

  10. Nuclear reactor melt arrest and coolability device

    Energy Technology Data Exchange (ETDEWEB)

    Theofanous, Theo G.; Dinh, Nam Truc; Wachowiak, Richard M.

    2016-06-14

    Example embodiments provide a Basemat-Internal Melt Arrest and Coolability device (BiMAC) that offers improved spatial and mechanical characteristics for use in damage prevention and risk mitigation in accident scenarios. Example embodiments may include a BiMAC having an inclination of less than 10-degrees from the basemat floor and/or coolant channels of less than 4 inches in diameter, while maintaining minimum safety margins required by the Nuclear Regulatory Commission.

  11. Applicability of base-isolation R D in non-reactor facilities to a nuclear reactor plant

    Energy Technology Data Exchange (ETDEWEB)

    Seidensticker, R.W.; Chang, Y.W.

    1990-01-01

    Seismic isolation is gaining increased attention worldwide for use in a wide spectrum of critical facilities, ranging from hospitals and computing centers to nuclear power plants. While the fundamental principles and technology are applicable to all of these facilities, the degree of assurance that the actual behavior of the isolation systems is as specified varies with the nature of the facility involved. Obviously, the level of effort to provide such assurance for a nuclear power plant will be much greater than that required for, say, a critical computer facility. The question, therefore, is to what extent can research and development (R D) for non-nuclear use be used to provide technological data needed for seismic isolation of a nuclear power plant. This question, of course is not unique to seismic isolation. Virtually every structural component, system, or piece of equipment used in nuclear power plants is also used in non- nuclear facilities. Experience shows that considerable effort is needed to adapt conventional technology into a nuclear power plant. Usually, more thorough analysis is required, material and fabrication quality-control requirements are more stringent as are controls on field installation. In addition, increased emphasis on maintainability and inservice inspection throughout the life of the plant is generally required to gain acceptance in nuclear power plant application. This paper reviews the R D programs ongoing for seismic isolation in non-nuclear facilities and related experience and makes a preliminary assessment of the extent to which such R D and experience can be used for nuclear power plant application. Ways are suggested to improve the usefulness of such non-nuclear R D in providing the high level of confidence required for the use of seismic isolation in a nuclear reactor plant. 2 refs.

  12. Installation of a new type of nuclear reactor in Mexico: advantages and disadvantages; Instalacion de un nuevo tipo de reactor nuclear en Mexico: ventajas y desventajas

    Energy Technology Data Exchange (ETDEWEB)

    Jurado P, M.; Martin del Campo M, C. [FI-UNAM, 04510 Mexico D.F. (Mexico)]. e-mail: mjp_green@hotmail.com

    2005-07-01

    In this work the main advantages and disadvantages of the installation of a new type of nuclear reactor different to the BWR type reactor in Mexico are presented. A revision of the advanced reactors is made that are at the moment in operation and of the advanced reactors that are in construction or one has already planned its construction in the short term. Specifically the A BWR and EPR reactors are analyzed. (Author)

  13. Facility Configuration Study of the High Temperature Gas-Cooled Reactor Component Test Facility

    Energy Technology Data Exchange (ETDEWEB)

    S. L. Austad; L. E. Guillen; D. S. Ferguson; B. L. Blakely; D. M. Pace; D. Lopez; J. D. Zolynski; B. L. Cowley; V. J. Balls; E.A. Harvego, P.E.; C.W. McKnight, P.E.; R.S. Stewart; B.D. Christensen

    2008-04-01

    A test facility, referred to as the High Temperature Gas-Cooled Reactor Component Test Facility or CTF, will be sited at Idaho National Laboratory for the purposes of supporting development of high temperature gas thermal-hydraulic technologies (helium, helium-Nitrogen, CO2, etc.) as applied in heat transport and heat transfer applications in High Temperature Gas-Cooled Reactors. Such applications include, but are not limited to: primary coolant; secondary coolant; intermediate, secondary, and tertiary heat transfer; and demonstration of processes requiring high temperatures such as hydrogen production. The facility will initially support completion of the Next Generation Nuclear Plant. It will secondarily be open for use by the full range of suppliers, end-users, facilitators, government laboratories, and others in the domestic and international community supporting the development and application of High Temperature Gas-Cooled Reactor technology. This pre-conceptual facility configuration study, which forms the basis for a cost estimate to support CTF scoping and planning, accomplishes the following objectives: • Identifies pre-conceptual design requirements • Develops test loop equipment schematics and layout • Identifies space allocations for each of the facility functions, as required • Develops a pre-conceptual site layout including transportation, parking and support structures, and railway systems • Identifies pre-conceptual utility and support system needs • Establishes pre-conceptual electrical one-line drawings and schedule for development of power needs.

  14. Parallelization and automatic data distribution for nuclear reactor simulations

    Energy Technology Data Exchange (ETDEWEB)

    Liebrock, L.M. [Liebrock-Hicks Research, Calumet, MI (United States)

    1997-07-01

    Detailed attempts at realistic nuclear reactor simulations currently take many times real time to execute on high performance workstations. Even the fastest sequential machine can not run these simulations fast enough to ensure that the best corrective measure is used during a nuclear accident to prevent a minor malfunction from becoming a major catastrophe. Since sequential computers have nearly reached the speed of light barrier, these simulations will have to be run in parallel to make significant improvements in speed. In physical reactor plants, parallelism abounds. Fluids flow, controls change, and reactions occur in parallel with only adjacent components directly affecting each other. These do not occur in the sequentialized manner, with global instantaneous effects, that is often used in simulators. Development of parallel algorithms that more closely approximate the real-world operation of a reactor may, in addition to speeding up the simulations, actually improve the accuracy and reliability of the predictions generated. Three types of parallel architecture (shared memory machines, distributed memory multicomputers, and distributed networks) are briefly reviewed as targets for parallelization of nuclear reactor simulation. Various parallelization models (loop-based model, shared memory model, functional model, data parallel model, and a combined functional and data parallel model) are discussed along with their advantages and disadvantages for nuclear reactor simulation. A variety of tools are introduced for each of the models. Emphasis is placed on the data parallel model as the primary focus for two-phase flow simulation. Tools to support data parallel programming for multiple component applications and special parallelization considerations are also discussed.

  15. Exploring Stochastic Sampling in Nuclear Data Uncertainties Assessment for Reactor Physics Applications and Validation Studies

    Directory of Open Access Journals (Sweden)

    Alexander Vasiliev

    2016-12-01

    Full Text Available The quantification of uncertainties of various calculation results, caused by the uncertainties associated with the input nuclear data, is a common task in nuclear reactor physics applications. Modern computation resources and improved knowledge on nuclear data allow nowadays to significantly advance the capabilities for practical investigations. Stochastic sampling is the method which has received recently a high momentum for its use and exploration in the domain of reactor design and safety analysis. An application of a stochastic sampling based tool towards nuclear reactor dosimetry studies is considered in the given paper with certain exemplary test evaluations. The stochastic sampling not only allows the input nuclear data uncertainties propagation through the calculations, but also an associated correlation analysis performance with no additional computation costs and for any parameters of interest can be done. Thus, an example of assessment of the Pearson correlation coefficients for several models, used in practical validation studies, is shown here. As a next step, the analysis of the obtained information is proposed for discussion, with focus on the systems similarities assessment. The benefits of the employed method and tools with respect to practical reactor dosimetry studies are consequently outlined.

  16. Transient behavior of a nuclear reactor coupled to an accelerator

    Science.gov (United States)

    Sadineni, Suresh Babu

    Accelerator Driven Systems (ADS) present one of the most viable solutions for transmutation and effective utilization of nuclear fuel. Spent fuel from reactors will be partitioned to separate plutonium and other minor actinides to be transmuted in the ADS. Without the ADS, minor actinides must be stored at a geologic repository for long periods of time. One problem with ADS is understanding the control issues that arise when coupling an accelerator to a reactor. "ADSTRANS" was developed to predict the transient behavior of a nuclear reactor coupled to an accelerator. It was based on MCNPX, a radiation transport code developed at the LANL, and upon a numerical model of the neutron transport equation. MCNPX was used to generate the neutron "source" term that occurs when the accelerator is fired. ADSTRANS coupled MCNPX to a separate finite difference code that solved the transient neutron transport equation. A cylindrical axisymmetric reactor with steel shielding was considered for this analysis. Multiple neutron energy groups, neutron precursor groups and neutron poisons were considered. ENDF/B cross-section data obtained through MCNPX was also employed. The reactor was assumed to be isothermal and near zero power level. Unique features of this code are: (1) it predicts the neutron behavior of an ADS for different reactor geometry, material concentration, both electron and proton particle accelerators, and target material, (2) it develops input files for MCNPX to simulate neutron production, runs MCNPX, and retrieves information from the MCNPX output files. Neutron production predicted by MCNPX for a 20 MeV electron accelerator and lead target was compared with experimental data from the Idaho Accelerator Center and found to be in good agreement. The spatial neutron flux distribution and transient neutron flux in the reactor as predicted by the code were compared with analytical solutions and found to be in good agreement. Fuel burnup and poison buildup were also as

  17. Basic Model of a Control Assembly Drop in Nuclear Reactors

    Directory of Open Access Journals (Sweden)

    Radek BULÍN

    2013-06-01

    Full Text Available This paper is focused on the modelling and dynamic analysis of a nonlinear system representing a control assembly of the VVER 440/V213 nuclear reactor. A simple rigid body model intended for basic dynamic analyses is introduced. It contains the influences of the pressurized water and mainly the eects of possible control assembly contacts with guiding tubes inside the reactor. Another approach based on a complex multibody model is further described and the suitability of both modelling approaches is discussed.

  18. Nuclear reactor for breeding U.sup.233

    Science.gov (United States)

    Bohanan, Charles S.; Jones, David H.; Raab, Jr., Harry F.; Radkowsky, Alvin

    1976-01-01

    A light-water-cooled nuclear reactor capable of breeding U.sup.233 for use in a light-water breeder reactor includes physically separated regions containing U.sup.235 fissile material and U.sup.238 fertile material and Th.sup.232 fertile material and Pu.sup.239 fissile material, if available. Preferably the U.sup.235 fissile material and U.sup.238 fertile material are contained in longitudinally movable seed regions and the Pu.sup.239 fissile material and Th.sup.232 fertile material are contained in blanket regions surrounding the seed regions.

  19. Non-destructive-Testing of Nuclear Fuel Element by Means of Neutron Imaging Technique

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

    Nuclear fuel element is the key component of nuclear reactor. People have to make strictly testing of the element to make sure the reactor operating safely. Neutron imaging is one of Non-destructive-Testing (NDT) techniques, which are very important techniques for

  20. Thermohydraulic and nuclear modeling of natural fission reactors

    Science.gov (United States)

    Viggato, Jason Charles

    Experimental verification of proposed nuclear waste storage schemes in geologic repositories is not possible, however, a natural analog exists in the form of ancient natural reactors that existed in uranium-rich ores. Two billion years ago, the enrichment of natural uranium was high enough to allow a sustained chain reaction in the presence of water as a moderator. Several natural reactors occurred in Gabon, Africa and were discovered in the early 1970's. These reactors operated at low power levels for hundreds of thousands of years. Heated water generated from the reactors also leached uranium from the surrounding rock strata and deposited it in the reactor cores. This increased the concentration of uranium in the core over time and served to "refuel" the reactor. This has strong implications in the design of modern geologic repositories for spent nuclear fuel. The possibility of accidental fission events in man-made repositories exists and the geologic evidence from Oklo suggests how those events may progress and enhance local concentrations of uranium. Based on a review of the literature, a comprehensive code was developed to model the thermohydraulic behavior and criticality conditions that may have existed in the Oklo reactor core. A two-dimensional numerical model that incorporates modeling of fluid flow, temperatures, and nuclear fission and subsequent heat generation was developed for the Oklo natural reactors. The operating temperatures ranged from about 456 K to about 721 K. Critical reactions were observed for a wide range of concentrations and porosity values (9 to 30 percent UO2 and 10 to 20 percent porosity). Periodic operation occurred in the computer model prediction with UO2 concentrations of 30 percent in the core and 5 percent in the surrounding material. For saturated conditions and 30 percent porosity, the model predicted temperature transients with a period of about 5 hours. Kuroda predicted 3 to 4 hour durations for temperature transients

  1. Nuclear Thermal Propulsion Ground Test History

    Science.gov (United States)

    Gerrish, Harold P.

    2014-01-01

    Nuclear Thermal Propulsion (NTP) was started in 1955 under the Atomic Energy Commission as project Rover and was assigned to Los Alamos National Laboratory. The Nevada Test Site was selected in 1956 and facility construction began in 1957. The KIWI-A was tested on July 1, 1959 for 5 minutes at 70MW. KIWI-A1 was tested on July 8, 1960 for 6 minutes at 85MW. KIWI-A3 was tested on October 10, 1960 for 5 minutes at 100MW. The National Aeronautics and Space Administration (NASA) was formed in 1958. On August 31, 1960 the AEC and NASA established the Space Nuclear Propulsion Office and named Harold Finger as Director. Immediately following the formation of SNPO, contracts were awarded for the Reactor In Flight Test (RIFT), master plan for the Nuclear Rocket Engine Development Station (NRDS), and the Nuclear Engine for Rocket Vehicle Application (NERVA). From December 7, 1961 to November 30, 1962, the KIWI-B1A, KIWI-B1B, and KIWI-B4A were tested at test cell A. The last two engines were only tested for several seconds before noticeable failure of the fuel elements. Harold Finger called a stop to any further hot fire testing until the problem was well understood. The KIWI-B4A cold flow test showed the problem to be related to fluid dynamics of hydrogen interstitial flow causing fuel element vibrations. President Kennedy visited the NTS one week after the KIWI-B4A failure and got to see the engine starting to be disassembled in the maintenance facility. The KIWI-B4D and KIWI-B4E were modified to not have the vibration problems and were tested in test cell C. The NERVA NRX program started testing in early 1964 with NRX-A1 cold flow test series (unfueled graphite core), NRX-A2 and NRX-A3 power test series up to 1122 MW for 13 minutes. In March 1966, the NRX-EST (Engine System Test) was the first breadboard using flight functional relationship and total operating time of 116 minutes. The NRX-EST demonstrated the feasibility of a hot bleed cycle. The NRX-A5 had multiple start

  2. Instrumentation Requirements for the Engineering Evaluation of Nuclear-Electric Spacecraft

    Science.gov (United States)

    Apel, W. C.

    1961-01-01

    Spacecraft employing nuclear-electric propulsion are being proposed for missions to Venus and distances beyond. These spacecraft utilize a nuclear reactor to provide thermal energy to a turboalternator which generates electric power for an ion motor and the other spacecraft systems. This Report discusses the instrumentation and communications system needed to evaluate a nuclear-electric spacecraft in flight, along with the problems expected. A representative spacecraft design is presented, which leads to a discussion of the instrumentation needed to evaluate such a spacecraft. A basic communications system is considered for transmitting the spacecraft data to Earth. The instrumentation and communications system, as well as all electronic systems on a nuclear-electric spacecraft, will be operating in high temperature and nuclear-radiation environments. The problems caused by these environments are discussed, and possible solutions are offered.

  3. Instrumentation Requirements for the Engineering Evaluation of Nuclear-Electric Spacecraft

    Science.gov (United States)

    Apel, W. C.

    1961-01-01

    Spacecraft employing nuclear-electric propulsion are being proposed for missions to Venus and distances beyond. These spacecraft utilize a nuclear reactor to provide thermal energy to a turboalternator which generates electric power for an ion motor and the other spacecraft systems. This Report discusses the instrumentation and communications system needed to evaluate a nuclear-electric spacecraft in flight, along with the problems expected. A representative spacecraft design is presented, which leads to a discussion of the instrumentation needed to evaluate such a spacecraft. A basic communications system is considered for transmitting the spacecraft data to Earth. The instrumentation and communications system, as well as all electronic systems on a nuclear-electric spacecraft, will be operating in high temperature and nuclear-radiation environments. The problems caused by these environments are discussed, and possible solutions are offered.

  4. Study of fuel assemblies for the nuclear reactor GFR; Estudio de ensambles de combustible para el reactor nuclear GFR

    Energy Technology Data Exchange (ETDEWEB)

    Reyes R, R.; Martin del Campo M, C.; Francois L, J. L. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac 8532, Jiutepec, Morelos 62550 (Mexico)]. e-mail: ricarera@yahoo.com.mx

    2008-07-01

    In the present work the criticality calculations for two models of fuel assembly were realized to study the nuclear reactor cooled by gas (Gas Fast Reactor) of IV Generation. Model 1 is an assembly with hexagonal adjustment of fuel rods with reflector in the axial ends higher and lower, the coolant flows between the rods. Model 2 is an hexagonal assembly type block with spheres dispersion and cylindrical channels for where the coolant with reflector in the axial ends also flows. The materials selected for each component of the assemblies, should be resistant to the radiation of fast neutrons and high operation temperatures, for what in both models the following materials were chosen: a mixture of uranium carbide more plutonium for the fuel; a mixture of silicon carbide in different theoretical density percentages for structures and shieldings; helium gas like coolant and a zirconium carbide mixture like reflector, which fulfill the restrictions of being resistant to the high operation temperatures and means of irradiation. General considerations were taken, which are common parameters to both types of assemblies, like size and materials used in the different parts of each model of assembly. The criticality calculations were obtained with the help of the MCNPx code, based on the Monte Carlo method. It was realized a validation of the atomic density data of each component of the assemblies, to have the certainty of the proportionate values that they were introduced of correct way in the code. The results show that model 1 makes better use of the fissile material in a assembly that has the same dimensions externally. That is to say, that from the only considered viewpoint, the neutron one, model 1 is better than model 2. (Author)

  5. Acrylic acid and electric power cogeneration in an SOFC reactor.

    Science.gov (United States)

    Ji, Baofeng; Wang, Jibo; Chu, Wenling; Yang, Weishen; Lin, Liwu

    2009-04-21

    A highly efficient catalyst, MoV(0.3)Te(0.17)Nb(0.12)O, used for acrylic acid (AA) production from propane, was used as an anodic catalyst in an SOFC reactor, from which AA and electric power were cogenerated at 400-450 degrees C.

  6. Nuclear data uncertainty quantification and data assimilation for a lead-cooled fast reactor : Using integral experiments for improved accuracy

    OpenAIRE

    Alhassan, Erwin

    2015-01-01

    For the successful deployment of advanced nuclear systems and optimization of current reactor designs, high quality nuclear data are required. Before nuclear data can be used in applications they must first be evaluated, tested and validated against a set of integral experiments, and then converted into formats usable for applications. The evaluation process in the past was usually done by using differential experimental data which was then complemented with nuclear model calculations. This t...

  7. Requirements for advanced simulation of nuclear reactor and chemicalseparation plants.

    Energy Technology Data Exchange (ETDEWEB)

    Palmiotti, G.; Cahalan, J.; Pfeiffer, P.; Sofu, T.; Taiwo, T.; Wei,T.; Yacout, A.; Yang, W.; Siegel, A.; Insepov, Z.; Anitescu, M.; Hovland,P.; Pereira, C.; Regalbuto, M.; Copple, J.; Willamson, M.

    2006-12-11

    This report presents requirements for advanced simulation of nuclear reactor and chemical processing plants that are of interest to the Global Nuclear Energy Partnership (GNEP) initiative. Justification for advanced simulation and some examples of grand challenges that will benefit from it are provided. An integrated software tool that has its main components, whenever possible based on first principles, is proposed as possible future approach for dealing with the complex problems linked to the simulation of nuclear reactor and chemical processing plants. The main benefits that are associated with a better integrated simulation have been identified as: a reduction of design margins, a decrease of the number of experiments in support of the design process, a shortening of the developmental design cycle, and a better understanding of the physical phenomena and the related underlying fundamental processes. For each component of the proposed integrated software tool, background information, functional requirements, current tools and approach, and proposed future approaches have been provided. Whenever possible, current uncertainties have been quoted and existing limitations have been presented. Desired target accuracies with associated benefits to the different aspects of the nuclear reactor and chemical processing plants were also given. In many cases the possible gains associated with a better simulation have been identified, quantified, and translated into economical benefits.

  8. Synergistic smart fuel for in-pile nuclear reactor measurements

    Energy Technology Data Exchange (ETDEWEB)

    Smith, J.A.; Kotter, D.K. [Idaho National Laboratories, Idaho Falls (United States); Ali, R.A.; Garrett, S.L. [Penn State University, University Park, State College, PA 16801 (United States)

    2013-07-01

    The thermo-acoustic fuel rod sensor developed in this research has demonstrated a novel technique for monitoring the temperature within the core of a nuclear reactor or the temperature of the surrounding heat-transfer fluid. It uses the heat from the nuclear fuel to generate sustained acoustic oscillations whose frequency will be indicative of the temperature. Converting a nuclear fuel rod into this type of thermo-acoustic sensor simply requires the insertion of a porous material (stack). This sensor has demonstrated a synergy with the elevated temperatures that exist within the nuclear reactor using materials that have only minimal susceptibility to high-energy particle fluxes. When the sensor is in operation, the sound waves radiated from the fuel rod resonator will propagate through the surrounding cooling fluid. The frequency of these oscillations is directly correlated with an effective temperature within the fuel rod resonator. This device is self-powered and is operational even in case of total loss of power of the reactor.

  9. Optimization Algorithms for Nuclear Reactor Power Control

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yeong Min; Oh, Won Jong; Oh, Seung Jin; Chun, Won Gee; Lee, Yoon Joon [Jeju National University, Jeju (Korea, Republic of)

    2010-10-15

    One of the control techniques that could replace the present conventional PID controllers in nuclear plants is the linear quadratic regulator (LQR) method. The most attractive feature of the LQR method is that it can provide the systematic environments for the control design. However, the LQR approach heavily depends on the selection of cost function and the determination of the suitable weighting matrices of cost function is not an easy task, particularly when the system order is high. The purpose of this paper is to develop an efficient and reliable algorithm that could optimize the weighting matrices of the LQR system

  10. The Harnessed Atom: Nuclear Energy & Electricity.

    Science.gov (United States)

    Department of Energy, Washington, DC. Nuclear Energy Office.

    This document is part of a nuclear energy curriculum designed for grades six through eight. The complete kit includes a written text, review exercises, activities for the students, and a teachers guide. The 19 lessons in the curriculum are divided into four units including: (1) "Energy and Electricity"; (2) "Understanding Atoms and Radiation"; (3)…

  11. Application of probabilistic risk assessment in nuclear and environmental licensing processes of nuclear reactors in Brazil

    Energy Technology Data Exchange (ETDEWEB)

    Mata, Jonatas F.C. da; Vasconcelos, Vanderley de; Mesquita, Amir Z., E-mail: jonatasfmata@yahoo.com.br, E-mail: vasconv@cdtn.br, E-mail: amir@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    The nuclear accident at Fukushima Daiichi, occurred in Japan in 2011, brought reflections, worldwide, on the management of nuclear and environmental licensing processes of existing nuclear reactors. One of the key lessons learned in this matter, is that the studies of Probabilistic Safety Assessment and Severe Accidents are becoming essential, even in the early stage of a nuclear development project. In Brazil, Brazilian Nuclear Energy Commission, CNEN, conducts the nuclear licensing. The organism responsible for the environmental licensing is Brazilian Institute of Environment and Renewable Natural Resources, IBAMA. In the scope of the licensing processes of these two institutions, the safety analysis is essentially deterministic, complemented by probabilistic studies. The Probabilistic Safety Assessment (PSA) is the study performed to evaluate the behavior of the nuclear reactor in a sequence of events that may lead to the melting of its core. It includes both probability and consequence estimation of these events, which are called Severe Accidents, allowing to obtain the risk assessment of the plant. Thus, the possible shortcomings in the design of systems are identified, providing basis for safety assessment and improving safety. During the environmental licensing, a Quantitative Risk Analysis (QRA), including probabilistic evaluations, is required in order to support the development of the Risk Analysis Study, the Risk Management Program and the Emergency Plan. This article aims to provide an overview of probabilistic risk assessment methodologies and their applications in nuclear and environmental licensing processes of nuclear reactors in Brazil. (author)

  12. Development of advanced automatic control system for nuclear ship. 2. Perfect automatic operation after reactor scram events

    Energy Technology Data Exchange (ETDEWEB)

    Yabuuchi, Noriaki; Nakazawa, Toshio; Takahashi, Hiroki; Shimazaki, Junya; Hoshi, Tsutao [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1997-11-01

    An automatic operation system has been developed for the purpose of realizing a perfect automatic plant operation after reactor scram events. The goal of the automatic operation after a reactor scram event is to bring the reactor hot stand-by condition automatically. The basic functions of this system are as follows; to monitor actions of the equipments of safety actions after a reactor scram, to control necessary control equipments to bring a reactor to a hot stand-by condition automatically, and to energize a decay heat removal system. The performance evaluation on this system was carried out by comparing the results using to Nuclear Ship Engineering Simulation System (NESSY) and the those measured in the scram test of the nuclear ship `Mutsu`. As the result, it was showed that this system had the sufficient performance to bring a reactor to a hot syand-by condition quickly and safety. (author)

  13. The nuclear battery: a very small reactor power supply for remote locations

    Energy Technology Data Exchange (ETDEWEB)

    Kozier, K.S. (Atomic Energy of Canada Ltd., Pinawa, MB (Canada). Whiteshell Nuclear Research Establishment)

    The Nuclear Battery is a small reactor power supply being developed by Atomic Energy of Canada Limited for use in locations that are remote from utility grids and natural gas pipelines. Key technical features of the Nuclear Battery reactor core include a heat-pipe primary heat transport system, graphite neutron moderator, low-enriched uranium TRISO coated-particle fuel, and the use of burnable poisons for long-term reactivity control. An external secondary heat transport system extracts useful heat energy that may be converted into electricity in an organic Rankine cycle engine, or used to produce a high-pressure steam. The reference design is capable of producing about 2400 kW(t) (about 600 kW(e) net) for 15 full-power years without refueling. (author).

  14. The nuclear battery: a very small reactor power supply for remote locations; Technical note

    Energy Technology Data Exchange (ETDEWEB)

    Kozier, K.S. (Atomic Energy of Canada Ltd., Pinawa, MB (Canada). Whiteshell Labs.)

    1992-08-01

    The Nuclear Battery is a small reactor power supply being developed by Atomic Energy of Canada Limited for use in locations that are remote from utility grids and natural gas pipelines. Key technical features of the Nuclear Battery reactor core include a heat-pipe primary heat transport system, graphite neutron moderator, low enriched uranium TRISO coated-particle fuel and the use of burnable poisons for long-term reactivity control. An external secondary heat transport system extracts useful heat energy that may be converted into electricity in an organic Rankine cycle engine, or used to produce high-pressure steam. The reference design is capable of producing about 2400 kW(t) (about 500 kW(e) net) for 15 full-power years without refuelling. (orig.).

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-04-01

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

  16. Sub-Critical Nuclear Reactor Based on FFAG-Accelerator

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Hee Seok; Kang, Hung Sik; Lee, Tae Yeon [Pohang Accelerator Laboratory, Pohang (Korea, Republic of)

    2011-10-15

    After the East-Japan earthquake and the subsequent nuclear disaster, the anti-nuclear mood has been wide spread. It is very unfortunate both for nuclear science community and for the future of mankind, which is threatened by two serious challenges, the global warming caused by the greenhouse effect and the shortage of energy cause by the petroleum exhaustion. While the nuclear energy seemed to be the only solution to these problems, it is clear that it has its own problems, one of which broke out so strikingly in Japan. There are also other problems such as the radiotoxic nuclear wastes that survive up to even tens of thousands years and the limited reserves of Uranium. To solve these problems of nuclear fission energy, accelerator-based sub-critical nuclear reactor was once proposed. (Its details will be explained below.) First of all, it is safe in a disaster such as an earthquake, because the deriving accelerator stops immediately by the earthquake. It also minimizes the nuclear waste problem by reducing the amount of the toxic waste and shortening their half lifetime to only a few hundred years. Finally, it solves the Uranium reserve problem because it can use Thorium as its fuel. The Thorium reserve is much larger than that of Uranium. Although the idea of the accelerator-driven nuclear reactor was proposed long time ago, it has not been utilized yet first by technical difficulty and economical reasons. The accelerator-based system needs 1 GeV, 10 MW power proton accelerator. A conventional linear accelerator would need several hundred m length, which is highly costly particularly in Korea because of the high land cost. However, recent technologies make it possible to realize that scale accelerator by a reasonable size. That is the fixed-field alternating gradient (FFAG) accelerator that is described in this article

  17. High Efficiency Nuclear Power Plants Using Liquid Fluoride Thorium Reactor Technology

    Science.gov (United States)

    Juhasz, Albert J.; Rarick, Richard A.; Rangarajan, Rajmohan

    2009-01-01

    An overall system analysis approach is used to propose potential conceptual designs of advanced terrestrial nuclear power plants based on Oak Ridge National Laboratory (ORNL) Molten Salt Reactor (MSR) experience and utilizing Closed Cycle Gas Turbine (CCGT) thermal-to-electric energy conversion technology. In particular conceptual designs for an advanced 1 GWe power plant with turbine reheat and compressor intercooling at a 950 K turbine inlet temperature (TIT), as well as near term 100 MWe demonstration plants with TITs of 950 and 1200 K are presented. Power plant performance data were obtained for TITs ranging from 650 to 1300 K by use of a Closed Brayton Cycle (CBC) systems code which considered the interaction between major sub-systems, including the Liquid Fluoride Thorium Reactor (LFTR), heat source and heat sink heat exchangers, turbo-generator machinery, and an electric power generation and transmission system. Optional off-shore submarine installation of the power plant is a major consideration.

  18. Nuclear Waste Transmutation in Subcritical Reactors Driven by Target-Distributed Accelerators

    CERN Document Server

    Blanovsky, A

    2004-01-01

    A radioactive waste transmutation system based extensively on existing nuclear power technology is presented. By replacing the control rods with neutron sources, we could maintain good power distribution and perform long-lived waste burning in high flux subcritical reactors. The design is based on a small pressurized water reactor, fission electric cell (FEC), target-distributed accelerator (TDA) and power monitoring system with in-core gamma-ray detectors, now under development in several countries. The TDA, in which an FEC electric field compensates for lost beam energy in the target, offers a new approach to obtain large neutron fluxes. The analysis takes into consideration a wide range of TDA design aspects including the wave model of observed relativistic phenomena, in-core microwave power source, the FEC with a multistage collector (anode) and layered cathode.

  19. "What--me worry?" "Why so serious?": a personal view on the Fukushima nuclear reactor accidents.

    Science.gov (United States)

    Gallucci, Raymond

    2012-09-01

    Infrequently, it seems that a significant accident precursor or, worse, an actual accident, involving a commercial nuclear power reactor occurs to remind us of the need to reexamine the safety of this important electrical power technology from a risk perspective. Twenty-five years since the major core damage accident at Chernobyl in the Ukraine, the Fukushima reactor complex in Japan experienced multiple core damages as a result of an earthquake-induced tsunami beyond either the earthquake or tsunami design basis for the site. Although the tsunami itself killed tens of thousands of people and left the area devastated and virtually uninhabitable, much concern still arose from the potential radioactive releases from the damaged reactors, even though there was little population left in the area to be affected. As a lifelong probabilistic safety analyst in nuclear engineering, even I must admit to a recurrence of the doubt regarding nuclear power safety after Fukushima that I had experienced after Three Mile Island and Chernobyl. This article is my attempt to "recover" my personal perspective on acceptable risk by examining both the domestic and worldwide history of commercial nuclear power plant accidents and attempting to quantify the risk in terms of the frequency of core damage that one might glean from a review of operational history.

  20. Detectability prediction for a thermoacoustic sensor in the breazeale nuclear reactor pool

    Energy Technology Data Exchange (ETDEWEB)

    Smith, James [Idaho National Laboratory, Idaho Falls, ID (United States); Hrisko, Joshua [Idaho National Laboratory, Idaho Falls, ID (United States); Garrett, Steven [Idaho National Laboratory, Idaho Falls, ID (United States)

    2016-03-01

    Laboratory experiments have suggested that thermoacoustic engines can be in- corporated within nuclear fuel rods. Such engines would radiate sounds that could be used to measure and acoustically-telemeter information about the op- eration of the nuclear reactor (e.g., coolant temperature or uxes of neutrons or other energetic particles) or the physical condition of the nuclear fuel itself (e.g., changes in temperature, evolved gases) that are encoded as the frequency and/or amplitude of the radiated sound [IEEE Measurement and Instrumen- tation 16(3), 18-25 (2013)]. For such acoustic information to be detectable, it is important to characterize the vibroacoustical environments within reactors. Measurements will be presented of the background noise spectra (with and with- out coolant pumps) and reverberation times within the 70,000 gallon pool that cools and shields the fuel in the 1 MW research reactor on Penn State's campus using two hydrophones, a piezoelectric projector, and an accelerometer. Sev- eral signal-processing techniques will be demonstrated to enhance the measured results. Background vibrational measurement were also taken at the 250 MW Advanced Test Reactor, located at the Idaho National Laboratory, using ac- celerometers mounted outside the reactor's pressure vessel and on plumbing will also be presented. The detectability predictions made in the thesis were validated in September 2015 using a nuclear ssion-heated thermoacoustic sensor that was placed in the core of the Breazeale Nuclear Reactor on Penn State's campus. Some features of the thermoacoustic device used in that experiment will also be revealed. [Work supported by the U.S. Department of Energy.

  1. FLOWSHEET EVALUATION FOR THE DISSOLVING AND NEUTRALIZATION OF SODIUM REACTOR EXPERIMENT USED NUCLEAR FUEL

    Energy Technology Data Exchange (ETDEWEB)

    Daniel, W. E.; Hansen, E. K.; Shehee, T. C.

    2012-10-30

    This report includes the literature review, hydrogen off-gas calculations, and hydrogen generation tests to determine that H-Canyon can safely dissolve the Sodium Reactor Experiment (SRE; thorium fuel), Ford Nuclear Reactor (FNR; aluminum alloy fuel), and Denmark Reactor (DR-3; silicide fuel, aluminum alloy fuel, and aluminum oxide fuel) assemblies in the L-Bundles with respect to the hydrogen levels in the projected peak off-gas rates. This is provided that the number of L-Bundles charged to the dissolver is controlled. Examination of SRE dissolution for potential issues has aided in predicting the optimal batching scenario. The calculations detailed in this report demonstrate that the FNR, SRE, and DR-3 used nuclear fuel (UNF) are bounded by MURR UNF and may be charged using the controls outlined for MURR dissolution in a prior report.

  2. Insight on the inconsistencies of Barkhausen signal measurements for radiation damage on nuclear reactor steel

    Energy Technology Data Exchange (ETDEWEB)

    Barroso, Soraia Pirfo; Fitzpatrick, Michael E. [Materials Engineering, The Open University, Walton Hall, Milton Keynes, MK7 6AA (United Kingdom); Gillemot, Ferenc; Horváth, Marta; Horváth, Ákos; Szekely, Richard [Hungarian Academy of Sciences Centre for Energy Research (MTA EK), P.O. Box 49 H-1525, Budapest 114 (Hungary)

    2014-02-18

    This paper focuses on the use of magnetic measurements, using Barkhausen signals to determine the irradiation effects, attempting to predict fracture toughness changes on nuclear reactor structural materials and correlating these measurements to mechanical testing and microstructure. For this study, two types of nuclear reactor materials were investigated: one sensitive to irradiation effects, the JRQ IAEA's reference material (A533B- -type); and one resistant material, 15KH2MFA WWER's reactor pressure vessel steel. The samples were carefully identified within the original heat block, i.e. forged or rolled plate. These calibrated samples were irradiated at different neutron fluences up to 10{sup 23} n/m{sup 2}. We show how microstructural anisotropy can mask the irradiation effects in the magnetic measurements. A correlation between irradiation effects and the magnetic measurements is explained based on this study.

  3. Software reliability and safety in nuclear reactor protection systems

    Energy Technology Data Exchange (ETDEWEB)

    Lawrence, J.D. [Lawrence Livermore National Lab., CA (United States)

    1993-11-01

    Planning the development, use and regulation of computer systems in nuclear reactor protection systems in such a way as to enhance reliability and safety is a complex issue. This report is one of a series of reports from the Computer Safety and Reliability Group, Lawrence Livermore that investigates different aspects of computer software in reactor National Laboratory, that investigates different aspects of computer software in reactor protection systems. There are two central themes in the report, First, software considerations cannot be fully understood in isolation from computer hardware and application considerations. Second, the process of engineering reliability and safety into a computer system requires activities to be carried out throughout the software life cycle. The report discusses the many activities that can be carried out during the software life cycle to improve the safety and reliability of the resulting product. The viewpoint is primarily that of the assessor, or auditor.

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

    Science.gov (United States)

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

    2012-06-01

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

  5. Neutron analysis of the fuel of high temperature nuclear reactors; Analisis neutronico del combustible de reactores nucleares de alta temperatura

    Energy Technology Data Exchange (ETDEWEB)

    Bastida O, G. E.; Francois L, J. L., E-mail: gbo729@yahoo.com.mx [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico)

    2014-10-15

    In this work a neutron analysis of the fuel of some high temperature nuclear reactors is presented, studying its main features, besides some alternatives of compound fuel by uranium and plutonium, and of coolant: sodium and helium. For this study was necessary the use of a code able to carry out a reliable calculation of the main parameters of the fuel. The use of the Monte Carlo method was convenient to simulate the neutrons transport in the reactor core, which is the base of the Serpent code, with which the calculations will be made for the analysis. (Author)

  6. The opening of electricity markets: the stakes for the nuclear energy; L'ouverture des marches de l'electricite: les enjeux pour le nucleaire

    Energy Technology Data Exchange (ETDEWEB)

    Maillard, M. [Ministere de l' Economie, des Finances et de l' Industrie, 75 - Paris (France). Direction Generale de L' Energie et des Matieres Premieres

    2002-07-01

    This article is a reprint of the opening talk of M. Maillard, general director of DGEMP (general direction of energy and raw materials, French ministry of economy, finances and industry), given at the 2002 annual meeting of the French society of nuclear energy (SFEN). In his talk, M. Maillard presents, first, the international context of the liberalization of electricity markets (the lessons learnt after the Californian crisis, the situation of the existing nuclear park, the questions relative to the nuclear revival, the conditions of a conservation of the nuclear option beyond the existing facilities). Then he analyzes the French situation and its specificities (pluri-annual planning of investments, mastered liberalization and competitiveness of the nuclear energy, preparation of future date-lines in agreement with the decommissioning of the oldest PWR reactors). (J.S.)

  7. ABWR Nuclear Power Plant : Safety nuclear electricity and low cost; La Central Nuclear ABWR: Electricidad de origen nuclear segura y de costa previsible

    Energy Technology Data Exchange (ETDEWEB)

    Redding, J.; Segarr, J.

    2002-07-01

    The Advanced Boiling Water Reactor (ABWR) nuclear plant defines the new generation of advanced nuclear plant designs that have achieved higher levels of safety and can satisfy the increasing needs for low cost electricity. Two ABWRs in Japan are now in their fifth cycles of operation. By every measure, these units have proven themselves successful. The construction of two more ABWRs in Taiwan at the Lungmen site is once again proceeding smoothly. Attention is now being focused on europe where the ABWR has been adapted to meet European safety requirements through utility sponsored programs. (Author)

  8. Review of Transient Testing of Fast Reactor Fuels in the Transient REActor Test Facility (TREAT)

    Energy Technology Data Exchange (ETDEWEB)

    Jensen, C.; Wachs, D.; Carmack, J.; Woolstenhulme, N.

    2017-01-01

    The restart of the Transient REActor Test (TREAT) facility provides a unique opportunity to engage the fast reactor fuels community to reinitiate in-pile experimental safety studies. Historically, the TREAT facility played a critical role in characterizing the behavior of both metal and oxide fast reactor fuels under off-normal conditions, irradiating hundreds of fuel pins to support fast reactor fuel development programs. The resulting test data has provided validation for a multitude of fuel performance and severe accident analysis computer codes. This paper will provide a review of the historical database of TREAT experiments including experiment design, instrumentation, test objectives, and salient findings. Additionally, the paper will provide an introduction to the current and future experiment plans of the U.S. transient testing program at TREAT.

  9. First measurements of (236)U concentrations and (236)U/(239)Pu isotopic ratios in a Southern Hemisphere soil far from nuclear test or reactor sites.

    Science.gov (United States)

    Srncik, M; Tims, S G; De Cesare, M; Fifield, L K

    2014-06-01

    The variation of the (236)U and (239)Pu concentrations as a function of depth has been studied in a soil profile at a site in the Southern Hemisphere well removed from nuclear weapon test sites. Total inventories of (236)U and (239)Pu as well as the (236)U/(239)Pu isotopic ratio were derived. For this investigation a soil core from an undisturbed forest area in the Herbert River catchment (17°30' - 19°S) which is located in north-eastern Queensland (Australia) was chosen. The chemical separation of U and Pu was carried out with a double column which has the advantage of the extraction of both elements from a relatively large soil sample (∼20 g) within a day. The samples were measured by Accelerator Mass Spectrometry using the 14UD pelletron accelerator at the Australian National University. The highest atom concentrations of both (236)U and (239)Pu were found at a depth of 2-3 cm. The (236)U/(239)Pu isotopic ratio in fallout at this site, as deduced from the ratio of the (236)U and (239)Pu inventories, is 0.085 ± 0.003 which is clearly lower than the Northern Hemisphere value of ∼0.2. The (236)U inventory of (8.4 ± 0.3) × 10(11) at/m(2) was more than an order of magnitude lower than values reported for the Northern Hemisphere. The (239)Pu activity concentrations are in excellent agreement with a previous study and the (239+240)Pu inventory was (13.85 ± 0.29) Bq/m(2). The weighted mean (240)Pu/(239)Pu isotopic ratio of 0.142 ± 0.005 is slightly lower than the value for global fallout, but our results are consistent with the average ratio of 0.173 ± 0.027 for the southern equatorial region (0-30°S).

  10. Evaluation of High-Performance Space Nuclear Electric Generators for Electric Propulsion Application

    Science.gov (United States)

    Woodcock, Gordon; Kross, Dennis A. (Technical Monitor)

    2002-01-01

    Electric propulsion applications are enhanced by high power-to-mass ratios for their electric power sources. At multi-megawatt levels, we can expect thrust production systems to be less than 5 kg/kWe. Application of nuclear electric propulsion to human Mars missions becomes an attractive alternative to nuclear thermal propulsion if the propulsion system is less than about 10 kg/kWe. Recent references have projected megawatt-plus nuclear electric sources at specific mass values from less than 1 kg/kWe to about 5 kg/kWe. Various assumptions are made regarding power generation cycle (turbogenerator; MHD (magnetohydrodynamics)) and reactor heat source design. The present paper compares heat source and power generation options on the basis of a parametric model that emphasizes heat transfer design and realizable hardware concept. Pressure drop (important!) is included in the power cycle analysis, and MHD and turbogenerator cycles are compared. Results indicate that power source specific mass less than 5 kg/kWe is attainable, even if peak temperatures achievable are limited to 1500 K. Projections of specific mass less than 1 kg/kWe are unrealistic, even at the highest peak temperatures considered.

  11. SAFT inspections for developing empirical database of fabrication flaws in nuclear reactor pressure vessels

    Science.gov (United States)

    Doctor, Steven R.; Schuster, George J.; Pardini, Allan F.

    1998-03-01

    The Pacific Northwest National Laboratory (PNNL) is developing a methodology for estimating the size and density distribution of fabrication flaws in U.S. nuclear reactor pressure vessels. This involves the nondestructive evaluation (NDE) of reactor pressure vessel materials and the destructive validation of the flaws found. NDE has been performed on reactor pressure vessel material made by Babcock & Wilcox and Combustion Engineering. A metallographic analysis is being performed to validate the flaw density and size distributions estimated from the 2500 indications of fabrication flaws that were detected and characterized in the very sensitive SAFT-UT (synthetic aperture focusing technique for ultrasonic testing) inspection data from the Pressure Vessel Research User Facility (PVRUF) vessel at Oak Ridge National Laboratory. Research plans are also described for expanding the work to include other reactor pressure vessel materials.

  12. 10 CFR Appendix A to Part 110 - Illustrative List of Nuclear Reactor Equipment Under NRC Export Licensing Authority

    Science.gov (United States)

    2010-01-01

    ... designed for inserting or removing fuel in an operating nuclear reactor. (3) Complete reactor control rod... contain fuel elements and the primary coolant in a nuclear reactor at an operating pressure in excess of... diffuser plates especially designed or prepared for use in a nuclear reactor. (8) Reactor control......

  13. 76 FR 31381 - Office Of New Reactors; Proposed Revision 4 to Standard Review Plan; Section 8.1 on Electric...

    Science.gov (United States)

    2011-05-31

    ... Analysis Reports for Nuclear Power Plants,'' on a proposed Revision 4 to Standard Review Plan (SRP... Position (BTP) 8-8 on ``Onsite (Emergency Diesel Generators) and Offsite Power Sources Allowed Outage Time... COMMISSION Office Of New Reactors; Proposed Revision 4 to Standard Review Plan; Section 8.1 on Electric...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-12-01

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

  15. Fission-Produced (99)Mo Without a Nuclear Reactor.

    Science.gov (United States)

    Youker, Amanda J; Chemerisov, Sergey D; Tkac, Peter; Kalensky, Michael; Heltemes, Thad A; Rotsch, David A; Vandegrift, George F; Krebs, John F; Makarashvili, Vakho; Stepinski, Dominique C

    2017-03-01

    (99)Mo, the parent of the widely used medical isotope (99m)Tc, is currently produced by irradiation of enriched uranium in nuclear reactors. The supply of this isotope is encumbered by the aging of these reactors and concerns about international transportation and nuclear proliferation. Methods: We report results for the production of (99)Mo from the accelerator-driven subcritical fission of an aqueous solution containing low enriched uranium. The predominately fast neutrons generated by impinging high-energy electrons onto a tantalum convertor are moderated to thermal energies to increase fission processes. The separation, recovery, and purification of (99)Mo were demonstrated using a recycled uranyl sulfate solution. Conclusion: The (99)Mo yield and purity were found to be unaffected by reuse of the previously irradiated and processed uranyl sulfate solution. Results from a 51.8-GBq (99)Mo production run are presented.

  16. Temperature measuring analysis of the nuclear reactor fuel assembly

    Science.gov (United States)

    F., Urban; Ľ., Kučák; Bereznai, J.; Závodný, Z.; Muškát, P.

    2014-08-01

    Study was based on rapid changes of measured temperature values from the thermocouple in the VVER 440 nuclear reactor fuel assembly. Task was to determine origin of fluctuations of the temperature values by experiments on physical model of the fuel assembly. During an experiment, heated water was circulating in the system and cold water inlet through central tube to record sensitivity of the temperature sensor. Two positions of the sensor was used. First, just above the central tube in the physical model fuel assembly axis and second at the position of the thermocouple in the VVER 440 nuclear reactor fuel assembly. Dependency of the temperature values on time are presented in the diagram form in the paper.

  17. Inception and evolution of Oklo natural nuclear reactors

    Science.gov (United States)

    Bentridi, Salah-Eddine; Gall, Benoît; Gauthier-Lafaye, François; Seghour, Abdeslam; Medjadi, Djamel-Eddine

    2011-11-01

    The occurrence of more than 15 natural nuclear Reactor Zones (RZ) in a geological environment remains a mystery even 40 years after their discovery. The present work gives for the first time an explanation of the chemical and physical processes that caused the start-up of the fission reactions with two opposite processes, uranium enrichments and progressive impoverishment in 235U. Based on Monte-Carlo neutronics simulations, a solution space was defined taking into account realistic combinations of relevant parameters acting on geological conditions and neutron transport physics. This study explains criticality occurrence, operation, expansion and end of life conditions of Oklo natural nuclear reactors, from the smallest to the biggest ones.

  18. Systems and methods for dismantling a nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Heim, Robert R; Adams, Scott Ryan; Cole, Matthew Denver; Kirby, William E; Linnebur, Paul Damon

    2014-10-28

    Systems and methods for dismantling a nuclear reactor are described. In one aspect the system includes a remotely controlled heavy manipulator ("manipulator") operatively coupled to a support structure, and a control station in a non-contaminated portion of a workspace. The support structure provides the manipulator with top down access into a bioshield of a nuclear reactor. At least one computing device in the control station provides remote control to perform operations including: (a) dismantling, using the manipulator, a graphite moderator, concrete walls, and a ceiling of the bioshield, the manipulator being provided with automated access to all internal portions of the bioshield; (b) loading, using the manipulator, contaminated graphite blocks from the graphite core and other components from the bioshield into one or more waste containers; and (c) dispersing, using the manipulator, dust suppression and contamination fixing spray to contaminated matter.

  19. ZEEP: Canada's first nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Green, R.E.; Okazaki, A. [retired, Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2015-09-15

    In 1905 Albert Einstein published his historic paper on special relativity, which contained the equation E=mc 2. The significance of this mass-energy relationship became evident with the discovery of nuclear fission in 1939, when it was realized that large amounts of energy would be released in a fission chain reaction. Canadian scientists were involved in this field from the beginning and their efforts resulted in the startup in September 1945 of the ZEEP reactor at Chalk River, the first reactor to go critical outside the USA. In this paper we recall some of the events that led to the construction of ZEEP, and describe the role it played in the development of the Canadian nuclear energy program. (author)

  20. Westinghouse Small Modular Reactor nuclear steam supply system design

    Energy Technology Data Exchange (ETDEWEB)

    Memmott, M. J.; Harkness, A. W.; Van Wyk, J. [Westinghouse Electric Company LLC, 600 Cranberry Woods Drive, Cranberry Twp. PA 16066 (United States)

    2012-07-01

    The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor (iPWR), in which all of the components typically associated with the nuclear steam supply system (NSSS) of a nuclear power plant are incorporated within a single reactor pressure vessel. This paper is the first in a series of four papers which describe the design and functionality of the Westinghouse SMR. Also described in this series are the key drivers influencing the design of the Westinghouse SMR and the unique passive safety features of the Westinghouse SMR. Several critical motivators contributed to the development and integration of the Westinghouse SMR design. These design driving motivators dictated the final configuration of the Westinghouse SMR to varying degrees, depending on the specific features under consideration. These design drivers include safety, economics, AP1000{sup R} reactor expertise and experience, research and development requirements, functionality of systems and components, size of the systems and vessels, simplicity of design, and licensing requirements. The Westinghouse SMR NSSS consists of an integral reactor vessel within a compact containment vessel. The core is located in the bottom of the reactor vessel and is composed of 89 modified Westinghouse 17x17 Robust Fuel Assemblies (RFA). These modified fuel assemblies have an active core length of only 2.4 m (8 ft) long, and the entirety of the core is encompassed by a radial reflector. The Westinghouse SMR core operates on a 24 month fuel cycle. The reactor vessel is approximately 24.4 m (80 ft) long and 3.7 m (12 ft) in diameter in order to facilitate standard rail shipping to the site. The reactor vessel houses hot and cold leg channels to facilitate coolant flow, control rod drive mechanisms (CRDM), instrumentation and cabling, an intermediate flange to separate flow and instrumentation and facilitate simpler refueling, a pressurizer, a straight tube, recirculating steam

  1. Specific schedule conditions for the formation of personnel of A or B category working in nuclear facilities. Option nuclear reactor

    CERN Document Server

    Int. At. Energy Agency, Wien

    2002-01-01

    This document describes the specific dispositions relative to the nuclear reactor domain, for the formation to the conventional and radiation risks prevention of personnel of A or B category working in nuclear facilities. The application domain, the applicable documents, the liability, the specificity of the nuclear reactor and of the retraining, the Passerelle formation, are presented. (A.L.B.)

  2. Towards an efficient multiphysics model for nuclear reactor dynamics

    Directory of Open Access Journals (Sweden)

    Obaidurrahman K.

    2015-01-01

    Full Text Available Availability of fast computer resources nowadays has facilitated more in-depth modeling of complex engineering systems which involve strong multiphysics interactions. This multiphysics modeling is an important necessity in nuclear reactor safety studies where efforts are being made worldwide to combine the knowledge from all associated disciplines at one place to accomplish the most realistic simulation of involved phenomenon. On these lines coupled modeling of nuclear reactor neutron kinetics, fuel heat transfer and coolant transport is a regular practice nowadays for transient analysis of reactor core. However optimization between modeling accuracy and computational economy has always been a challenging task to ensure the adequate degree of reliability in such extensive numerical exercises. Complex reactor core modeling involves estimation of evolving 3-D core thermal state, which in turn demands an expensive multichannel based detailed core thermal hydraulics model. A novel approach of power weighted coupling between core neutronics and thermal hydraulics presented in this work aims to reduce the bulk of core thermal calculations in core dynamics modeling to a significant extent without compromising accuracy of computation. Coupled core model has been validated against a series of international benchmarks. Accuracy and computational efficiency of the proposed multiphysics model has been demonstrated by analyzing a reactivity initiated transient.

  3. Hydrogen-or-Fossil-Combustion Nuclear Combined-Cycle Systems for Base- and Peak-Load Electricity Production

    Energy Technology Data Exchange (ETDEWEB)

    Forsberg, Charles W [ORNL; Conklin, Jim [ORNL

    2007-09-01

    A combined-cycle power plant is described that uses (1) heat from a high-temperature nuclear reactor to meet base-load electrical demands and (2) heat from the same high-temperature reactor and burning natural gas, jet fuel, or hydrogen to meet peak-load electrical demands. For base-load electricity production, fresh air is compressed; then flows through a heat exchanger, where it is heated to between 700 and 900 C by heat provided by a high-temperature nuclear reactor via an intermediate heat-transport loop; and finally exits through a high-temperature gas turbine to produce electricity. The hot exhaust from the Brayton-cycle gas turbine is then fed to a heat recovery steam generator that provides steam to a steam turbine for added electrical power production. To meet peak electricity demand, the air is first compressed and then heated with the heat from a high-temperature reactor. Natural gas, jet fuel, or hydrogen is then injected into the hot air in a combustion chamber, combusts, and heats the air to 1300 C-the operating conditions for a standard natural-gas-fired combined-cycle plant. The hot gas then flows through a gas turbine and a heat recovery steam generator before being sent to the exhaust stack. The higher temperatures increase the plant efficiency and power output. If hydrogen is used, it can be produced at night using energy from the nuclear reactor and stored until needed. With hydrogen serving as the auxiliary fuel for peak power production, the electricity output to the electric grid can vary from zero (i.e., when hydrogen is being produced) to the maximum peak power while the nuclear reactor operates at constant load. Because nuclear heat raises air temperatures above the auto-ignition temperatures of the various fuels and powers the air compressor, the power output can be varied rapidly (compared with the capabilities of fossil-fired turbines) to meet spinning reserve requirements and stabilize the electric grid. This combined cycle uses the

  4. Adaptation possibilities of the nuclear electricity production to the demand; Possibilites d'adaptation de la production d'electricite nucleaires a la demande

    Energy Technology Data Exchange (ETDEWEB)

    Acket, C

    2009-02-15

    Meeting about the CO{sub 2} emissions from electric heating showed that the leak of adaptation from nuclear reactors led the use of thermal power plants (coal, petroleum, gas) to control the demand variations. This argument is analyzed in the document: is it possible to replace those thermal power plants by nuclear reactors? In this framework the author analyzes the network needs, the electricity sources and the demand answer and the specificities of the nuclear. (A.L.B.)

  5. A Nuclear Reactor Transient Methodology Based on Discrete Ordinates Method

    Directory of Open Access Journals (Sweden)

    Shun Zhang

    2014-01-01

    Full Text Available With the rapid development of nuclear power industry, simulating and analyzing the reactor transient are of great significance for the nuclear safety. The traditional diffusion theory is not suitable for small volume or strong absorption problem. In this paper, we have studied the application of discrete ordinates method in the numerical solution of space-time kinetics equation. The fully implicit time integration was applied and the precursor equations were solved by analytical method. In order to improve efficiency of the transport theory, we also adopted some advanced acceleration methods. Numerical results of the TWIGL benchmark problem presented demonstrate the accuracy and efficiency of this methodology.

  6. Neutron physics for nuclear reactors unpublished writings by Enrico Fermi

    CERN Document Server

    Fermi, Enrico; Pisanti, O

    2010-01-01

    This unique volume gives an accurate and very detailed description of the functioning and operation of basic nuclear reactors, as emerging from yet unpublished papers by Nobel Laureate Enrico Fermi. In the first part, the entire course of lectures on Neutron Physics delivered by Fermi at Los Alamos is reported, according to the version made by Anthony P French. Here, the fundamental physical phenomena are described very clearly and comprehensively, giving the appropriate physics grounds for the functioning of nuclear piles. In the second part, all the patents issued by Fermi (and coworkers) on

  7. A coupled nuclear reactor thermal energy storage system for enhanced load following operation

    Science.gov (United States)

    Alameri, Saeed A.

    Nuclear power plants usually provide base-load electric power and operate most economically at a constant power level. In an energy grid with a high fraction of renewable energy sources, future nuclear reactors may be subject to significantly variable power demands. These variable power demands can negatively impact the effective capacity factor of the reactor and result in severe economic penalties. Coupling the reactor to a large Thermal Energy Storage (TES) block will allow the reactor to better respond to variable power demands. In the system described in this thesis, a Prismatic-core Advanced High Temperature Reactor (PAHTR) operates at constant power with heat provided to a TES block that supplies power as needed to a secondary energy conversion system. The PAHTR is designed to have a power rating of 300 MW th, with 19.75 wt% enriched Tri-Structural-Isotropic UO 2 fuel and a five year operating cycle. The passive molten salt TES system will operate in the latent heat region with an energy storage capacity of 150 MWd. Multiple smaller TES blocks are used instead of one large block to enhance the efficiency and maintenance complexity of the system. A transient model of the coupled reactor/TES system is developed to study the behavior of the system in response to varying load demands. The model uses six-delayed group point kinetics and decay heat models coupled to thermal-hydraulic and heat transfer models of the reactor and TES system. Based on the transient results, the preferred TES design consists of 1000 blocks, each containing 11000 LiCl phase change material tubes. A safety assessment of major reactor events demonstrates the inherent safety of the coupled system. The loss of forced circulation study determined the minimum required air convection heat removal rate from the reactor core and the lowest possible reduced primary flow rate that can maintain the reactor in a safe condition. The loss of ultimate heat sink study demonstrated the ability of the TES

  8. Determination of 36Cl in nuclear waste from reactor decommissioning

    DEFF Research Database (Denmark)

    Hou, Xiaolin; Frøsig, Lars; Nielsen, Sven Poul

    2007-01-01

    An analytical method for the determination of Cl-36 in nuclear waste such as graphite, heavy concrete, steel, aluminum, and lead was developed. Several methods were investigated for decomposing the samples. AgCl precipitation was used to separate Cl-36 from the matrix elements, followed by ion...... of this analytical method for Cl-36 is 14 mBq. The method has been used to determine Cl-36 in heavy concrete, aluminum, and graphite from the Danish DR-2 research reactor....

  9. Circuit for power variation rate measurements in a nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Moisin, L.H.

    1980-01-01

    An asychronous digital circuit for the power variation rate of a nuclear reactor is proposed. The circuit is based on the fact that the variation rate can be obtained by a simple division between the difference of two time normalized adjacent measurements of the neutron flux and the total value of the first measurement. The circuit maintains a constant precision of the counting rate due to the effect of an automatic time constant switch. 4 references.

  10. La política nuclear espanyola: el caos del reactor nuclear Argos

    OpenAIRE

    Barca i Salom, Francesc Xavier

    2000-01-01

    L’11 de juny de 1962 s’inaugurava a l’Escola Tècnica Superior d’Enginyeria Industrial de Barcelona un reactor nuclear experimental, que era batejat amb el nom mític d’Argos. Tota la premsa barcelonina se’n feu ressò i el presentava com el primer reactor construït íntegrament a Espanya per la Junta d’Energia Nuclear. La idea de dotar l’Escola d’un reactor nuclear havia nascut, però, set anys abans, precisament en el mateix moment de la creació de la Càtedra Ferran Tallada d’enginyeria...

  11. Space-reactor electric systems: subsystem technology assessment

    Energy Technology Data Exchange (ETDEWEB)

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

    1983-03-29

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

  12. Neutron dose estimation in a zero power nuclear reactor

    Science.gov (United States)

    Triviño, S.; Vedelago, J.; Cantargi, F.; Keil, W.; Figueroa, R.; Mattea, F.; Chautemps, A.; Santibañez, M.; Valente, M.

    2016-10-01

    This work presents the characterization and contribution of neutron and gamma components to the absorbed dose in a zero power nuclear reactor. A dosimetric method based on Fricke gel was implemented to evaluate the separation between dose components in the mixed field. The validation of this proposed method was performed by means of direct measurements of neutron flux in different positions using Au and Mg-Ni activation foils. Monte Carlo simulations were conversely performed using the MCNP main code with a dedicated subroutine to incorporate the exact complete geometry of the nuclear reactor facility. Once nuclear fuel elements were defined, the simulations computed the different contributions to the absorbed dose in specific positions inside the core. Thermal/epithermal contributions of absorbed dose were assessed by means of Fricke gel dosimetry using different isotopic compositions aimed at modifying the sensitivity of the dosimeter for specific dose components. Clear distinctions between gamma and neutron capture dose were obtained. Both Monte Carlo simulations and experimental results provided reliable estimations about neutron flux rate as well as dose rate during the reactor operation. Simulations and experimental results are in good agreement in every positions measured and simulated in the core.

  13. BENCHMARK EVALUATION OF THE START-UP CORE REACTOR PHYSICS MEASUREMENTS OF THE HIGH TEMPERATURE ENGINEERING TEST REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    John Darrell Bess

    2010-05-01

    The benchmark evaluation of the start-up core reactor physics measurements performed with Japan’s High Temperature Engineering Test Reactor, in support of the Next Generation Nuclear Plant Project and Very High Temperature Reactor Program activities at the Idaho National Laboratory, has been completed. The evaluation was performed using MCNP5 with ENDF/B-VII.0 nuclear data libraries and according to guidelines provided for inclusion in the International Reactor Physics Experiment Evaluation Project Handbook. Results provided include updated evaluation of the initial six critical core configurations (five annular and one fully-loaded). The calculated keff eigenvalues agree within 1s of the benchmark values. Reactor physics measurements that were evaluated include reactivity effects measurements such as excess reactivity during the core loading process and shutdown margins for the fully-loaded core, four isothermal temperature reactivity coefficient measurements for the fully-loaded core, and axial reaction rate measurements in the instrumentation columns of three core configurations. The calculated values agree well with the benchmark experiment measurements. Fully subcritical and warm critical configurations of the fully-loaded core were also assessed. The calculated keff eigenvalues for these two configurations also agree within 1s of the benchmark values. The reactor physics measurement data can be used in the validation and design development of future High Temperature Gas-cooled Reactor systems.

  14. Hydraulic Shuttle Irradiation System (HSIS) Recently Installed in the Advanced Test Reactor (ATR)

    Energy Technology Data Exchange (ETDEWEB)

    A. Joseph Palmer; Gerry L. McCormick; Shannon J. Corrigan

    2010-06-01

    2010 International Congress on Advances in Nuclear Power Plants (ICAPP’10) ANS Annual Meeting Imbedded Topical San Diego, CA June 13 – 17, 2010 Hydraulic Shuttle Irradiation System (HSIS) Recently Installed in the Advanced Test Reactor (ATR) Author: A. Joseph Palmer, Mechanical Engineer, Irradiation Test Programs, 208-526-8700, Joe.Palmer@INL.gov Affiliation: Idaho National Laboratory P.O. Box 1625, MS-3840 Idaho Falls, ID 83415 INL/CON-10-17680 ABSTRACT Most test reactors are equipped with shuttle facilities (sometimes called rabbit tubes) whereby small capsules can be inserted into the reactor and retrieved during power operations. With the installation of Hydraulic Shuttle Irradiation System (HSIS) this capability has been restored to the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). The general design and operating principles of this system were patterned after the hydraulic rabbit at Oak Ridge National Laboratory’s (ORNL) High Flux Isotope Reactor (HFIR), which has operated successfully for many years. Using primary coolant as the motive medium the HSIS system is designed to simultaneously transport fourteen shuttle capsules, each 16 mm OD x 57 mm long, to and from the B-7 position of the reactor. The B-7 position is one of the higher flux positions in the reactor with typical thermal and fast (>1 Mev) fluxes of 2.8E+14 n/cm2/sec and 1.9E+14 n/cm2/sec respectively. The available space inside each shuttle is approximately 14 mm diameter x 50 mm long. The shuttle containers are made from titanium which was selected for its low neutron activation properties and durability. Shuttles can be irradiated for time periods ranging from a few minutes to several months. The Send and Receive Station (SRS) for the HSIS is located 2.5 m deep in the ATR canal which allows irradiated shuttles to be easily moved from the SRS to a wet loaded cask, or transport pig. The HSIS system first irradiated (empty) shuttles in September 2009 and has since completed

  15. 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

  16. Evaluation of LLNL's Nuclear Accident Dosimeters at the CALIBAN Reactor September 2010

    Energy Technology Data Exchange (ETDEWEB)

    Hickman, D P; Wysong, A R; Heinrichs, D P; Wong, C T; Merritt, M J; Topper, J D; Gressmann, F A; Madden, D J

    2011-06-21

    The Lawrence Livermore National Laboratory uses neutron activation elements in a Panasonic TLD holder as a personnel nuclear accident dosimeter (PNAD). The LLNL PNAD has periodically been