WorldWideScience

Sample records for general electric nuclear test reactor

  1. Safety Evaluation Report related to the renewal of the operating license for the General Electric-Nuclear Test Reactor (GE-NTR) (Docket No. 50-73)

    International Nuclear Information System (INIS)

    1984-09-01

    This Safety Evaluation Report for the application filed by the General Electric Company (GE) for a renewal license number R-33 to continue to operate its research reactor has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is owned and operated by GE and is located in Pleasanton, California. The staff concludes that the reactor can continue to be operated by GE without endangering the health and safety of the public

  2. Generalities about nuclear reactors

    International Nuclear Information System (INIS)

    Jaouen, C.; Beroux, P.

    2012-01-01

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

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

    International Nuclear Information System (INIS)

    1987-07-01

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

  4. General Electric Nuclear Energy

    International Nuclear Information System (INIS)

    2001-01-01

    The ESBWR is a 1380 MWe boiling water reactor with improved operating safety margins and passive safety systems. He stated that the ESBWR derived from earlier GE plant design certification efforts and is the result of eight years of International cooperative work. He stated that the biggest challenge is to cross the regulatory hurdles associated with the inspections, tests, analyses, and acceptance criteria (ITAAC) and combined license (COL) programs. He further stated that he did not know how long it might take to license the ESBWR, in part, because the last GE design certification took about 8 to 10 years. Dr. Rao also provided a brief overview of the GE Nuclear Advance Liquid Metal S-PRISM design

  5. 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.

  6. Reactors for nuclear electric propulsion

    International Nuclear Information System (INIS)

    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

  7. Nuclear reactor development in China for non-electrical applications

    International Nuclear Information System (INIS)

    Sun Yuliang; Zhong Daxin; Dong Duo; Xu Yuanhui

    1998-01-01

    In parallel to its vigorous program of nuclear power generation, China has attached great importance to the development of nuclear reactors for non-electrical applications. The Institute of Nuclear Energy Technology (INET) in Beijing has been developing technologies of the water-cooled heating reactor and the modular high temperature gas-cooled reactor. In 1989, a 5 MW water cooled test reactor was erected. Currently, an industrial demonstration nuclear heating plant is being projected. Feasibility studies are being made of sea-water desalination using the INET developed nuclear heating reactor as heat source. Also, a 10 MW high temperature gas-cooled test reactor is being constructed at INET in the framework of China's national high-tech program. The paper gives an overview of China's energy market situation. With respect to China's technology development of high temperature gas-cooled reactors and water cooled heating reactors, the paper describes some general requirements on the technical development, reviews the national programs and activities, describes briefly the design and safety features of the reactor concepts, discusses aspects of application potentials. (author)

  8. 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.

  9. Examination on the testing method for evaluating life of electric wires and cables for nuclear reactors

    International Nuclear Information System (INIS)

    Seguchi, Tadao; Morita, Yosuke; Yoshida, Kenzo

    1984-01-01

    Regarding the method of environmental test on the electric wires and cables used for the safety system in the containment vessels of nuclear power plants, the draft recommendation was issued by the Institute of Electrical Engineers of Japan in 1982. Its contents follow the IEEE Standard of USA, and are composed of the tests on the deterioration in normal operation, the case of LOCA and the prevention of spread of fire. In this report, as to the testing method regarding normal operation, the appropriate method and its basis are described in view of the recent experimental data. In the draft recommendation, the successive method carrying out irradiation after thermal deterioration is adopted, and both testing conditions are given. However, the fundamental problems remain in the propriety of the acceleration of deterioration and the multiplied effect of heat and radiation. The qualitative and quantitative data on these problems have been accumulated in various countries, therefore, the examination of the testing method was carried out based on these data. The dose rate dependence of radiation deterioration, the multiplied effect of radiation and heat, and the correlation of thermal deterioration rate with temperature are discussed. The appropriate method is proposed. (Kako, I.)

  10. Reactor design for nuclear electric propulsion

    International Nuclear Information System (INIS)

    Koenig, D.R.; Ranken, W.A.

    1979-01-01

    Conceptual design studies of a nuclear power plant for electric propulsion of spacecrafts have been on going for several years. An attractive concept which has evolved from these studies and which has been described in previous publications, is a heat-pipe cooled, fast spectrum nuclear reactor that provides 3 MW of thermal energy to out-of-core thermionic converters. The primary motivation for using heat pipes is to provide redundancy in the core cooling system that is not available in gas or liquid-metal cooled reactors. Detailed investigation of the consequences of heat pipe failures has resulted in modifications to the basic reactor design and has led to consideration of an entirely different core design. The new design features an integral laminated core configuration consisting of alternating layers of UO 2 and molybdenum sheets that span the entire diameter of the core. Design characteristics are presented and compared for the two reactors

  11. Electric-stepping-motor tests for a control-drum actuator of a nuclear reactor

    Science.gov (United States)

    Kieffer, A. W.

    1972-01-01

    Experimental tests were conducted on two stepping motors for application as reactor control-drum actuators. Various control-drum loads with frictional resistances ranging from approximately zero to 40 N-m and inertias ranging from zero to 0.424 kg-sq m were tested.

  12. Service to the Electric Utility Industry by the Ford Nuclear Reactor, University of Michigan

    International Nuclear Information System (INIS)

    Burn, R.R.; Simpson, P.A.; Cook, G.M.

    1993-01-01

    Since 1977, the staff of the University of Michigan's Ford Nuclear Reactor has been providing irradiation, testing, analytical, and training services to electric utilities and to suppliers of the nuclear electric utility industry. This paper discusses the reactor's irradiation facilities; reactor programs and utilization; materials testing programs; neutron activation analysis activities; and training programs conducted

  13. Fire safety requirements for electrical cables towards nuclear reactor safety

    International Nuclear Information System (INIS)

    Raju, M.R.

    2002-01-01

    Full text: Electrical power supply forms a very important part of any nuclear reactor. Power supplies have been categorized in to class I, II, III and IV from reliability point. The safety related equipment are provided with highly reliable power supply to achieve the safety of very high order. Vast network of cables in a nuclear reactor are grouped and segregated to ensure availability of power to at least one group under all anticipated occurrences. Since fire can result in failures leading to unavailability of power caused by common cause, both passive and active fire protection methods are adopted in addition to fire detection system. The paper describes the requirement for passive fire protection to electrical cables viz. fire barrier and fire breaks. The paper gives an account of the tests required to standardize the products. Fire safety implementation for cables in research reactors is described

  14. Nondestructive testing of nuclear reactor components integrity

    International Nuclear Information System (INIS)

    Mala, M.; Miklos, M.

    2011-01-01

    Nuclear energy must respond to current challenges in the energy market. The significant parameters are increase of the nuclear fuel price, closed fuel cycle, reduction and safe and the final disposal of high level radioactive waste. Nowadays, the discussions on suitable energy mix are taking place not only here in Czech Republic, but also in many other European countries. It is necessary to establish an appropriate ratio among the production of electricity from conventional, nuclear and renewable energy sources. Also, it is necessary to find ways how to streamline the economy, central part of the nuclear fuel cycle and thereby to increase the competitiveness of nuclear energy. This streamlining can be carried out by improving utilization of existing nuclear fuel with maintaining a high degree of nuclear facilities safety. Increasing operational reliability and safety together with increasing utilization of nuclear fuel place increasing demands on monitoring of changes during fuel burnup. The potential fuel assembly damages in light water reactors are prevented by the introduction of new procedures and programs of the fuel assembly monitoring. One of them is the Post Irradiation Inspection Program (PIIP) which is a good tool for monitoring of chemical regime impact on the fuel assembly cladding behavior. Main nondestructive techniques that are used at nuclear power plants for the fuel assembly integrity evaluation are ultrasonic measurements, eddy current measurements, radiographic testing, acoustic techniques and others. Ultrasonic system is usual tool for leak fuel rod evaluation and it is also used at Temelin NPP. Since 2009, Temelin NPP has cooperated with Research Center Rez Ltd in frame of PIIP program at both units WWER 1000. This program was established for US VVantage6 fuel assemblies and also it continues for Russian TVSA-T fuel assemblies. (author)

  15. Ground test facility for nuclear testing of space reactor subsystems

    International Nuclear Information System (INIS)

    Quapp, W.J.; Watts, K.D.

    1985-01-01

    Two major reactor facilities at the INEL have been identified as easily adaptable for supporting the nuclear testing of the SP-100 reactor subsystem. They are the Engineering Test Reactor (ETR) and the Loss of Fluid Test Reactor (LOFT). In addition, there are machine shops, analytical laboratories, hot cells, and the supporting services (fire protection, safety, security, medical, waste management, etc.) necessary to conducting a nuclear test program. This paper presents the conceptual approach for modifying these reactor facilities for the ground engineering test facility for the SP-100 nuclear subsystem. 4 figs

  16. Proposal of space reactor for nuclear electric propulsion system

    International Nuclear Information System (INIS)

    Nishiyama, Takaaki; Nagata, Hidetaka; Nakashima, Hideki

    2009-01-01

    A nuclear reactor installed in spacecrafts is considered here. The nuclear reactor could stably provide an enough amount of electric power in deep space missions. Most of the nuclear reactors that have been developed up to now in the United States and the former Soviet Union have used uranium with 90% enrichment of 235 U as a fuel. On the other hand, in Japan, because the uranium that can be used is enriched to below 20%, the miniaturization of the reactor core is difficult. A Light-water nuclear reactor is an exception that could make the reactor core small. Then, the reactor core composition and characteristic are evaluated for the cases with the enrichment of the uranium fuel as 20%. We take up here Graphite reactor, Light-water reactor, and Sodium-cooled one. (author)

  17. Diagnosis of electric equipment at the Dalat Nuclear Research Reactor

    International Nuclear Information System (INIS)

    Nguyen Truong Sinh

    1999-01-01

    The Dalat Nuclear Research Reactor (DNRR) is a pool type of its kind in the world: Soviet-designed core and control system harmoniously integrated into the left-over infrastructure of the former American-made TRIGA MARK II reactor, which includes the reactor tank and shielding, graphite reflector, beam tubes and thermal column. The reactor is mainly used for radioisotope and radiopharmaceutical production, elemental analysis using neutron activation techniques, neutron beam exploitation, silicon doping, and reactor physics experimentation. For safe operation of the reactor maintenance work has been carried out for the reactor control and instrumentation, reactor cooling, ventilation, radiomonitoring, mechanical, normal electric supply systems as well as emergency electric diesel generators and the water treatment station. Technical management of the reactor includes periodical maintenance as required by technical specifications, training, re-training and control of knowledge for reactor staff. During recent years, periodic preventive maintenance (PPM) has been carried out for the electric machines of the technological systems. (author)

  18. Nuclear reactors for space electric power

    International Nuclear Information System (INIS)

    Buden, D.

    1978-06-01

    The Los Alamos Scientific Laboratory is studying reactor power plants for space applications in the late 1980s and 1990s. The study is concentrating on high-temperature, compact, fast reactors that can be coupled with various radiation shielding systems and thermoelectric, dynamic, or thermionic electric power conversion systems, depending on the mission. Lifetimes of 7 to 10 yr at full power, at converter operating temperatures of 1275 to 1675 0 K, are being studied. The systems are being designed such that no single-failure modes exist that will cause a complete loss of power. In fact, to meet the long lifetimes, highly redundant design features are being emphasized. Questions have been raised about safety since the COSMOS 954 incident. ''Fail-safe'' means to prevent exposure of the population to radioactive material, meeting the environmental guidelines established by the U.S. Government have been and continue to be a necessary requirement for any space reactor program. The major safety feature to prevent prelaunch and launch radioactive material hazards is not operating the reactor before achieving the prescribed orbit. Design features in the reactor ensure that accidental criticality cannot occur. High orbits (above 400 to 500 nautical miles) have sufficient lifetimes to allow radioactive elements to decay to safe levels. The major proposed applications for satellites with reactors in Earth orbit are in geosynchronous orbit (19,400 nautical miles). In missions at geosynchronous orbit, where orbital lifetimes are practically indefinite, the safety considerations are negligible. Orbits below 400 to 500 nautical miles are the ones where a safety issue is involved in case of satellite malfunction. The potential missions, the question of why reactors are being considered as a prime power candidate, reactor features, and safety considerations will be discussed

  19. Thermionic reactor power conditioner design for nuclear electric propulsion.

    Science.gov (United States)

    Jacobsen, A. S.; Tasca, D. M.

    1971-01-01

    Consideration of the effects of various thermionic reactor parameters and requirements upon spacecraft power conditioning design. A basic spacecraft is defined using nuclear electric propulsion, requiring approximately 120 kWe. The interrelationships of reactor operating characteristics and power conditioning requirements are discussed and evaluated, and the effects on power conditioner design and performance are presented.

  20. Nuclear fuels for material test reactors

    International Nuclear Information System (INIS)

    Ramanathan, L.V.; Durazzo, M.; Freitas, C.T. de

    1982-01-01

    Experimental results related do the development of nuclear fuels for reactors cooled and moderated by water have been presented cylindrical and plate type fuels have been described in which the core consists of U compouns dispersed in an Al matrix and is clad with aluminium. Fabrication details involving rollmilling, swaging or hot pressing have been described. Corrosion and irradiation test results are also discussed. The performance of the different types of fuels indicates that it is possible to locally fabricate fuel plates with U 3 O 8 +Al cores (20% enriched U) for use in operating Brazilian research reactors. (Author) [pt

  1. A quality assurance program for nuclear power reactor materials tests at the Ford nuclear reactor

    International Nuclear Information System (INIS)

    Burn, R.R.

    1989-01-01

    The University of Michigan Nuclear Reactor Laboratory Quality Assurance Program has been established to assure that materials testing services provided to electric utilities produce accurate results in accordance with industry standards, sound engineering practice, and customer requirements. The program was prepared to comply with applicable requirements of 10CFR50, Appendix B, of the Code of Federal Regulations and a standard of the American National Standards Institute (ANSI), N45.2. The paper discusses the quality assurance program applicability, organization, qualification and training of personnel, material identification and control, examination and testing, measuring and test equipment, nonconforming test equipment, records, audits, and distribution

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

    International Nuclear Information System (INIS)

    Bottimore, R.R.

    1980-12-01

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

  3. Electrical cabling system associated at a nuclear reactor

    International Nuclear Information System (INIS)

    Dejeux, P.; Desfontaines, G.

    1988-01-01

    This cabling system for an electrical device in a nuclear reactor comprises at least a first cable issued of the device, a second cable comprising a first portion, a second portion and a third portion joining the second by a multiple quick fitting connector capable to connect at least ten second portions at ten other third portions of the second cable [fr

  4. Nuclear reactors for electric power generation

    International Nuclear Information System (INIS)

    Hoogenboom, J.E.

    1987-01-01

    In this article the operation of a nuclear power plant, the status quo about the application of nuclear energy in the world are explained, the subjects of discussion between supporters and adversaries nowadays and the prospects for prolonged usage of nuclear power are summarized, viewed from the actual technical possibilities. 2 refs.; 7 figs.; 2 tabs

  5. Nuclear science. U.S. electricity needs and DOE's civilian reactor development program

    International Nuclear Information System (INIS)

    England-Joseph, Judy; Allen, Robert E. Jr.; Fitzgerald, Duane; Young, Edward E. Jr.; Leavens, William P.; Bell, Jacqueline

    1990-05-01

    Electricity projections developed by the North American Electric Reliability Council (NERC) appear to be the best available estimates of future U.S. electricity needs. NERC, which represents all segments of the utility industry, forecasts that before 1998 certain regions of the country, particularly in the more heavily populated eastern half of the United States, may experience shortfalls during summer peak demand periods. These forecasts considered the utility companies' plans, as of 1989, to meet electricity needs during the period; these plans include such measures as constructing additional generators and conducting demand management programs. Working closely with the nuclear industry, DOE is supporting the development of several reactor technologies to ensure that nuclear power remains a viable electricity supply option. In fiscal year 1990, DOE's Civilian Reactor Development Program was funded at $253 million. DOE is using these funds to support industry-led efforts to develop light water reactors (LWR), advanced liquid-metal reactors (LMR), and modular high-temperature gas-cooled reactors (MHTGR) that are safe, environmentally acceptable, and economically competitive. The utility company officials we spoke with, all of whom were in the Southeast, generally supported DOE's efforts in developing these technologies. However, most of the officials do not plan to purchase nuclear reactors until after 2000 because of the high costs of constructing nuclear reactors and current public opposition to nuclear power

  6. The behavior of fission products during nuclear rocket reactor tests

    International Nuclear Information System (INIS)

    Bokor, P.C.; Kirk, W.L.; Bohl, R.J.

    1991-01-01

    Fission product release from nuclear rocket propulsion reactor fuel is an important consideration for nuclear rocket development and application. Fission product data from the last six reactors of the Rover program are collected in this paper to provide as basis for addressing development and testing issues. Fission product loss from the fuel will depend on fuel composition and reactor design and operating parameters. During ground testing, fission products can be contained downstream of the reactor. The last Rover reactor tested, the Nuclear Furnance, was mated to an effluent clean-up system that was effective in preventing the discharge of fission products into the atmosphere

  7. Nuclear reactor power for an electrically powered orbital transfer vehicle

    Science.gov (United States)

    Jaffe, L.; Beatty, R.; Bhandari, P.; Chow, E.; Deininger, W.; Ewell, R.; Fujita, T.; Grossman, M.; Kia, T.; Nesmith, B.

    1987-01-01

    To help determine the systems requirements for a 300-kWe space nuclear reactor power system, a mission and spacecraft have been examined which utilize electric propulsion and this nuclear reactor power for multiple transfers of cargo between low earth orbit (LEO) and geosynchronous earth orbit (GEO). A propulsion system employing ion thrusters and xenon propellant was selected. Propellant and thrusters are replaced after each sortie to GEO. The mass of the Orbital Transfer Vehicle (OTV), empty and dry, is 11,000 kg; nominal propellant load is 5000 kg. The OTV operates between a circular orbit at 925 km altitude, 28.5 deg inclination, and GEO. Cargo is brought to the OTV by Shuttle and an Orbital Maneuvering Vehicle (OMV); the OTV then takes it to GEO. The OTV can also bring cargo back from GEO, for transfer by OMV to the Shuttle. OTV propellant is resupplied and the ion thrusters are replaced by the OMV before each trip to GEO. At the end of mission life, the OTV's electric propulsion is used to place it in a heliocentric orbit so that the reactor will not return to earth. The nominal cargo capability to GEO is 6000 kg with a transit time of 120 days; 1350 kg can be transferred in 90 days, and 14,300 kg in 240 days. These capabilities can be considerably increased by using separate Shuttle launches to bring up propellant and cargo, or by changing to mercury propellant.

  8. Nuclear reactor power for an electrically powered orbital transfer vehicle

    International Nuclear Information System (INIS)

    Jaffe, L.; Beatty, R.; Bhandari, P.

    1987-01-01

    To help determine the systems requirements for a 300-kWe space nuclear reactor power system, a mission and spacecraft have been examined which utilize electric propulsion and this nuclear reactor power for multiple transfers of cargo between low Earth orbit (LEO) and geosynchronous Earth orbit (GEO). A propulsion system employing ion thrusters and xenon propellant was selected. Propellant and thrusters are replaced after each sortie to GEO. The mass of the Orbital Transfer Vehicle (OTV), empty and dry, is 11,000 kg; nominal propellant load is 5000 kg. The OTV operates between a circular orbit at 925 km altitude, 28.5 deg inclination, and GEO. Cargo is brought to the OTV by Shuttle and an Orbital Maneuvering Vehicle (OMV); the OTV then takes it to GEO. The OTV can also bring cargo back from GEO, for transfer by OMV to the Shuttle. OTV propellant is resupplied and the ion thrusters are replaced by the OMV before each trip to GEO. At the end of mission life, the OTV's electric propulsion is used to place it in a heliocentric orbit so that the reactor will not return to Earth. The nominal cargo capability to GEO is 6000 kg with a transit time of 120 days; 1350 kg can be transferred in 90 days, and 14,300 kg in 240 days. These capabilities can be considerably increased by using separate Shuttle launches to bring up propellant and cargo, or by changing to mercury propellant

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

    International Nuclear Information System (INIS)

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

    1976-08-01

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

  10. Non-electric applications of pool-type nuclear reactors

    International Nuclear Information System (INIS)

    Adamov, E.O.; Cherkashov, Yu.M.; Romenkov, A.A.

    1997-01-01

    This paper recommends the use of pool-type light water reactors for thermal energy production. Safety and reliability of these reactors were already demonstrated to the public by the long-term operation of swimming pool research reactors. The paper presents the design experience of two projects: Apatity Underground Nuclear Heating Plant and Nuclear Sea-Water Desalination Plant. The simplicity of pool-type reactors, the ease of their manufacturing and maintenance make this type of a heat source attractive to the countries without a developed nuclear industry. (author). 6 figs, 1 tab

  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. Nuclear reactor vessel surface inspecting technique applying electric resistance probe

    International Nuclear Information System (INIS)

    Yamaguchi, T.; Enami, K.; Yoshioka, M.

    1975-01-01

    A new technique for inspecting the inner surface of the PWR type nuclear reactor vessel by use of an electric resistance probe is introduced, centering on a data processing system. This system is composed of a mini-computer, a system typewriter, an interface unit, a D-A converter and controller, and X-Y recorder and others. Its functions are judging flaws and making flaw detection maps. In order to judge flaws by flaw detection signals, three kinds of flaw judging methods have been developed. In case there is a flaw, its position and depth are calculated and listed on the system typewriter. The flaw detection maps are expressed in four kinds of modes and they are displayed on the X-Y recorder. (auth.)

  13. Nuclear power reactors

    International Nuclear Information System (INIS)

    1982-11-01

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

  14. Nuclear reactors. Introduction

    International Nuclear Information System (INIS)

    Boiron, P.

    1997-01-01

    This paper is an introduction to the 'nuclear reactors' volume of the Engineers Techniques collection. It gives a general presentation of the different articles of the volume which deal with: the physical basis (neutron physics and ionizing radiations-matter interactions, neutron moderation and diffusion), the basic concepts and functioning of nuclear reactors (possible fuel-moderator-coolant-structure combinations, research and materials testing reactors, reactors theory and neutron characteristics, neutron calculations for reactor cores, thermo-hydraulics, fluid-structure interactions and thermomechanical behaviour of fuels in PWRs and fast breeder reactors, thermal and mechanical effects on reactors structure), the industrial reactors (light water, pressurized water, boiling water, graphite moderated, fast breeder, high temperature and heavy water reactors), and the technology of PWRs (conceiving and building rules, nuclear parks and safety, reactor components and site selection). (J.S.)

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

    International Nuclear Information System (INIS)

    Fletcher, C.D.

    1986-01-01

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

  16. Chinese nuclear heating test reactor and demonstration plant

    International Nuclear Information System (INIS)

    Wang Dazhong; Ma Changwen; Dong Duo; Lin Jiagui

    1992-01-01

    In this report the importance of nuclear district heating is discussed. From the viewpoint of environmental protection, uses of energy resources and transport, the development of nuclear heating in China is necessary. The development program of district nuclear heating in China is given in the report. At the time being, commissioning of the 5 MW Test Heating Reactor is going on. A 200 MWt Demonstration Plant will be built. In this report, the main characteristics of these reactors are given. It shows this type of reactor has a high inherent safety. Further the report points out that for this type of reactor the stability is very important. Some experimental results of the driving facility are included in the report. (orig.)

  17. Status report on nuclear reactors for space electric power

    International Nuclear Information System (INIS)

    Buden, D.

    1978-01-01

    The Los Alamos Scientific Laboratory is studying reactor power plants for space applications in the late 1980s and 1990s. The study is concentrating on high-temperature, compact, fast reactors that can be coupled with various radiation shielding systems and thermoelectric, dynamic, or thermionic electric power conversion systems, depending on the mission. Increased questions have been raised about safety since the COSMOS 954 incident. High orbits (above 400 to 500 nautical miles) have sufficient lifetimes to allow radioactive elements to decay to safe levels. The major proposed applications for satellites with reactors in Earth orbit are in geosynchronous orbit (19,400 nautical miles). In missions at geosynchronous orbit where orbital lifetimes are practically indefinite, the safety considerations are negligible. The potential missions, why reactors are being considered as a prime power candidate, reactor features, and safety considerations are discussed

  18. Reactor/Brayton power systems for nuclear electric spacecraft

    Science.gov (United States)

    Layton, J. P.

    1980-01-01

    Studies are currently underway to assess the technological feasibility of a nuclear-reactor-powered spacecraft propelled by electric thrusters. This vehicle would be capable of performing detailed exploration of the outer planets of the solar system during the remainder of this century. The purpose of this study was to provide comparative information on a closed cycle gas turbine power conversion system. The results have shown that the performance is very competitive and that a 400 kWe space power system is dimensionally compatible with a single Space Shuttle launch. Performance parameters of system mass and radiator area were determined for systems from 100 to 1000 kWe. A 400 kWe reference system received primary attention. The components of this system were defined and a conceptual layout was developed with encouraging results. The preliminary mass determination for the complete power system was very close to the desired goal of 20 kg/kWe. Use of more advanced technology (higher turbine inlet temperature) will substantially improve system performance characteristics.

  19. Nuclear reactor

    International Nuclear Information System (INIS)

    Miyashita, Akio.

    1981-01-01

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

  20. Standard technical specifications for General Electric boiling water reactors

    International Nuclear Information System (INIS)

    1979-08-01

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

  1. Improving the fidelity of electrically heated nuclear systems testing using simulated neutronic feedback

    International Nuclear Information System (INIS)

    Bragg-Sitton, Shannon M.; Godfroy, Thomas J.; Webster, Kenny

    2010-01-01

    Nonnuclear test platforms and methodologies can be employed to reduce the overall cost, risk and complexity of testing nuclear systems while allowing one to evaluate the operation of an integrated nuclear system within a reasonable timeframe, providing valuable input to the overall system design. In a nonnuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Standard electric test techniques allow one to fully assess thermal, heat transfer, and stress related attributes of a given system, but these approaches fail 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 testing with nuclear fuel elements installed. 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. This paper summarizes the results of initial system dynamic response testing for two electrically heated reactor concepts: a heat pipe-cooled reactor simulator with integrated heat exchanger and a gas-cooled reactor simulator with integrated Brayton power conversion system. Initial applications apply a simplified reactor kinetics model with either a single or an averaged measured state point. Preliminary results demonstrate the applicability of the dynamic test methodology to any reactor type, elucidating the variation in system response characteristics in different reactor concepts. These results suggest a need to further enhance the dynamic test approach by incorporating a more accurate model of the reactor dynamics and improved hardware instrumentation for better state estimation in application of the

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

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

    International Nuclear Information System (INIS)

    Bonnemay, A.; Dansac Bon, V.

    1978-01-01

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

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  5. Nuclear electric capacity expansion in Mexico: system effects of reactor size and cost

    International Nuclear Information System (INIS)

    Thayer, G.R.; Abbey, D.S.; Hardie, R.W.; Enriquez, R.P.; Uria, E.G.

    1984-01-01

    Mexico's electrical generation capacity could more than double over the next ten years - from about 15 GWe currently to as much as 35 GWe in 1990. While new capacity additions will be predominantly oil-fired in the 1980's, nuclear power will become increasingly important in the 1990's. This study investigated the appropriate size of new, nuclear capacity additions by assessing the implications of installing different size reactors into Mexico's electrical grid. Included in the assessments of reactor sizes are estimates of electrical generation costs and comparisons of the effective load-carrying capability of a 10 GWe nuclear capacity expansion

  6. Technical management on commissioning test of nuclear heating reactor

    International Nuclear Information System (INIS)

    Zhang Yajun; Su Qingshan

    1999-01-01

    The commissioning is the last construction stage of a nuclear heating project. The commissioning quality will directly affect on the safe operation and availability of the heating reactor. The author presents the whole test process until the completion of the test report from the point of test documents, including the preparation and execution of the test, the management of the various unexpected events during the test. And it will be emphatically discussed that the managing procedures of the various unexpected events during the test, including temporary control change, setpoint change, unexpected events and design change

  7. Nuclear reactors

    International Nuclear Information System (INIS)

    Barre, Bertrand

    2015-10-01

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

  8. Portland General Electric Company report on the operating and startup experience with control and instrumentation and electrical systems at the Trojan Nuclear Plant

    International Nuclear Information System (INIS)

    Zimmerman, G.A.

    1977-01-01

    The Trojan Nuclear Plant is an 1178 MWe nuclear plant located on the Columbia River 40 miles northwest of Portland, Oregon. The Nuclear Stream Supply System vendor is Westinghouse with a General Electric turbine generator. The reactor is rated and licensed for 3423 MWt (1178 MWe) and the turbine generator is designed for 3570 MWt(1219 MWe). The startup phase testing of Trojan commenced on November 21, 1975, upon receipt of our NRC Operating License. The startup testing program was completed on May 22, 1976, following 100 hours of full-power operation, at which time a scheduled summer maintenance outage began. Some of the highlights and milestones of the startup testing program are described

  9. Generic risk insights for General Electric boiling water reactors

    International Nuclear Information System (INIS)

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

    1991-05-01

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

  10. A European view of the use of nuclear reactors for applications other than electricity generation

    International Nuclear Information System (INIS)

    Marsham, T.N.; Brierley, G.

    Energy demands and temperature ranges employed by heat-consuming industrial processes are analyzed. Matching heat demand to reactor size is a problem. Emphasis is placed on HTGR's providing heat in the range 300-800 deg C. Further non-electrical uses of nuclear power, like nuclear ship propulsion, are analyzed. (E.C.B.)

  11. Environmental qualification test of electrical penetration for nuclear power stations

    International Nuclear Information System (INIS)

    Kooziro, Tetsuya; Nakagawa, Akitoshi; Toyoda, Shigeru; Uno, Shunpei

    1979-01-01

    Environmental qualification test was conducted according to IEEE Std. 323-1974 in order to evaluate the safety and reliability of electrical penetration of PWR type nuclear power station. Electrical penetration is the assemblies of electric cables attached to the containment vessel and penetrate through the vessel. Since it is a part of the vessel, it is deemed to be one of the primary safety equipments that are important for the safety and reliability of nuclear power stations. Environmental tests were conducted continuously as to heat cycle, vibration and LOCA with the full size specimens of bushing type, pigtail type and triaxial cable type and at the same time thermal life and irradiation tests were conducted on the insulation materials used, in order to obtain the comprehensive evaluation of their electrical and mechanical characteristics. As the result, they all satisfied the requirements for the circuits for actual use during and after various environmental qualification tests according to IEEE Std. 323. (author)

  12. Guidebook to nuclear reactors

    International Nuclear Information System (INIS)

    Nero, A.V. Jr.

    1976-05-01

    A general introduction to reactor physics and theory is followed by descriptions of commercial nuclear reactor types. Future directions for nuclear power are also discussed. The technical level of the material is suitable for laymen

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

    International Nuclear Information System (INIS)

    Vejvoda, E.J.; Pottmeyer, J.A.; DeLorenzo, D.S.; Weyns-Rollosson, M.I.; Duncan, D.R.

    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

  14. 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.

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

    International Nuclear Information System (INIS)

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

    1989-06-01

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

  16. Supply of appropriate nuclear technology for the developing world: small power reactors for electricity generation

    International Nuclear Information System (INIS)

    Heising-Goodman, C.D.

    1981-01-01

    This paper reviews the supply of small nuclear power plants (200 to 500 MWe electrical generating capacity) available on today's market, including the pre-fabricated designs of the United Kingdom's Rolls Royce Ltd and the French Alsthom-Atlantique Company. Also, the Russian VVER-440 conventionally built light-water reactor design is reviewed, including information on the Soviet Union's plans for expansion of its reactor-building capacity. A section of the paper also explores the characteristics of LDC electricity grids, reviewing methods available for incorporating larger plants into smaller grids as the Israelis are planning. Future trends in reactor supply and effects on proliferation rates are also discussed, reviewing the potential of the Indian 220 MWe pressurised heavy-water reactor, South Korean and Jananese potential for reactor exports in the Far East, and the Argentine-Brazilian nuclear programme in Latin America. This study suggests that small reactor designs for electrical power production and other applications, such as seawater desalination, can be made economical relative to diesel technology if traditional scaling laws can be altered by adopting and standardising a pre-fabricated nuclear power plant design. Also, economy can be gained if sufficient attention is concentrated on the design, construction and operating experience of suitably sized conventionally built reactor systems. (author)

  17. Probability of a surface rupture offset beneath a nuclear test reactor

    International Nuclear Information System (INIS)

    Reed, J.W.; Meehan, R.L.; Crellin, G.L.

    1981-01-01

    A probabilistic analysis was conducted to determine the likelihood of a surface rupture offset of any size beneath the 50 megawatt General Electric Test Reactor (GETR), which is located at the Vallecitos Nuclear Center near Pleasanton, California. Geologic faults have been observed at the GETR site. These faults may be due to surface folds, landslides, or deep tectonic movement. They are referred to in the paper as 'existing faults;' however, use of this term does not imply that they are tectonic in origin. The objective of the analysis was to evaluate whether a conservative estimate of the probability of occurrence of a future fault movement is sufficiently low so that movement beneath the reactor building need not be considered as a design basis event. The reactor building is located between two existing faults which are approximately 1320 feet apart. If a fault movement occurs in the future, it is conservatively assumed to occur either on the existing faults or between the faults, or on a fault(s) and between the two faults at the same time. The probabilistic model included the possibility of movements occurring due to unknown, undiscovered faults in the region. For this part, movements were assumed to occur according to a Poisson process. For the possibility of new faults occurring due to the two existing faults, a hazard function was used which increases with time since the last offset. (orig./RW)

  18. Analysis of the General Electric Company swell tests with RELAP4/MOD7

    International Nuclear Information System (INIS)

    Fischer, S.R.; Hendrix, C.E.

    1979-01-01

    The RELAP4/MOD7 nuclear reactor transient analysis code, presently being developed by EG and G Idaho, Inc., will incorporate several significant improvements over earlier versions of RELAP4. As part of the development of RELAP4/MOD7, a thorough assessment of the capability of the code to simulate water reactor LOCA phenomena is being made. This assessment is accomplished in part by comparing results from code calculations with test data from experimental facilities. Simulations of the General Electric Company (GE) level swell tests were performed as part of the code checkout. In these tests, a pressurized vessel partially filled with nearly saturated water was blown down through a simulated break located near the top of the vessel. Comparison of RELAP4 calculations with data from these experiments indicates that the code has the capability to model the unequal phase velocity flow and resulting density gradients that might occur in a BWR steam line break transient. Comparisons of RELAP4 calculations with data from two level swell experiments are presented

  19. Study of reactor Brayton power systems for nuclear electric spacecraft

    Science.gov (United States)

    1979-01-01

    The feasibility of using Brayton power systems for nuclear electric spacecraft was investigated. The primary performance parameters of systems mass and radiator area were determined for systems from 100 to 1000 kW sub e. Mathematical models of all system components were used to determine masses and volumes. Two completely independent systems provide propulsion power so that no single-point failure can jeopardize a mission. The waste heat radiators utilize armored heat pipes to limit meteorite puncture. The armor thickness was statistically determined to achieve the required probability of survival. A 400 kW sub e reference system received primary attention as required by the contract. The components of this system were defined and a conceptual layout was developed with encouraging results. An arrangement with redundant Brayton power systems having a 1500 K (2240 F) turbine inlet temperature was shown to be compatible with the dimensions of the space shuttle orbiter payload bay.

  20. Applications of nuclear reactor power systems to electric propulsion missions.

    Science.gov (United States)

    Schaupp, R. W.; Sawyer, C. D.

    1971-01-01

    The performance of nuclear electric propulsion systems (NEP) has been evaluated for a wide variety of missions in an attempt to establish the commonality of NEP system requirements. Emphasis was given to those requirements and system characteristics that serve as guidelines for current technology development programs. Various interactions and tradeoffs between NEP system and mission parameters are described. The results show that the most significant factors in selecting NEP system size are launch mode (direct or spiral escape) and, to a weaker extent, launch vehicle capability. Other factors such as mission, payload, and thrust time constraints, have little influence, thus allowing one NEP system to be used for many missions. The results indicated that a 100 kWe NEP would be suitable for most direct escape missions and a 250 kWe NEP system would be suitable for more demanding missions that use the spiral escape mode.

  1. 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

  2. Wiring installation for electric devices above the roof slab of a nuclear reactor

    International Nuclear Information System (INIS)

    Jahnke, S.

    1986-01-01

    The wiring installation is situated inside the nuclear reactor building. It includes, associated to electric devices, a first cable which extends from the device to a fixed connector arranged above the cover. A second cable is connected to the said fixed connector and to a connector fixed on a plate situated out of the reactor. According to the present invention each second cable has several sections. A first section can be connected to the said fixed connector situated above the cover and to a fixed lead-in connector of a fluid-tight conduit above the reactor core. A second section is inside the conduit. A third section can be connected to a lead-out connector fixed on the plate which is out of the reactor. The invention applies more particularly to pressurized water nuclear reactors [fr

  3. Nuclear reactor capable of electric power generation during in-service inspection

    International Nuclear Information System (INIS)

    Nakamura, Shinsuke; Nogami, Hitoshi.

    1992-01-01

    The nuclear power plant according to the present invention can generate electric power even in a period when one of a pair of reactors is put to in-service inspection. That is, the nuclear power plant of the present invention comprises a system constitution of two nuclear reactors each of 50% thermal power and one turbine power generator of 100% electric power. Further, facilities of various systems relevant to the two reactors each of 50% thermal power, as a pair, are used in common as much as possible in order to reduce the cost for construction and maintenance/ inspection. Further, a reactor building and a turbine building disposed in adjacent with each for paired two reactors each of 50% thermal power are arranged vertically. This arrangement can facilitate the common use of the facilities for various systems and equipments to attain branching and joining of fluids in reactor feed water systems and main steam system pipelines easily with low pressure loss and low impact shocks. The facility utilization factor of such reactors is remarkably improved by doubling the period of continuous power generation. As a result, economic property is remarkably improved. (I.S.)

  4. Systematic evaluation program review of NRC Safety Topic VI-10.A associated with the electrical, instrumentation and control portions of the testing of reactor trip system and engineered safety features, including response time for the Dresden station, Unit II nuclear power plant

    International Nuclear Information System (INIS)

    St Leger-Barter, G.

    1980-11-01

    This report documents the technical evaluation and review of NRC Safety Topic VI-10.A, associated with the electrical, instrumentation, and control portions of the testing of reactor trip systems and engineered safety features including response time for the Dresden II nuclear power plant, using current licensing criteria

  5. 10 CFR 110.26 - General license for the export of nuclear reactor components.

    Science.gov (United States)

    2010-01-01

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

  6. 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.

  7. Very high temperature measurements: Applications to nuclear reactor safety tests

    International Nuclear Information System (INIS)

    Parga, Clemente-Jose

    2013-01-01

    This PhD dissertation focuses on the improvement of very high temperature thermometry (1100 deg. C to 2480 deg. C), with special emphasis on the application to the field of nuclear reactor safety and severe accident research. Two main projects were undertaken to achieve this objective: - The development, testing and transposition of high-temperature fixed point (HTFP) metal-carbon eutectic cells, from metrology laboratory precision (±0.001 deg. C) to applied research with a reasonable degradation of uncertainties (±3-5 deg. C). - The corrosion study and metallurgical characterization of Type-C thermocouple (service temp. 2300 deg. C) prospective sheath material was undertaken to extend the survivability of TCs used for molten metallic/oxide corium thermometry (below 2000 deg. C)

  8. Nuclear reactor

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  9. General view about reactor safety nuclear power plants in Brazil

    International Nuclear Information System (INIS)

    Gasparian, A.E.; Silva, D.E.; Salvatore, J.E.L.; Lima, J.M. de

    1991-01-01

    In this paper the authors describe the principles and goals that have guided, as well as the methods that have been used by the National Commission of Nuclear Energy (CNEN) to set forth measures aiming at providing safety to the Brazilian nuclear power plants. The status of the licensing process of these power plants is shown. The performance and the results obtained so far in relation to the nuclear safety are also described. (author)

  10. Nuclear reactor

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  11. The Perspective of Small and Medium Size Nuclear Power Reactors in the Brazilian Isolated Electricity Grids

    International Nuclear Information System (INIS)

    Moutinho dos Santos, E.

    1998-01-01

    As a consequence of the international crisis that has shaken the nuclear industry almost all over the world during the late 1980s, new academic contributions, which could settle a new basis for the nuclear energy in Brazil, have been scarce. This paper aims to partially fulfill this gap. We discuss some perspectives for the nuclear option to regain some prominence in the Brazilian energy matrix. Recent developments in the nuclear industry, including advances in plant design, have been opening interesting markets for small and medium nuclear power reactors (SMNPRs). We access the suitability and feasibility of such new technology in the Brazilian isolated electricity grids. We conclude by saying that, although the difficulties are still huge, SMNPRs may be a good strategy for Brazil to revitalize its nuclear policy. (author)

  12. Expert systems for analysis and management assistance of nuclear reactor electrical power supplies

    International Nuclear Information System (INIS)

    Evrard, J.M.; Souchet, Y.

    1988-01-01

    Electrical power supplies of nuclear plants are very complex systems. They are studied in many ways: failure consequences, probabilistic risk assessement, failure diagnosis, corrective actions in case of incident ... Knowledge base technology (expert systems) is very suited to solve these problems. A common structural representation can generate specific functional representations; thus we get coherency and easy evolution. This paper shows a two facet methodology: the plant description (both static and dynamic) and reasoning about it. Current applications are developed for pressurized water reactors and gas cooled reactors, using the SPIRAL expert system shell buil at CEA [fr

  13. Nuclear Reactor Physics

    Science.gov (United States)

    Stacey, Weston M.

    2001-02-01

    An authoritative textbook and up-to-date professional's guide to basic and advanced principles and practices Nuclear reactors now account for a significant portion of the electrical power generated worldwide. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. Nuclear reactor physics is the core discipline of nuclear engineering, and as the first comprehensive textbook and reference on basic and advanced nuclear reactor physics to appear in a quarter century, this book fills a large gap in the professional literature. Nuclear Reactor Physics is a textbook for students new to the subject, for others who need a basic understanding of how nuclear reactors work, as well as for those who are, or wish to become, specialists in nuclear reactor physics and reactor physics computations. It is also a valuable resource for engineers responsible for the operation of nuclear reactors. Dr. Weston Stacey begins with clear presentations of the basic physical principles, nuclear data, and computational methodology needed to understand both the static and dynamic behaviors of nuclear reactors. This is followed by in-depth discussions of advanced concepts, including extensive treatment of neutron transport computational methods. As an aid to comprehension and quick mastery of computational skills, he provides numerous examples illustrating step-by-step procedures for performing the calculations described and chapter-end problems. Nuclear Reactor Physics is a useful textbook and working reference. It is an excellent self-teaching guide for research scientists, engineers, and technicians involved in industrial, research, and military applications of nuclear reactors, as well as government regulators who wish to increase their understanding of nuclear reactors.

  14. Nuclear reactor shutdown system

    International Nuclear Information System (INIS)

    Mangus, J.D.; Cooper, M.H.

    1982-01-01

    An improved nuclear reactor shutdown system is described comprising a temperature sensitive device connected to control the electric power supply to a magnetic latch holding a body of a neutron absorbing material. The temperature sensitive device is exposed to the reactor coolant so that when the reactor coolant temperature rises above a specific level, the temperature sensitive device will cause deenergization of the magnetic latch to allow the body of neutron absorbing material to enter the reactor core. (author)

  15. IEEE Std 383-1974: IEEE standard for type test of Class IE electric cables, field splices, and connections for nuclear power generating stations

    International Nuclear Information System (INIS)

    Anon.

    1992-01-01

    This standard provides direction for establishing type tests which may be used in qualifying Class 1E electric cables, field splices, and other connections for service in nuclear power generating stations. General guidelines for qualifications are given in IEEE Std 323-1974, Standard for Qualifying Class IE Electric Equipment for Nuclear Power Generating Stations. Categories of cables covered are those used for power control and instrumentation services. Though intended primarily to pertain to cable for field installation, this guide may also be used for the qualification of internal wiring of manufactured devices. This guide does not cover cables for service within the reactor vessel

  16. Heat-electricity convertion systems for a Brazilian space micro nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Guimaraes, Lamartine N.F.; Marcelino, Natalia B.; Placco, Guilherme M.; Nascimento, Jamil A.; Borges, Eduardo M., E-mail: guimarae@ieav.cta.br, E-mail: lamartine.guimaraes@pq.cnpq.br, E-mail: jamil@ieav.cta.br, E-mail: jalnsgf@outlook.com, E-mail: borges.em@hotmail.com, E-mail: ecorborges@hotmail.com, E-mail: ivayolini@gmail.com, E-mail: guilherme_placco@ig.com.br [Instituto de Estudos Avancados (IEAv/DCTA), Sao Jose dos Campos, SP (Brazil); Barrios Junior, Ary Garcia, E-mail: arygarcia89@yahoo.com [Faculdade de Tecnologia Sao Francisco (FATESF), Jacarei, SP (Brazil)

    2013-07-01

    This contribution will discuss the evolution work in the development of thermal cycles to allow the development of heat-electricity conversion for the Brazilian space micro nuclear Reactor. Namely, innovative core and nuclear fuel elements, Brayton cycle, Stirling engine, heat pipes, passive multi-fluid turbine, among others. This work is basically to set up the experimental labs that will allow the specification and design of the space equipment. Also, some discussion of the cost so far, and possible other applications will be presented. (author)

  17. Heat-electricity convertion systems for a Brazilian space micro nuclear reactor

    International Nuclear Information System (INIS)

    Guimaraes, Lamartine N.F.; Marcelino, Natalia B.; Placco, Guilherme M.; Nascimento, Jamil A.; Borges, Eduardo M.; Barrios Junior, Ary Garcia

    2013-01-01

    This contribution will discuss the evolution work in the development of thermal cycles to allow the development of heat-electricity conversion for the Brazilian space micro nuclear Reactor. Namely, innovative core and nuclear fuel elements, Brayton cycle, Stirling engine, heat pipes, passive multi-fluid turbine, among others. This work is basically to set up the experimental labs that will allow the specification and design of the space equipment. Also, some discussion of the cost so far, and possible other applications will be presented. (author)

  18. Electric failure on the reactor n.3 of the nuclear power plant of Dampierre

    International Nuclear Information System (INIS)

    2007-05-01

    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.)

  19. TRAC development at General Electric

    International Nuclear Information System (INIS)

    Andersen, J.G.M.; Shaug, J.C.; Shiralkar, B.S.

    1987-01-01

    TRAC is a computer code for transient analysis of light water reactors. The BWR version of TRAC has been developed as a result of a close cooperation between General Electric Company and Idaho National Engineering Laboratory. Up through 1985 the development work at General Electric was jointly funded by General Electric, the Nuclear Regulatory Commission and Electric Power Research Institute under the Refill-Reflood and FIST programs. At INEL (which has the main responsibility for the NRC version of TRAC-BWR) this work has led to the development of TRACBD1 and TRACBF1, while at GE, TRACB04 was the final product of the Refill-Reflood and FIST programs. TRAC development has continued at General Electric after the completion of these programs with the evolution of the TRACG code. The purpose of the paper is to describe this work. The TRAC development at General Electric can be divided into two main categories: extended benchmark capability and improved user convenience

  20. Combined development of international nuclear fusion test reactors

    International Nuclear Information System (INIS)

    Anon.

    1987-01-01

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

  1. Nuclear reactor

    International Nuclear Information System (INIS)

    Tilliette, Z.

    1975-01-01

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

  2. Nuclear reactor

    International Nuclear Information System (INIS)

    Rau, P.

    1980-01-01

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

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

    International Nuclear Information System (INIS)

    Touati, Said; Chennai, Salim; Souli, Aissa

    2015-01-01

    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)

  4. 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)

  5. 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.

  6. Small nuclear power reactor emergency electric power supply system reliability comparative analysis

    International Nuclear Information System (INIS)

    Bonfietti, Gerson

    2003-01-01

    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)

  7. Nuclear start-up, testing and core management of the Fast Test Reactor (FTR)

    International Nuclear Information System (INIS)

    Bennett, R.A.; Daughtry, J.W.; Harris, R.A.; Jones, D.H.; Nelson, J.V.; Rawlins, J.A.; Rothrock, R.B.; Sevenich, R.A.; Zimmerman, B.D.

    1980-01-01

    Plans for the nuclear start-up, low and high power physics testing, and core management of the Fast Test Reactor (FTR) are described. Owing to the arrangement of the fuel-handling system, which permits continuous instrument lead access to experiments during refuelling, it is most efficient to load the reactor in an asymmetric fashion, filling one-third core sectors at a time. The core neutron level will be monitored during this process using both in-core and ex-core detectors. A variety of physics tests are planned following the core loading. Because of the experimental purpose of the reactor, these tests will include a comprehensive characterization programme involving both active and passive neutron and gamma measurements. Following start-up tests, the FTR will be operated as a fast neutron irradiation facility, to test a wide variety of fast reactor core components and materials. Nuclear analyses will be made prior to each irradiation cycle to confirm that the planned arrangement of standard and experimental components satisfies all safety and operational constraints, and that all experiments are located so as to achieve their desired irradiation environment. (author)

  8. Utilization of nuclear energy for generating electric power in the FRG, with special regard to LWR-type reactors

    International Nuclear Information System (INIS)

    Vollradt, J.

    1977-01-01

    Comments on interdependencies in energy industry and energy generation as seen by energy supply utilities, stating that the generation of electric power in Germany can only be based on coal and nuclear energy in the long run, are followed by the most important, fundamental, nuclear-physical, technological and in part political interdependencies prevailing in the starting situation of 1955/58 when the construction of nuclear power plant reactors began. Then the development ranging to the 28000 MW nuclear power output to be expected in 1985 is outlined, totalling in 115000 MW electric power in the FRG. Finally, using the respectively latest order, the technical set up of each of the reactor types with 1300 MWe unit power offered by German manufacturers are described: BBC/BBR PWR-type reactor Neupotz, KWU-PWR-type reactor Hamm and KWU PWR-type reactor double unit B+C Gundremmingen. (orig.) [de

  9. Advanced marine reactor MRX and application to nuclear barge supplying electricity and heat

    International Nuclear Information System (INIS)

    Ishida, Toshihisa; Kusunoki, Tsuyoshi; Odano, Naoteru; Yoritsune, Tsutomu; Fukuhara, Yoshifumi; Ochiai, Masa-aki

    2000-01-01

    The basic design concept of an advanced marine reactor MRX has been established with adoption of several new technologies. The MRX is an integral-type PWR with 100 MWt aimed basically for use of ship propulsion. Adoption of a water-filled containment together with the integral type reactor makes the reactor light-weight and compact greatly. A engineered safety system is a simplified passive system, function of which is confirmed by the safety analysis. The MRX can be applied to an energy supply system of electricity and heat co-generation by installing it on a barge. Concept of a nuclear barge with the MRX of 334 MWt output is presented for use of supplying electricity, fresh water and hot water. Combined system of electric generation and desalination with the RO process can deliver variable output of electricity and fresh water according a demand. Latent heat of the exhausted steam from the turbine can be used effectively to raise the temperature of cold water as heat supply. (author)

  10. Reactor type choice and characteristics for a small nuclear heat and electricity co-generation plant

    International Nuclear Information System (INIS)

    Liu Kukui; Li Manchang; Tang Chuanbao

    1997-01-01

    In China heat supply consumes more than 70 percent of the primary energy resource, which makes for heavy traffic and transportation and produces a lot of polluting materials such as NO x , SO x and CO 2 because of use of the fossil fuel. The utilization of nuclear power into the heat and electricity co-generation plant contributes to the global environmental protection. The basic concept of the nuclear system is an integral type reactor with three circuits. The primary circuit equipment is enclosed in and linked up directly with reactor vessel. The third circuit produces steam for heat and electricity supply. This paper presents basic requirements, reactor type choice, design characteristics, economy for a nuclear co-generation plant and its future application. The choice of the main parameters and the main technological process is the key problem of the nuclear plant design. To make this paper clearer, take for example a double-reactor plant of 450 x 2MW thermal power. There are two sorts of main technological processes. One is a water-water-steam process. Another is water-steam-steam process. Compared the two sorts, the design which adopted the water-water-steam technological process has much more advantage. The system is simplified, the operation reliability is increased, the primary pressure reduces a lot, the temperature difference between the secondary and the third circuits becomes larger, so the size and capacity of the main components will be smaller, the scale and the cost of the building will be cut down. In this design, the secondary circuit pressure is the highest among that of the three circuits. So the primary circuit radioactivity can not leak into the third circuit in case of accidents. (author)

  11. The HTR-10 test reactor project and potential use of HTGR for non-electric application in China

    International Nuclear Information System (INIS)

    Sun Yuliang; Zhong Daxin; Xu Yuanhui; Wu Zhongxin

    1997-01-01

    Coal is the dominant source of energy in China. This use of coal results in two significant problems for China; it is a major burden on the train, road and waterway transportation infrastructures and it is a significant source of environmental pollution. In order to ease the problems caused by the burning of coal and to help reduce the energy supply shortage in China, national policy has directed the development of nuclear power. This includes the erection of nuclear power plants with water cooled reactors and the development of advanced nuclear reactor types, specifically, the high temperature gas cooled reactor (HTGR). The HTGR was chosen for its favorable safety features and its ability to provide high reactor outlet coolant temperatures for efficient power generation and high quality process heat for industrial applications. As the initial modular HTGR development activity within the Chinese High Technology Programme, a 10MW helium cooled test reactor is currently under construction on the site of the Institute of Nuclear Energy Technology northwest of Beijing. This plant features a pebble-bed helium cooled reactor with initial criticality anticipated in 1999. There will be two phases of high temperature heat utilization from the HTR-10. The first phase will utilize a reactor outlet temperature of 700 deg. C with a steam generator providing steam for a steam turbine cycle which works on an electrical/heat co-generation basis. The second phase is planned for a core outlet temperature of 900 deg. C to investigate a steam cycle/gas turbine combined cycle system with the gas turbine and the steam cycle being independently parallel in the secondary side of the plant. This paper provides a review of the technical design, licensing, safety and construction schedule for the HTR-10. It also addresses the potential uses of the HTGR for non-electric applications in China including process steam for the petrochemical industry, heavy oil recovery, coal conversion and

  12. Nuclear reactors

    International Nuclear Information System (INIS)

    Prescott, R.F.

    1976-01-01

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

  13. 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.

  14. Characterization and testing of materials for nuclear reactors. Proceedings of a technical meeting

    International Nuclear Information System (INIS)

    2007-03-01

    Nuclear techniques in general and neutrons based methods in particular have played and will continue to play an important role in research in materials science and technology. Today the world is looking at nuclear fission and nuclear fusion as the main sources of energy supply for the future. Research reactors have played a key role in the development of nuclear technology. A materials development programme will thus play a major role in the design and development of new nuclear power plants, for the extension of the life of operating reactors as well as for fusion reactors. Against this background, the IAEA had organized a Technical Meeting on Development, Characterization and Testing of Materials - With Special Reference to the Energy Sector under the activity on specific applications of research reactors. The meeting was held in Vienna, May 29- June 2, 2006. There was also participation by experts in techniques, complementary to neutrons. The participants for the technical meeting were experts in the utilization of nuclear techniques namely the high flux and medium flux research reactors, fusion research and positron annihilation. They presented the design, development and utilization of the facilities at their respective centres for materials characterization with main focus on materials for nuclear energy, both fission and fusion. In core irradiation of materials, development of instrument for residual stress measurement in large and / or irradiated specimen, neutron radiography for inspection of irradiated fuel, work on oxide dispersion strengthened (ODS) steels and SiC composites, relevant to future power systems were cited as application of nuclear techniques in fission reactors. The use of neutron scattering for helium bubbles in steel, application of positron annihilation to study helium bubbles in Cu, Ti-stabilized stainless steel and voidswelling studies etc. show that these techniques have an important role in the development of materials for energy

  15. Online calibration method for condition monitoring of nuclear reactor instrumentations based on electrical signature analysis

    International Nuclear Information System (INIS)

    Syaiful Bakhri

    2013-01-01

    Electrical signature analysis currently becomes an alternative in condition monitoring in nuclear power plants not only for stationary components such as sensors, measurement and instrumentation channels, and other components but also for dynamic components such as electric motors, pumps, generator or actuators. In order to guarantee the accuracy, the calibration of monitoring system is a necessary which practically is performed offline, under limited schedules and certain tight procedures. This research aims to introduce online calibration technique for electrical signature condition monitoring in order that the accuracy can be maintained continuously which in turn increases the reactor safety as a whole. The research was performed step by stepin detail from the conventional technique, online calibration using baseline information and online calibration using differential gain adjustment. Online calibration based on differential gain adjustment provides better results than other techniques even tough under extreme gain insertion as well as external disturbances such as supply voltages. (author)

  16. On the classification of structures, systems and components of nuclear research and test reactors

    International Nuclear Information System (INIS)

    Mattar Neto, Miguel

    2009-01-01

    The classification of structures, systems and components of nuclear reactors is a relevant issue related to their design because it is directly associated with their safety functions. There is an important statement regarding quality standards and records that says Structures, systems, and components important to safety shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed. The definition of the codes, standards and technical requirements applied to the nuclear reactor design, fabrication, inspection and tests may be seen as the main result from this statement. There are well established guides to classify structures, systems and components for nuclear power reactors such as the Pressurized Water Reactors but one can not say the same for nuclear research and test reactors. The nuclear reactors safety functions are those required to the safe reactor operation, the safe reactor shutdown and continued safe conditions, the response to anticipated transients, the response to potential accidents and the control of radioactive material. So, it is proposed in this paper an approach to develop the classification of structures, systems and components of these reactors based on their intended safety functions in order to define the applicable set of codes, standards and technical requirements. (author)

  17. Nuclear reactor

    International Nuclear Information System (INIS)

    Scholz, M.

    1976-01-01

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

  18. Nuclear reactor container

    International Nuclear Information System (INIS)

    Hosaka, Seiichi.

    1988-01-01

    Cables coverd with non-halogen covering material are used as electric wire cables wired for supplying electric power to a reactor recycling pump. Silicone rubber having specified molecular formula is used for the non-halogen covering material. As a result, formation of chlorine in a nuclear reactor container can be eliminated and increase in the deposited salts to SUS pipeways, etc. can be prevented, to avoid the occurrence of stress corrosion cracks. (H.T.)

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

    International Nuclear Information System (INIS)

    Chung, Moon Ki; Park, Chun Kyeong; Yang, Seon Kyu; Chung, Chang Hwan; Chun, Shee Yeong; Song, Cheol Hwa; Chun, Hyeong Gil; Chang, Seok Kyu; Chung, Heung Joon; Won, Soon Yeon; Cho, Yeong Ro; Kim, Bok Deuk; Min, Kyeong Ho

    1994-07-01

    The present research aims at the development of the thermal hydraulic verification test technology for the reactor safety system of the conventional and advanced nuclear power plant and the development of the advanced thermal hydraulic measuring techniques. (Author)

  20. Nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    1978-05-10

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

  1. Nuclear reactors

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  2. Alteration in reactor installation (addition of Unit 2) in Shimane Nuclear Power Station, Chugoku Electric Power Co., Inc. (inquiry)

    International Nuclear Information System (INIS)

    1983-01-01

    An inquiry was made by the Ministry of International Trade and Industry to Nuclear Safety Commission on the addition of Unit 2 in Shimane Nuclear Power Station of The Chugoku Electric Power Co., Inc., concerning the technical capability of Chugoku Electric Power Co., Inc., and the plant safety. The NSC requested the Committee on Examination of Reactor Safety to make a deliberation on this subject. Both the technical capability and the safety of Unit 1 were already confirmed by MITI. Unit 2 to be newly added in the Shimane Nuclear Power Station is a BWR power plant with electric output of 820 MW. The examination made by MITI is described: the technical capability of Chugoku Electric Power Co., Inc., the safety of Unit 2 about its siting, reactor proper, reactor cooling system, radioactive waste management, etc. (J.P.N.)

  3. Nuclear reactor

    International Nuclear Information System (INIS)

    Sasaki, Tomozo.

    1987-01-01

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

  4. Nuclear reactors

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  5. Nuclear reactor types

    International Nuclear Information System (INIS)

    Jones, P.M.S.

    1987-01-01

    The characteristics of different reactor types designed to exploit controlled fission reactions are explained. Reactors vary from low power research devices to high power devices especially designed to produce heat, either for direct use or to produce steam to drive turbines to generate electricity or propel ships. A general outline of basic reactors (thermal and fast) is given and then the different designs considered. The first are gas cooled, including the Magnox reactors (a list of UK Magnox stations and reactor performance is given), advanced gas cooled reactors (a list of UK AGRs is given) and the high temperature reactor. Light water cooled reactors (pressurized water [PWR] and boiling water [BWR] reactors) are considered next. Heavy water reactors are explained and listed. The pressurized heavy water reactors (including CANDU type reactors), boiling light water, steam generating heavy water reactors and gas cooled heavy water reactors all come into this category. Fast reactors (liquid metal fast breeder reactors and gas cooled fast reactors) and then water-cooled graphite-moderated reactors (RBMK) (the type at Chernobyl-4) are discussed. (U.K.)

  6. Alteration in reactor installations (Unit 1 and 2 reactor facilities) in the Hamaoka Nuclear Power Station of The Chubu Electric Power Co., Inc. (report)

    International Nuclear Information System (INIS)

    1982-01-01

    A report by the Nuclear Safety Commission to the Ministry of International Trade and Industry concerning the alteration in Unit 1 and 2 reactor facilities in the Hamaoka Nuclear Power Station, Chubu Electric Power Co., Inc., was presented. The technical capabilities for the alteration of reactor facilities in Chubu Electric Power Co., Inc., were confirmed to be adequate. The safety of the reactor facilities after the alteration was confirmed to be adequate. The items of examination made for the confirmation of the safety are as follows: reactor core design (nuclear design, mechanical design, mixed reactor core), the analysis of abnormal transients in operation, the analysis of various accidents, the analysis of credible accidents for site evaluation. (Mori, K.)

  7. Refurbish research and test reactors corresponding to global age of nuclear energy

    International Nuclear Information System (INIS)

    Mishima, Kaichiro; Oyama, Yukio; Okamoto, Koji; Yamana, Hajime; Yamaguchi, Akira

    2011-01-01

    This special article featured arguments for refurbishment of research and test reactors corresponding to global age of nuclear energy, based on the report: 'Investigation of research facilities necessary for future joint usage' issued by the special committee of Atomic Energy Society of Japan (AESJ) in September 2010. It consisted of six papers titled as 'Introduction-establishment of AESJ special committee for investigation', 'State of research and test reactors in Japan', 'State of overseas research and test reactors', 'Needs analysis for research and test reactors', 'Proposal of AESJ special committee' and 'Summary and future issues'. In order to develop human resources and promote research and development needed in global age of nuclear energy, research and test reactors would be refurbished as an Asian regional center of excellence. (T. Tanaka)

  8. On exposure of workers in nuclear reactor facilities for test and in nuclear reactor facilities in research and development stage in fiscal 1988

    International Nuclear Information System (INIS)

    1989-01-01

    The Law for Regulation on Nuclear Reactor requires the operators of nuclear reactors that the exposure dose of workers engaged in work for nuclear reactors should not exceed the limits specified in official notices that are issued based on the Law. The present article summarizes the contents of the Report on Radiation Management in 1988 submitted by the operators of nuclear reactor facilities for test and those of nuclear reactor facilities in research and development stage based on the Law, and the Report on Management of Exposure Dose of Workers submitted by them based on administrative notices. The reports demonstrate that the exposure of workers was below the permissible exposure dose in 1988 in all nuclear reactor facilities. The article presents data on the distribution of exposure dose among workers in all facilities with a nuclear reactor for test, and data on personal exposure of employees and non-employees and overall exposure of all workers in the facilities of Japan Atomic Energy Research Institute and Power Reactor and Nuclear Fuel Development Corporation. (N.K.)

  9. 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.

  10. The Nuclear option for U.S. electrical generating capacity additions utilizing boiling water reactor technology

    International Nuclear Information System (INIS)

    Garrity, T.F.; Wilkins, D.R.

    1993-01-01

    The technology status of the Advanced Boiling Water (ABWR) and Simplified Boiling Water (SBWR) reactors are presented along with an analysis of the economic potential of advanced nuclear power generation systems based on BWR technology to meet the projected domestic electrical generating capacity need through 2005. The forecasted capacity needs are determined for each domestic North American Electric Reliability Council (NERC) region. Extensive data sets detailing each NERC region's specific generation and load characteristics, and capital and fuel cost parameters are utilized in the economic analysis of the optimal generation additions to meet this need by use of an expansion planning model. In addition to a reference case, several sensitivity cases are performed with regard to capital costs and fuel price escalation

  11. Benchmark test of evaluated nuclear data files for fast reactor neutronics application

    International Nuclear Information System (INIS)

    Chiba, Go; Hazama, Taira; Iwai, Takehiko; Numata, Kazuyuki

    2007-07-01

    A benchmark test of the latest evaluated nuclear data files, JENDL-3.3, JEFF-3.1 and ENDF/B-VII.0, has been carried out for fast reactor neutronics application. For this benchmark test, experimental data obtained at fast critical assemblies and fast power reactors are utilized. In addition to comparing of numerical solutions with the experimental data, we have extracted several cross sections, in which differences between three nuclear data files affect significantly numerical solutions, by virtue of sensitivity analyses. This benchmark test concludes that ENDF/B-VII.0 predicts well the neutronics characteristics of fast neutron systems rather than the other nuclear data files. (author)

  12. Application of IAEA's International Nuclear Event Scale to events at testing/research reactors in Japan

    International Nuclear Information System (INIS)

    Nozawa, Masao; Watanabe, Norio

    1999-01-01

    The International Nuclear Event Scale (INES) is a means for providing prompt, clear and consistent information related to nuclear events and facilitating communication between the nuclear community, the media and the public on such events. This paper describes the INES rating process for events at testing/research reactors and nuclear fuel processing facilities and experience on the application of the INES scale in Japan. (author)

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

    International Nuclear Information System (INIS)

    Takeda, Takeshi; Nakagawa, Shigeaki; Tachibana, Yukio; Takada, Eiji; Kunitomi, Kazuhiko

    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)

  14. Nuclear reactor

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  15. Nuclear reactor

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  16. Concerning control of radiation exposure to workers in nuclear reactor facilities for testing and nuclear reactor facilities in research and development phase (fiscal 1987)

    International Nuclear Information System (INIS)

    1988-01-01

    A nuclear reactor operator is required by the Nuclear Reactor Control Law to ensure that the radiation dose to workers engaged in the operations of his nuclear reactor is controlled below the permissible exposure doses that are specified in notifications issued based on the Law. The present note briefly summarizes the data given in the Reports on Radiation Control, which have been submitted according to the Nuclear Reactor Control Law by the operators of nuclear reactor facilities for testing and those in the research and development phase, and the Reports on Control of Radiation Exposure to Workers submitted in accordance with the applicable administrative notices. According to these reports, the measured exposure to workers in 1987 were below the above-mentioned permissible exposure doses in all these nuclear facilities. The 1986 and 1987 measurements of radiation exposure dose to workers in nuclear reactor facilities for testing are tabulated. The measurements cover dose distribution among the facilities' personnel and workers of contractors. They also cover the total exposure dose for all workers in each of four plants operated under the Japan Atomic Energy Research Institute and the Power Reactor and Nuclear Fuel Development Corporation. (N.K.)

  17. Non Nuclear Testing of Reactor Systems In The Early Flight Fission Test Facilities (EFF-TF)

    International Nuclear Information System (INIS)

    Van Dyke, Melissa; Martin, James

    2004-01-01

    The Early Flight Fission-Test Facility (EFF-TF) can assist in the design and development of systems through highly effective non-nuclear testing of nuclear systems when technical issues associated with near-term space fission systems are 'non-nuclear' in nature (e.g. system's nuclear operations are understood). For many systems, thermal simulators can be used to closely mimic fission heat deposition. Axial power profile, radial power profile, and fuel pin thermal conductivity can be matched. In addition to component and subsystem testing, operational and lifetime issues associated with the steady state and transient performance of the integrated reactor module can be investigated. Instrumentation at the EFF-TF allows accurate measurement of temperature, pressure, strain, and bulk core deformation (useful for accurately simulating nuclear behavior). Ongoing research at the EFF-TF is geared towards facilitating research, development, system integration, and system utilization via cooperative efforts with DOE laboratories, industry, universities, and other Nasa centers. This paper describes the current efforts for the latter portion of 2003 and beginning of 2004. (authors)

  18. Nuclear reactor

    International Nuclear Information System (INIS)

    Anthony, A.J.; Gruber, E.A.

    1979-01-01

    A nuclear reactor with control rods in channels between fuel assemblies wherein the fuel assemblies incorporate guide rods which protrude outwardly into the control rod channels to prevent the control rods from engaging the fuel elements. The guide rods also extend back into the fuel assembly such that they are relatively rigid members. The guide rods are tied to the fuel assembly end or support plates and serve as structural members which are supported independently of the fuel element. Fuel element spacing and support means may be attached to the guide rods. 9 claims

  19. Data on loss of off-site electric power simulation tests of the high temperature engineering test reactor

    International Nuclear Information System (INIS)

    Takeda, Takeshi; Nakagawa, Shigeaki; Fujimoto, Nozomu; Tachibana, Yukio; Iyoku, Tatsuo

    2002-07-01

    The high temperature engineering test reactor (HTTR), the first high temperature gas-cooled reactor (HTGR) in Japan, achieved the first full power of 30 MW on December 7 in 2001. In the rise-to-power test of the HTTR, simulation tests on loss of off-site electric power from 15 and 30 MW operations were carried out by manual shutdown of off-site electric power. Because helium circulators and water pumps coasted down immediately after the loss of off-site electric power, flow rates of helium and water decreased to the scram points. To shut down the reactor safely, the subcriticality should be kept by the insertion of control rods and the auxiliary cooling system should cool the core continuously avoiding excessive cold shock to core graphite components. About 50 s later from the loss of off-site electric power, the auxiliary cooling system started up by supplying electricity from emergency power feeders. Temperature of hot plenum block among core graphite structures decreased continuously after the startup of the auxiliary cooling system. This report describes sequences of dynamic components and transient behaviors of the reactor and its cooling system during the simulation tests from 15 and 30 MW operations. (author)

  20. On exposure management of workers in nuclear reactor facilities for test and in nuclear reactor facilities in research and development stage in fiscal 1993

    International Nuclear Information System (INIS)

    1994-01-01

    The Law of Regulation on Nuclear Reactor requires the operators of nuclear reactors that the exposure dose of workers engaged in work for nuclear reactors should not exceed the limits specified in official notices that are issued based on the Law. The present article summarizes the contents of the Report on Radiation Management in 1993 submitted by the operators of nuclear reactor facilities for test and those of nuclear reactor facilities in research and development stage based on the Law, and the Report on Management of Exposure Dose of Workers submitted by them based on administrative notices. The reports demonstrate that the the exposure of workers was below the permissible exposure dose in 1993 in all nuclear reactor facilities. The article presents data on the distribution of exposure dose among workers in all facilities with a nuclear reactor for test, and data on personal exposure of employees and non-employees and overall exposure of all workers in the facilities of JAERI and PNC. (J.P.N.)

  1. Nuclear reactors

    International Nuclear Information System (INIS)

    Yoshioka, Michiko.

    1985-01-01

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

  2. Nuclear electronic equipment for control and monitoring panel. Procedure guide for on-site tests of nuclear reactor instruments

    International Nuclear Information System (INIS)

    1975-10-01

    By the use of a procedure for on-site testing of nuclear reactor instruments it should be possible to judge their ability to guarantee the reactor safety and availability at the moment of divergence or during operation. Such a procedure must therefore be created as a work implement for the quick and reliable installation of electronic devices necessary for nuclear reactor control and supervision. A standard document is proposed for this purpose, allowing a ''test programme'' to be set up before the equipment is installed on the site [fr

  3. Nuclear reactors

    International Nuclear Information System (INIS)

    Matheson, J.E.

    1983-01-01

    A nuclear reactor has an upper and a lower grid plate. Protrusions project from the upper grid plate. Fuel assemblies having end fittings fit between the grid plates. An arrangement is provided for accepting axial forces generated during the operation of the nuclear reactor by the flow of the cooling medium and thermal expansion and irradiation-induced growth of the fuel assembly, which comprises rods. Each fuel assembly rests on the lower grid plate and its upper end is elastically supported against the upper grid plate by the above-mentioned arrangement. The arrangement comprises four (for example) torsion springs each having a torsion tube and a torsion bar nested within the torsion tube and connected at one end thereto. The other end of the torsion bar is connected to an associated one of four lever arms. The torsion tube is rigidly connected to the other end fitting and the springs are disposed such that the lever arms are biassed against the protrusions. (author)

  4. Welding of sule elements for nuclear reactors with solid state YAG laser using instrumentated testing equipments

    International Nuclear Information System (INIS)

    Bourgault, F.; Lacoste, J.; Schley, R.; Kluzinski, C.; Piednoir, P.

    1985-09-01

    The instrumentation of the equipment for carrying out safety tests on fuel elements for nuclear reactors requires special thermocouples adapted to the prevailing agressive medium. The investigations described deal essentially with the operational and metallurgical weldability tests out on the safety test zircaloy piping in the pressurized water circuit (PHEBUS-programme) [fr

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

    International Nuclear Information System (INIS)

    Wei, T.; Tobias, M.

    1974-03-01

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

  6. 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.

  7. Light-water nuclear reactors

    International Nuclear Information System (INIS)

    Drevon, G.

    1983-01-01

    This work gives basic information on light-water reactors which is advanced enough for the reader to become familiar with the essential objectives and aspects of their design, their operation and their insertion in the industrial, economic and human environment. In view of the capital role of electric energy in the modern economy a significant place is given to electron-nuclear power stations, particularly those of the type adopted for the French programme. The work includes sixteen chapters. The first chapter relates the history and presents the various applications of light water reactors. The second refers to the general elementary knowledge of reactor physics. The third chapter deals with the high power light-water nuclear power station and thereby introduces the ensuing chapters which, up to and including chapter 13, are devoted to the components and the various aspects of the operation of power stations, in particular safety and the relationship with the environment. Chapter 14 provides information on the reactors adapted to applications other than the generation of electricity on an industrial scale. Chapter 15 shows the extent of the industrial effort devoted to light-water reactors and chapter 16 indicates the paths along which the present work is preparing the future of these reactors. The various chapters have been written to allow for separate consultation. An index of the main technical terms and a bibliography complete the work [fr

  8. 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

  9. Nuclear reactor

    International Nuclear Information System (INIS)

    Schweiger, F.; Glahe, E.

    1976-01-01

    In a nuclear reactor of the kind which is charged with spherical reaction elements and in which control rods are arranged to be thrust directly into the charge, each control rod has at least one screw thread on its external surface so that as the rod is thrust into the charge it is caused to rotate and thus make penetration easier. The length of each control rod may have two distinct portions, a latter portion which carries a screw thread and a lead-in portion which is shorter than the latter portion and which may carry a thread of greater pitch than that on the latter portion or may have a number of axially extending ribs instead of a thread

  10. Computer-aided testing and operational aids for PARR-1 nuclear reactor

    International Nuclear Information System (INIS)

    Ansari, S.A.

    1990-01-01

    The utilization of the plant computer of Pakistan Research Reactor (PARR-1) for automatic periodic testing of nuclear instrumentation in the reactor is described. Computer algorithms have been developed for on-line acquisition and real-time processing of nuclear channel signals. The mean value, standard deviation, and probability distributions of nuclear channel signals are obtained in real time, and the computer generates a warning message if the signal error exceeds the maximum permissible error. In this way a faulty channel is automatically identified. Other real-time algorithms are also described that assist the operator in safe reactor operation by automatically computing approach-to-criticality during reactor start-up and the control rod worth determination

  11. Nuclear reactors

    International Nuclear Information System (INIS)

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

    1979-01-01

    In a nuclear reactor (e.g. one having coolant down-flow through a core to a hearth below) thermal insulation (e.g. of a floor of the hearth) comprises a layer of bricks and a layer of tiles thereon, with smaller clearances between the tiles than between the bricks but with the bricks being of reduced cross-section immediately adjacent the tiles so as to be surrounded by interconnected passages, of relatively large dimensions, constituting a continuous chamber extending behind the layer of tiles. By this arrangement, lateral coolant flow in the inter-brick clearances is much reduced. The reactor core is preferably formed of hexagonal columns, supported on diamond-shaped plates each supported on a pillar resting on one of the hearth-floor tiles. Each plate has an internal duct, four upper channels connecting the duct with coolant ducts in four core columns supported by the plate, and lower channels connecting the duct to a downwardly-open recess common to three plates, grouped to form a hexagon, at their mutually-adjacent corners. This provides mixing, and temperature-averaging, of coolant from twelve columns

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

    International Nuclear Information System (INIS)

    Banciu, Ortenzia; Vladescu, Gabriela

    2003-01-01

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

  13. 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.)

  14. Nuclear reactors

    International Nuclear Information System (INIS)

    Pearson, K.G.

    1977-01-01

    Reference is made to auxiliary means of cooling the nuclear fuel clusters used in light or heavy water cooled nuclear reactors. One method is to provide one or more spray cooling tubes. From holes in the side walls of those tubes coolant water may be sprayed laterally into the cluster against the rods. The flow of main coolant may thus be supplemented or even replaced by the auxiliary coolant. A difficulty, however, is that only those fuel rods close to a spray cooling tube can readily be reached by the auxiliary coolant. In the arrangement described, where the fuel rods are spaced apart by transverse grids, at least one of the interspaces between the grids is provided with an axially extending auxiliary coolant conduit having lateral holes through which an auxiliary coolant is sprayed into the cluster. A deflector is provided that extends from a transverse grid into a position in front of the holes and deflects auxiliary coolant on to parts of the fuel rods otherwise inaccessible to the auxiliary coolant. The construction of the deflector is described. (U.K.)

  15. The feature of high flux engineering test reactor and its role in nuclear power development

    International Nuclear Information System (INIS)

    Lu Guangquan

    1987-01-01

    The High Flux Engineering Test Reactor (HFETR) designed and built by Chinese own efforts reached to its initial criticality on Dec. 27, 1979, and then achieved high power operation on Dec. 16, 1980. Until Nov. 11. 1986, the reactor had been operated for thirteen cycles. The paper presents briefly main feature of HFETR and its utilization during past years. The paper also deals with its role in nuclear power development. Finally, author gives his opinion on comprehensive utilization of HFETR. (author)

  16. Assessment of two small-sized innovative nuclear reactors for electricity generation in Brazil using INPRO methodology

    International Nuclear Information System (INIS)

    Goncalves Filho, Orlando Joao Agostinho; Sefidvash, Farhang

    2009-01-01

    This paper presents the main results of the assessment study of two small-sized innovative reactors for electricity generation in Brazil using the methodology developed under the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO), co-ordinated by the International Atomic Energy Agency (IAEA). INPRO was initiated in 2001 and has the main objective of helping to ensure that nuclear energy is available to contribute in a sustainable manner to the energy needs of the 21st century. Brazil joined the INPRO project since its beginning and in 2005 submitted a proposal for the assessment using INPRO methodology of two small-sized reactors (IRIS - International Reactor Innovative and Secure, and FBNR - Fixed Bed Nuclear Reactor) as potential components of an innovative nuclear energy system (INS) completed by a conventional open nuclear fuel cycle based on enriched uranium. The scope of this assessment study was restricted to the reactor component of the INS and to the methodology areas of economics and safety for IRIS, and proliferation resistance and safety for FBNR. The results indicate that both IRIS and FBNR innovative designs comply mostly with the basic principles of the areas assessed and have potential to comply with the remaining ones. (author)

  17. 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.

  18. Energy, electricity and nuclear power

    International Nuclear Information System (INIS)

    Reuss, P.; Naudet, G.

    2008-01-01

    After an introduction recalling what energy is, the first part of this book presents the present day energy production and consumption and details more particularly the electricity 'vector' which is an almost perfect form of energy despite the fact that it is not a primary energy source: it must be generated from another energy source and no large scale storage of this energy is possible. The second part of the book is devoted to nuclear energy principles and to the related technologies. Content: 1 - What does energy mean?: the occurrence of the energy concept, the classical notion of energy, energy notion in modern physics, energy transformations, energy conservation, irreversibility of energy transformations, data and units used in the energy domain; 2 - energy production and consumption: energy systems, energy counting, reserves and potentialities of energy resources, production of primary energies, transport and storage of primary energies, energy consumption, energy saving, energy markets and prices, energy indicators; 3 - electric power: specificity of electricity and the electric system, power networks, power generation, electricity storage, power consumption and demand, power generation economics, electricity prices and market; 4 - physical principles of nuclear energy: nuclei structure and binding energy, radioactivity and nuclear reactions, nuclear reactions used in energy generation, basics of fission reactors physics; 5 - nuclear techniques: historical overview, main reactor types used today, perspectives; 6 - fuel cycle: general considerations, uranium mining, conversion, enrichment, fuel fabrication, back-end of the cycle, plutonium recycle in water cooled reactors; 7 - health and environmental aspects of nuclear energy: effects on ionizing radiations, basics of radiation protection, environmental impacts of nuclear energy, the nuclear wastes problem, specific risks; 8 - conclusion; 9 - appendixes (units, physics constants etc..)

  19. Nuclear reactors to come

    International Nuclear Information System (INIS)

    Lung, M.

    2002-01-01

    The demand for nuclear energy will continue to grow at least till 2050 because of mainly 6 reasons: 1) the steady increase of the world population, 2) China, India and Indonesia will reach higher social standard and their energy consumption will consequently grow, 3) fossil energy resources are dwindling, 4) coal will be little by little banned because of its major contribution to the emission of green house effect gas, 5) renewable energies need important technological jumps to be really efficient and to take the lead, and 6) fusion energy is not yet ready to take over. All these reasons draw a promising future for nuclear energy. Today 450 nuclear reactors are operating throughout the world producing 17% of the total electrical power demand. In order to benefit fully of this future, nuclear industry has to improve some characteristics of reactors: 1) a more efficient use of uranium (it means higher burnups), 2) a simplification and automation of reprocessing-recycling chain of processes, 3) efficient measures against proliferation and against any misuse for terrorist purposes, and 4) an enhancement of safety for the next generation of reactors. The characteristics of fast reactors and of high-temperature reactors will likely make these kinds of reactors the best tools for energy production in the second half of this century. (A.C.)

  20. Pressure test at the reactor building of the Embalse Nuclear Power Plant (CNE)

    International Nuclear Information System (INIS)

    Coutsiers, E.E.; Perrino, J.; Moreno, C.; Batistic, J.A.; Lolis, R.R.; Aviles, A.

    1991-01-01

    Upon request by the Licensing Authority, the reactor building (RB) in a nuclear power plant must be submitted to pressure tests. One of these tests is to be performed before startup and, then, a test must be carried out every 5 years in operation. The pre-operational tests took place in August 1981, under two values of relative pressure: 1.266 kg/cm 2 and 0.422 kg/cm 2 . Operational tests must only be made at the lower pressure and their objective is to verify that the loss speed remains within the range indicated in the corresponding technical specification. The first operational test was performed in August 1989. The personnel of the CNE took care of the preparation of the Work Plan, of aligning the various systems contained in the RB, of pressurization, of monitoring localized tightedness, of depressurization and of the general and quality control of the test. The measurements were carried out by the CISME (Center of Metrology Research and Service) of the National Institute of Industrial Technology (INTI) , which did also supply the necesary instruments and the data collection system. There is also a description of the work performed before the test, of the calculation method used for assessing the loss rate, of the test sequencies and of the results obtained. (Author) [es

  1. Nuclear reactors

    International Nuclear Information System (INIS)

    Middleton, J.E.

    1977-01-01

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

  2. Experiments and models of general corrosion and flow-assisted corrosion of materials in nuclear reactor environments

    Science.gov (United States)

    Cook, William Gordon

    Corrosion and material degradation issues are of concern to all industries. However, the nuclear power industry must conform to more stringent construction, fabrication and operational guidelines due to the perceived additional risk of operating with radioactive components. Thus corrosion and material integrity are of considerable concern for the operators of nuclear power plants and the bodies that govern their operations. In order to keep corrosion low and maintain adequate material integrity, knowledge of the processes that govern the material's breakdown and failure in a given environment are essential. The work presented here details the current understanding of the general corrosion of stainless steel and carbon steel in nuclear reactor primary heat transport systems (PHTS) and examines the mechanisms and possible mitigation techniques for flow-assisted corrosion (FAC) in CANDU outlet feeder pipes. Mechanistic models have been developed based on first principles and a 'solution-pores' mechanism of metal corrosion. The models predict corrosion rates and material transport in the PHTS of a pressurized water reactor (PWR) and the influence of electrochemistry on the corrosion and flow-assisted corrosion of carbon steel in the CANDU outlet feeders. In-situ probes, based on an electrical resistance technique, were developed to measure the real-time corrosion rate of reactor materials in high-temperature water. The probes were used to evaluate the effects of coolant pH and flow on FAC of carbon steel as well as demonstrate of the use of titanium dioxide as a coolant additive to mitigated FAC in CANDU outlet feeder pipes.

  3. Reliability and testing considerations in the design of nuclear reactor filtration systems

    International Nuclear Information System (INIS)

    O'Nan, A.; Williams, R.P.; Goldsmith, J.M.

    1975-01-01

    The high performance standards set by USAEC-DRL Regulatory Guides for nuclear reactor filtration systems pose difficult problems for on-site leakage tests. These problems are compounded by the crowded conditions inside reactor structures, and by the fact that, until recently, little consideration has been given by system designers to the needs of testing. Techniques for coping with testing problems on existing systems, and suggestions for improving the testability of future systems, are given. Test crew safety considations are discussed, and a pair of easily portable contaminant generators is described. (U.S.)

  4. Nuclear reactor

    International Nuclear Information System (INIS)

    Hattori, Sadao; Sato, Morihiko.

    1994-01-01

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

  5. Review of light--water reactor safety studies. Volume 3 of health and safety impacts of nuclear, geothermal, and fossil-fuel electric generation in California

    International Nuclear Information System (INIS)

    Nero, A.V.; Farnaam, M.R.K.

    1977-01-01

    This report summarizes and compares important studies of light-water nuclear reactor safety, emphasizing the Nuclear Regulatory Commission's Reactor Safety Study, work on risk assessment funded by the Electric Power Research Institute, and the Report of the American Physical Society study group on light-water reactor safety. These reports treat risk assessment for nuclear power plants and provide an introduction to the basic issues in reactor safety and the needs of the reactor safety research program. Earlier studies are treated more briefly. The report includes comments on the Reactor Safety Study. The manner in which these studies may be used and alterations which would increase their utility are discussed

  6. An overview of the Nuclear Electric Propulsion Space Test Program (NEPSTP) satellite

    International Nuclear Information System (INIS)

    Voss, S.S.; Reynolds, E.L.

    1994-01-01

    Early in 1992 the idea of purchasing a Russian designed and fabricated space reactor power system and integrating it with a US designed satellite went from fiction to reality with the purchase of the first two Topaz II reactors by the Strategic Defense Initiative Organization (now the Ballistic Missile Defense Organization (BMDO)). The New Mexico Alliance was formed to establish a ground test facility in which to perform nonnuclear systems testing of the Topaz II, and to evaluate the Topaz II system for flight testing with respect to safety, performance, and operability. In conjunction, SDIO requested that the Applied Physics Laboratory in Laurel, MD propose a mission and design a satellite in which the Topaz II could be used as the power source. The outcome of these two activities was the design of the Nuclear Electric Propulsion Space Test Program (NEPSTP) satellite which combines a modified Russian Topaz II power system with a US designed satellite to achieve a specified mission. Due to funding reduction within the SDIO, the Topaz II flight program was postponed indefinitely at the end of Fiscal year 1993. The purpose of this paper is to present an overview of the NEPSTP mission and the satellite design at the time the flight program ended

  7. Evaluating Russian space nuclear reactor technology for United States applications

    International Nuclear Information System (INIS)

    Polansky, G.F.; Schmidt, G.L.; Voss, S.S.; Reynolds, E.L.

    1994-01-01

    Space nuclear power and nuclear electric propulsion are considered important technologies for planetary exploration, as well as selected earth orbit applications. The Nuclear Electric Propulsion Space Test Program (NEPSTP) was intended to provide an early flight demonstration of these technologies at relatively low cost through extensive use of existing Russian technology. The key element of Russian technology employed in the program was the Topaz II reactor. Refocusing of the activities of the Ballistic Missile Defense Organization (BMDO), combined with budgetary pressures, forced the cancellation of the NEPSTP at the end of the 1993 fiscal year. The NEPSTP was faced with many unique flight qualification issues. In general, the launch of a spacecraft employing a nuclear reactor power system complicates many spacecraft qualification activities. However, the NEPSTP activities were further complicated because the reactor power system was a Russian design. Therefore, this program considered not only the unique flight qualification issues associated with space nuclear power, but also with differences between Russian and United States flight qualification procedures. This paper presents an overview of the NEPSTP. The program goals, the proposed mission, the spacecraft, and the Topaz II space nuclear power system are described. The subject of flight qualification is examined and the inherent difficulties of qualifying a space reactor are described. The differences between United States and Russian flight qualification procedures are explored. A plan is then described that was developed to determine an appropriate flight qualification program for the Topaz II reactor to support a possible NEPSTP launch

  8. Simulated nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    Berta, V.T.

    1993-01-01

    An apparatus for electrically simulating a nuclear reactor fuel assembly. It includes a heater assembly having a top end and a bottom end and a plurality of concentric heater tubes having electrical circuitry connected to a power source, and radially spaced from each other. An outer target tube and an inner target tube is concentric with the heater tubes and with each other, and the outer target tube surrounds and is radially spaced from the heater tubes. The inner target tube is surrounded by and radially spaced from the heater tubes and outer target tube. The top of the assembly is generally open to allow for the electrical power connection to the heater tubes, and the bottom of the assembly includes means for completing the electrical circuitry in the heater tubes to provide electrical resistance heating to simulate the power profile in a nuclear reactor. The embedded conductor elements in each heater tube is split into two halves for a substantial portion of its length and provided with electrical isolation such that each half of the conductor is joined at one end and is not joined at the other end

  9. Attitudes of the general public and electric power company employees toward nuclear power generation

    International Nuclear Information System (INIS)

    Komiyama, Hisashi

    1997-01-01

    We conducted an awareness survey targeted at members of the general public residing in urban areas and in areas scheduled for construction of nuclear power plants as well as employees of electric power company in order to determine the awareness and attitude structures of people residing near scheduled construction sites of nuclear power plants with respect to nuclear power generation, and to examine ways of making improvements in terms of promoting nuclear power plant construction sites. Analysis of those results revealed that there are no significant differences in the awareness and attitudes of people residing in urban areas and in areas near scheduled construction sites. On the contrary, a general sense of apprehension regarding the construction of nuclear power plants was observed common to both groups. In addition, significant differences in awareness and attitudes with respect to various factors were determined to exist between members of the general public residing in urban areas and scheduled construction sites and employees of electric power company. (author)

  10. Standard Guide for Conducting Supplemental Surveillance Tests for Nuclear Power Reactor Vessels, E 706 (IH)

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This guide discusses test procedures that can be used in conjunction with, but not as alternatives to, those required by Practices E185 and E2215 for the surveillance of nuclear reactor vessels. The supplemental mechanical property tests outlined permit the acquisition of additional information on radiation-induced changes in fracture toughness, notch ductility, and yield strength properties of the reactor vessel steels. 1.2 This guide provides recommendations for the preparation of test specimens for irradiation, and identifies special precautions and requirements for reactor surveillance operations and postirradiation test planning. Guidance on data reduction and computational procedures is also given. Reference is made to other ASTM test methods for the physical conduct of specimen tests and for raw data acquisition.

  11. 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

  12. Advanced Demonstration and Test Reactor Options Study

    International Nuclear Information System (INIS)

    Petti, David Andrew; Hill, R.; Gehin, J.; Gougar, Hans David; Strydom, Gerhard; Heidet, F.; Kinsey, J.; Grandy, Christopher; Qualls, A.; Brown, Nicholas; Powers, J.; Hoffman, E.; Croson, D.

    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'. Advanced reactors are

  13. Nuclear reactor

    International Nuclear Information System (INIS)

    Garabedian, G.

    1988-01-01

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

  14. In-reactor testing of the closed cycle gas core reactor---the nuclear light bulb concept

    International Nuclear Information System (INIS)

    Gauntt, R.O.; Slutz, S.A.; Harms, G.A.; Latham, T.S.; Roman, W.C.; Rodgers, R.J.

    1993-01-01

    The Nuclear Light Bulb (NLB) concept is an advanced closed cycle space propulsion rocket engine design that offers unprecidented performance characteristics in terms of specific impulse (>1800 s) and thrust (>445 kN). The NLB is a gas-core nuclear reactor making use of thermal radiation from a high temperature U-plasma core to heat the hydrogen propellant to very high temperatures (∼4000 K). The following paper describes analyses performed in support of the design of in-reactor tests that are planned to be performed in the Annular Core Research Reactor (ACRR) at Sandia National Laboratories in order to demonstrate the technical feasibility of this advanced concept. The tests will examine the stability of a hydrodynamically confined fissioning U-plasma under steady and transient conditions. Testing will also involve study of propellant heating by thermal radiation from the plasma and materials performance in the nuclear environment of the NLB. The analyses presented here include neutronic performance studies and U-plasma radiation heat-transport studies of small vortex-confined fissioning U-plasma experiments that are irradiated in the ACRR. These analyses indicate that high U-plasma temperatures (4000 to 9000 K) can be sustained in the ACRR for periods of time on the order of 5 to 20 s. These testing conditions are well suited to examine the stability and performance requirements necessary to demonstrate the feasibility of this concept

  15. Non-electrical Application of Nuclear Energy: Some General Issues and Prospects

    International Nuclear Information System (INIS)

    Kuznetsov, Yu. N.

    2008-01-01

    effective NDC structure allowing extraction of turbine steam to supply heat (via an intermediate circuit) to a desalination system producing a high-quality distillate. A nuclear plant consisting of two VK-300 facilities can supply heat to a desalination complex with the output of 300.000 m 3 /day, which allows distillate production at a cost of 0.58 dollars/m 3 . In this case, 357 MWe can be used to produce electricity at a cost of 0.029 dollars/kWh. The cost estimates have demonstrated that nuclear desalination facilities with RUTA reactors would be competitive with fossil fuel plants of the same capacity in the regions relying on expensive imported fuel. Another technology that offers a wide range of non-electricity applications is high-temperature gas cooled reactor (HTGR). The paper discusses HTGR application for production of synthesis gas (mixture of CO and H 2 ) by way of black and slate coal gasification, in radiation and chemical processes (such as polymerization of polyethylene, organic concrete, etc.), and in hydrogen production (steam conversion of methyl hydride, high-temperature electrolysis of water, thermochemical cycle). This seems to be the most expedient way - technically and economically - of HTGR application.(author)

  16. Nuclear research reactors

    International Nuclear Information System (INIS)

    1985-01-01

    It's presented data about nuclear research reactors in the world, retrieved from the Sien (Nuclear and Energetic Information System) data bank. The information are organized in table forms as follows: research reactors by countries; research reactors by type; research reactors by fuel and research reactors by purpose. (E.G.) [pt

  17. Electric arc apparatus for severing split-pin assemblies of guide tubes of nuclear reactors

    International Nuclear Information System (INIS)

    Burns, D.C.; Kauric, C.E.; Persang, J.C.

    1987-01-01

    This patent describes an apparatus for use in the replacement of an old split-pin assembly of a guide tube of a nuclear reactor by a new split-pin assembly, the old split-pin assembly including an old split pin and an old nut securing the old split pin to the guide tube, the old split-pin assembly and the guide tube being radioactive. The apparatus includes a metal disintegration machining tool, the tool having an electrode, means for mounting the tool submerged in a pool of water in engagement with the guide tube and with the old split-pin assembly secured to the guide tube, the tool being so mounted with the electrode in position to coact electrically with the last-named old split-pin assembly but not with the guide tube, and means, connected to the tool, for firing a disintegrating arc between the electrode and the assembly to disintegrate the assembly into readily removable fragments

  18. Nuclear reactor

    International Nuclear Information System (INIS)

    Schulze, I.; Gutscher, E.

    1980-01-01

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

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

    International Nuclear Information System (INIS)

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

    1989-12-01

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

  20. 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.

  1. Reactor operator screening test experiences

    International Nuclear Information System (INIS)

    O'Brien, W.J.; Penkala, J.L.; Witzig, W.F.

    1976-01-01

    When it became apparent to Duquesne Light Company of Pittsburgh, Pennsylvania, that the throughput of their candidate selection-Phase I training-reactor operator certification sequence was something short of acceptable, the utility decided to ask consultants to make recommendations with respect to candidate selection procedures. The recommendation implemented was to create a Nuclear Training Test that would predict the success of a candidate in completing Phase I training and subsequently qualify for reactor operator certification. The mechanics involved in developing and calibrating the Nuclear Training Test are described. An arbitration decision that resulted when a number of International Brotherhood of Electrical Workers union employees filed a grievance alleging that the selection examination was unfair, invalid, not job related, inappropriate, and discriminatorily evaluated is also discussed. The arbitration decision favored the use of the Nuclear Training Test

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

    International Nuclear Information System (INIS)

    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 various core deformations. The power delivered to the SAFE-100 prototype was then adjusted accordingly via kinetics calculations. The work presented in this paper will demonstrate virtual reactivity feedback as core power was increased from 1 kWt to 10 kWt, held approximately constant at 10 kWt, and then allowed to decrease based on the negative thermal reactivity coefficient. (authors)

  3. Nuclear reactor

    International Nuclear Information System (INIS)

    Gilroy, J.E.

    1980-01-01

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

  4. Nuclear reactor

    International Nuclear Information System (INIS)

    Hattori, Sadao; Sekine, Katsuhisa.

    1987-01-01

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

  5. Testing of reactor fuel materials using nuclear techniques

    International Nuclear Information System (INIS)

    Khouri, M.T.F.C.

    1978-01-01

    The tests presented here apply to: the quantitative determination of uranium in the core of fuel element plates by the detection of the number of neutrons produced in photo induced reactions in uranium; the determination of 235 U proportion in uranium dioxide samples, in the form of uranyl nitrate, by the technique of the detection of tracks produced by fission fragments and in pellet samples by passive gamma spectrometry and the checking of uranium homogenization distribution in fuel plates and uranium dioxide pellets. (Author) [pt

  6. 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 ...

  7. Construction and tests of a gamma device for experimental measurements of burnup of nuclear reactor fuel

    International Nuclear Information System (INIS)

    Brandao Junior, F.A.

    1982-01-01

    The gamma-scanning method is an important tool for the measurement of burnup of nuclear reactor fuel. The adequate knowledge of burnup allows for a better inventory of 'sensitive' fissile materials, better fuel management and provides insight on fuel behaviour and safety margins. This paper is related to the description, construction and operation of a first gamma scanning device, tested by irradiation of prototype PWR fuel pins, 14 cm long, in a Triga Mark-I reactor at very low power. Despite the limitations imposed by the low burnup, the experiment permitted a good checking of the main physical concepts and devices involved in the method. (Author) [pt

  8. Acoustic emission monitoring of preservice testing at Watts Bar Unit 1 Nuclear Reactor

    International Nuclear Information System (INIS)

    Hutton, P.H.; Pappas, R.A.; Friesel, M.A.

    1985-02-01

    Acoustic emission (AE) monitoring of selected pressure boundary areas at TVA's Watts Bar, Unit 1 Nuclear Plant in the US during hot functional preservice testing is described. Background, methodology, and results are included. The work discussed here is a major milestone in a program supported by the US 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 to AE monitoring during reactor operation. 3 refs., 6 figs

  9. Fabrication and Testing of a Modular Micro-Pocket Fission Detector Instrumentation System for Test Nuclear Reactors

    Science.gov (United States)

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

    2018-01-01

    Advancements in nuclear reactor core modeling and computational capability have encouraged further development of in-core neutron sensors. Measurement of the neutron-flux distribution within the reactor core provides a more complete understanding of the operating conditions in the reactor than typical ex-core sensors. Micro-Pocket Fission Detectors have been developed and tested previously but have been limited to single-node operation and have utilized highly specialized designs. The development of a widely deployable, multi-node Micro-Pocket Fission Detector assembly will enhance nuclear research capabilities. A modular, four-node Micro-Pocket Fission Detector array was designed, fabricated, and tested at Kansas State University. The array was constructed from materials that do not significantly perturb the neutron flux in the reactor core. All four sensor nodes were equally spaced axially in the array to span the fuel-region of the reactor core. The array was filled with neon gas, serving as an ionization medium in the small cavities of the Micro-Pocket Fission Detectors. The modular design of the instrument facilitates the testing and deployment of numerous sensor arrays. The unified design drastically improved device ruggedness and simplified construction from previous designs. Five 8-mm penetrations in the upper grid plate of the Kansas State University TRIGA Mk. II research nuclear reactor were utilized to deploy the array between fuel elements in the core. The Micro-Pocket Fission Detector array was coupled to an electronic support system which has been specially developed to support pulse-mode operation. The Micro-Pocket Fission Detector array composed of four sensors was used to monitor local neutron flux at a constant reactor power of 100 kWth at different axial locations simultaneously. The array was positioned at five different radial locations within the core to emulate the deployment of multiple arrays and develop a 2-dimensional measurement of

  10. Nuclear reactors; graphical symbols

    International Nuclear Information System (INIS)

    1987-11-01

    This standard contains graphical symbols that reveal the type of nuclear reactor and is used to design graphical and technical presentations. Distinguishing features for nuclear reactors are laid down in graphical symbols. (orig.) [de

  11. Nuclear reactor

    International Nuclear Information System (INIS)

    Sakurai, Mikio; Yamauchi, Koki.

    1983-01-01

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

  12. 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.

  13. Nuclear reactor

    International Nuclear Information System (INIS)

    Jungmann, A.

    1975-01-01

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

  14. Nuclear reactor

    International Nuclear Information System (INIS)

    Shirakawa, Toshihisa.

    1979-01-01

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

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

    International Nuclear Information System (INIS)

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

    1982-03-01

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

  16. Installation modification of the reactor No.2 of Ikata nuclear power plant of Shikoku Electric Power Company, Inc

    International Nuclear Information System (INIS)

    1980-01-01

    The application was made on August 25, 1979, from the president of the Shikoku Electric Power Company, Inc., to the Minister of International Trade and Industry, relating to the installation modification of the reactor No. 2 in the Ikata nuclear power plant. The inquiry was submitted on September 28, 1979, from the Minister of International Trade and Industry to the Nuclear Safety Commission, after the safety evaluation in the Ministry of International Trade and Industry, and the investigation and deliberation were started on October 1, 1979, in the Nuclear Safety Commission. The content of the modification is to add the circuit actuated by the abnormal low pressure signal of the reactor to the actuating circuit of the emergency core cooling system (ECCS) and to increase the new fuel storage capacity from about 1/3 core to about 2/3 core. The additional signal circuit is composed of the logic circuit of ''2 out of 4'' and is multichannel design. The circuit is independent from the reactor control system and the conventional signal circuit of the concurrence of low pressure in the reactor and low level in the pressurizer. With the addition of the circuit of abnormal low pressure signal of the reactor, the countermeasures for preventing ECCS start by mistake are also added. These modifications give no influence to the functions of the reactor control system and reactor protection system. The function and the performance of ECCS were analyzed and evaluated accompanying these modifications assuming the loss of coolant accident. Concerning the new fuel storage capacity, the type of racks is modified from angle type to can type, and the subcriticality is kept even at the time of water flood. (Nakai, Y.)

  17. Nuclear reactor

    International Nuclear Information System (INIS)

    Jolly, R.

    1979-01-01

    The support grid for the fuel rods of a liquid metal cooled fast breeder reactor has a regular hexagonal contour and contains a large number of unit cells arranged honeycomb fashion. The totality of these cells make up a hexagonal shape. The grid contains a number of strips of material, and there is a window in each of three sidewalls staggered by one sidewall. The other sidewalls have embossed protrusions, thus generating a guide lining or guide bead. The windows reduce the rigidity of the areas in the middle between the ends of the cells. (DG) [de

  18. 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.

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

    International Nuclear Information System (INIS)

    1979-01-01

    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

  20. 9th international conference on high-temperature reactors - coal and nuclear energy for electricity and gas generation

    International Nuclear Information System (INIS)

    Kelber, G.

    1987-01-01

    The site of the high-temperatur reactor in the Ruhr region neighbouring on a coal-fired power plant is not accidental. The potential of the high-temperature reactor as a central plant element for the supply of heat for heating purposes and process heat covers also the possibility of coal gasification and liquefaction. Therefore the high-temperature reactor is, in the long term, a ray of hope for the coal region, able to compensate for the production-related competitive disadvantages of local coal. It can contribute to guaranteeing in the long term the task of German hard coal as an essential pillar of our energy supply. The VGB as a technical association of thermal power plant operators is particularly committed to the integration of coal and nuclear energy. Within the bounds of its possibilities, it will contribute to promoting the safe and environmentally beneficial generation of electricity from the two primary energy sources. (orig./DG) [de

  1. DESIGN SAFETY PROBLEMS OF NUCLEAR REACTORS IN SPACE FOR ELECTRICAL POWER

    Energy Technology Data Exchange (ETDEWEB)

    Pickler, D A

    1963-06-15

    A general treatment is presented of some of the problems in the design safety of reactors which are to be operated in space. The basic requirements of these reachigh temperatures. The usual concept of a space reactor is described briefly, and the hazards of an assumed unmanned vehicle with an enriched-U-fueled reactor are examined during its launching, orbit, and reentry. Graphs are given for the dose vs distance downwind for an excursion of 100 Mw-sec, for the activity vs time after shutdown of a reactor which has been operated for 5 yr at 100 kw(t), and for the altitude vs orbital lifetime. Apparent conflicts between the basic requirements are discussed. (D.L.C.)

  2. Potential criticality accident at the General Electric Nuclear Fuel and Component Manufacturing Facility, May 29, 1991

    International Nuclear Information System (INIS)

    1991-08-01

    At the General Electric Nuclear Fuel and Component Manufacturing facility, located near Wilmington, North Carolina, on May 28 and 29, 1991, approximately 150 kilograms of uranium were inadvertently transferred from safe process tanks to an unsafe tank located at the waste treatment facility, thus creating the potential for a localized criticality safety problem. The excess uranium was ultimately safely recovered when the tank contents were centrifuged to remove the uranium-bearing material. Subsequently, the US Nuclear Regulatory Commission dispatched an Incident Investigation Team to determine what happened, to identify probable causes, and to make appropriate findings and conclusions. This report describes the incident, the methodology used by the team in its investigation, and presents the team's findings and conclusions. 48 figs., 8 tabs

  3. Nuclear reactor

    International Nuclear Information System (INIS)

    Schabert, H.P.; Weber, R.; Bauer, A.

    1975-01-01

    The refuelling of a PWR power reactor of about 1,200 MWe is performed by a transport pipe in the containment leading from an external to an internal fuel pit. A wagon to transport the fuel elements can go from a vertical loading position to an also vertical deloading position in the inner fuel pit via guide rollers. The necessary horizontal movement is effected by means of a cable line through the transport pipe which is inclined at least 10 0 . Gravity thus helps in the movement to the deloading position. The cable line with winch is fastened outside the containment. Swivelling devices tip the wagon from the horizontal to the vertical position or vice versa. Loading and deloading are done laterally. (TK/LH) [de

  4. Hot cell works and related irradiation tests in fission reactor for development of new materials for nuclear application

    International Nuclear Information System (INIS)

    Shikama, Tatsuo

    1999-01-01

    Present status of research works in Oarai Branch, Institute for Materials Research, Tohoku University, utilizing Japan Materials Testing Reactor and related hot cells will be described.Topics are mainly related with nuclear materials studies, excluding fissile materials, which is mainly aiming for development of materials for advanced nuclear systems such as a nuclear fusion reactor. Conflict between traditional and routined procedures and new demands will be described and future perspective is discussed. (author)

  5. Numerical solutions of the aerosol general dynamic equation for nuclear reactor safety studies

    International Nuclear Information System (INIS)

    Park, J.W.

    1988-01-01

    Methods and approximations inherent in modeling of aerosol dynamics and evolution for nuclear reactor source term estimation have been investigated. Several aerosol evolution problems are considered to assess numerical methods of solving the aerosol dynamic equation. A new condensational growth model is constructed by generalizing Mason's formula to arbitrary particle sizes, and arbitrary accommodation of the condensing vapor and background gas at particle surface. Analytical solution is developed for the aerosol growth equation employing the new condensation model. The space-dependent aerosol dynamic equation is solved to assess implications of spatial homogenization of aerosol distributions. The results of our findings are as follows. The sectional method solving the aerosol dynamic equation is quite efficient in modeling of coagulation problems, but should be improved for simulation of strong condensation problems. The J-space transform method is accurate in modeling of condensation problems, but is very slow. For the situation considered, the new condensation model predicts slower aerosol growth than the corresponding isothermal model as well as Mason's model, the effect of partial accommodation is considerable on the particle evolution, and the effect of the energy accommodation coefficient is more pronounced than that of the mass accommodation coefficient. For the initial conditions considered, the space-dependent aerosol dynamics leads to results that are substantially different from those based on the spatially homogeneous aerosol dynamic equation

  6. Nuclear reactors

    International Nuclear Information System (INIS)

    Prescott, R.F.

    1976-01-01

    In the system described the fuel elements are arranged vertically in groups and are supported in such a manner as to tend to tilt them towards the center of the respective group, the fuel elements being urged laterally into abutment with one another. The elements have interlocking bearing pads, whereby lateral movement of adjacent elements is resisted; this improves the stability of the reactor core during refuelling operations. Fuel elements may comprise clusters of parallel fuel pins enclosed in a wrapper of hexagonal cross section, with bearing pads in the form of spline-like ribs located on each side of the wrapper and extending parallel to the longitudinal axis of the fuel element, being interlockable with ribs on pads of adjacent fuel elements. The arrangement is applicable to a reactor core in which fuel elements and control rod guide tubes are arranged in modules each of which comprises a cluster of at least three fuel elements, one of which is rigidly supported whilst the others are resiliently tilted towards the center of the cluster so as to lean on the rigidly supported element. It is also applicable to modules comprising a cluster of six fuel elements, each resiliently tilted towards a central void to form a circular arch. The modules may include additional fuel elements located outside the clusters and also resiliently tilted towards the central voids, the latter being used to accommodate control rod guide tubes. The need for separate structural members to act as leaning posts is thus avoided. Such structural members are liable to irradiation embrittlement, that could lead to core failure. (U.K.)

  7. Nuclear reactor (1960)

    International Nuclear Information System (INIS)

    Maillard, M.L.

    1960-01-01

    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) [fr

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

    International Nuclear Information System (INIS)

    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

  9. 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.

  10. New generation of nuclear reactors

    International Nuclear Information System (INIS)

    Chwaszczewski, S.

    2000-01-01

    The development trends of the construction of nuclear reactors has been performed on the background of worldwide electricity demand for now and predicted for future. The social acceptance, political and economical circumstances has been also taken into account. Seems to Electric Power Research Institute (US) and other national authorities the advanced light water reactors have the best features and chances for further development and commercial applications in future

  11. Shock and vibration tests of uranium mononitride fuel pellets for a space power nuclear reactor

    Science.gov (United States)

    Adams, D. W.

    1972-01-01

    Shock and vibration tests were conducted on cylindrically shaped, depleted, uranium mononitride (UN) fuel pellets. The structural capabilities of the pellets were determined under exposure to shock and vibration loading which a nuclear reactor may encounter during launching into space. Various combinations of diametral and axial clearances between the pellets and their enclosing structures were tested. The results of these tests indicate that for present fabrication of UN pellets, a diametral clearance of 0.254 millimeter and an axial clearance of 0.025 millimeter are tolerable when subjected to launch-induced loads.

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

    International Nuclear Information System (INIS)

    Semyonov, Y.P.; Bakanov, Y.A.; Synyavsky, V.V.; Yuditsky, V.D.

    1997-01-01

    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)

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

    International Nuclear Information System (INIS)

    Perrin, J.S.

    1978-01-01

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

  14. Type test of Class 1E electric cables, field splices, and connections for nuclear power generating stations - 1975

    International Nuclear Information System (INIS)

    Anon.

    1976-01-01

    This Standard provides direction for establishing type tests which may be used in qualifying Class 1E electric cables, field splices, and other connections for service in nuclear power generating stations. General guidelines for qualifications are given in IEEE Std 323-1974, Standard for Qualifying Class 1E Electric Equipment for Nuclear Power Generating Stations. Categories of cables covered are those used for power control and instrumentation services. Though intended primarily to pertain to cable for field installation, this guide may also be used for the qualification of internal wiring of manufactured devices

  15. In core instrumentation for online nuclear heating measurements of material testing reactor

    International Nuclear Information System (INIS)

    Reynard, C.; Andre, J.; Brun, J.; Carette, M.; Janulyte, A.; Merroun, O.; Zerega, Y.; Lyoussi, A.; Bignan, G.; Chauvin, J-P.; Fourmentel, D.; Glayse, W.; Gonnier, C.; Guimbal, P.; Iracane, D.; Villard, J.-F.

    2010-01-01

    The present work focuses on nuclear heating. This work belongs to a new advanced research program called IN-CORE which means 'Instrumentation for Nuclear radiations and Calorimetry Online in REactor' between the LCP (University of Provence-CNRS) and the CEA (French Atomic Energy Commission) - Jules Horowitz Reactor (JHR) program. This program started in September 2009 and is dedicated to the conception and the design of an innovative mobile experimental device coupling several sensors and ray detectors for on line measurements of relevant physical parameters (photonic heating, neutronic flux ...) and for an accurate parametric mapping of experimental channels in the JHR Core. The work presented below is the first step of this program and concerns a brief state of the art related to measurement methods of nuclear heating phenomena in research reactor in general and MTR in particular. A special care is given to gamma heating measurements. A first part deals with numerical codes and models. The second one presents instrumentation divided into various kinds of sensor such as calorimeter measurements and gamma ionization chamber measurements. Their basic principles, characteristics such as metrological parameters, operating mode, disadvantages/advantages, ... are discussed. (author)

  16. The Next Generation Nuclear Plant/Advanced Gas Reactor Fuel Irradiation Experiments in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    Grover, S. Blaine

    2009-01-01

    The United States Department of Energy's Next Generation Nuclear Plant (NGNP) Program will be irradiating eight separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) 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 irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006, and the second experiment (AGR-2) is currently in the design phase. The design of test trains, as well as the support systems and fission product monitoring system that will monitor and control the experiment during irradiation will be discussed. In

  17. Nuclear reactor neutron shielding

    Science.gov (United States)

    Speaker, Daniel P; Neeley, Gary W; Inman, James B

    2017-09-12

    A nuclear reactor includes a reactor pressure vessel and a nuclear reactor core comprising fissile material disposed in a lower portion of the reactor pressure vessel. The lower portion of the reactor pressure vessel is disposed in a reactor cavity. An annular neutron stop is located at an elevation above the uppermost elevation of the nuclear reactor core. The annular neutron stop comprises neutron absorbing material filling an annular gap between the reactor pressure vessel and the wall of the reactor cavity. The annular neutron stop may comprise an outer neutron stop ring attached to the wall of the reactor cavity, and an inner neutron stop ring attached to the reactor pressure vessel. An excore instrument guide tube penetrates through the annular neutron stop, and a neutron plug comprising neutron absorbing material is disposed in the tube at the penetration through the neutron stop.

  18. Nuclear reactor coolant channels

    International Nuclear Information System (INIS)

    Macbeth, R.V.

    1978-01-01

    A nuclear reactor coolant channel is described that is suitable for sub-cooled reactors as in pressurised water reactors as well as for bulk boiling, as in boiling water reactors and steam generating nuclear reactors. The arrangement aims to improve heat transfer between the fuel elements and the coolant. Full constructional details are given. See also other similar patents by the author. (U.K.)

  19. A follow-up test of failed fuel element of a nuclear reactor

    International Nuclear Information System (INIS)

    Peerasathien, W.

    1974-01-01

    This thesis is a result of test of a number of nuclear fuel rods which have not been used for a long time due to leakage of radioactivity. Water was circulated through each fuel rod in a test cylinder and radioactivity in water was measured. It was found that the detection of Cesium-137 which has a long half-life, does not indicate the extent of leakage of short-lived radioisotopes, some of which are gaseous. These gases are harmful to the reactor operators and users. A better result was obtained by placing the failed fuel rod in the test cylinder close to the reactor to induce fission. Short half-life gases or other nuclides of the same series were then directly measured

  20. Nuclear reactor

    International Nuclear Information System (INIS)

    Irion, L.; Tautz, J.; Ulrych, G.

    1976-01-01

    This additional patent complements the arrangement of non-return valves to prevent loss of cooling water on fracture of external tubes in the main coolant circuit (according to PS 24 24 427.7) by ensuring that the easily movable valves only operate in case of a fault, but do not flutter in operation, because the direction of flow is not the same at each location where they are installed. The remedy for this undesirable effect consists of allocating 1 non-return valve unit with 5 to 10 valves to each (of several) ducts for the cooling water intake. These units are installed in the annular space between the reactor vessel and the pressure vessel below the inlet of the ducts. Due to flow guidance surfaces in the same space, the incoming cooling water is deflected downwards and as the guiding surfaces are closed at the sides, must pass parallel to the valves of the non-return valve unit. On fracture of the external cooling water inlet pipe concerned, all valves of this unit close due to reversal of flow on the outlet side. (TK) [de

  1. Nuclear rocket engine reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lanin, Anatoly

    2013-07-01

    Covers a new technology of nuclear reactors and the related materials aspects. Integrates physics, materials science and engineering Serves as a basic book for nuclear engineers and nuclear physicists. 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.

  2. Advancing nuclear technology and research. The advanced test reactor national scientific user facility

    Energy Technology Data Exchange (ETDEWEB)

    Benson, Jeff B; Marshall, Frances M [Idaho National Laboratory, Idaho Falls, ID (United States); Allen, Todd R [Univ. of Wisconsin, Madison, WI (United States)

    2012-03-15

    The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is one of the world's premier test reactors for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material radiation 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. The mission of the ATR NSUF is to provide access to world-class facilities, thereby facilitating the advancement of nuclear science and technology. Cost free access to the ATR, INL post irradiation examination facilities, and partner facilities is granted based on technical merit to U.S. university-led experiment teams conducting non-proprietary research. Proposals are selected via independent technical peer review and relevance to United States Department of Energy. To increase overall research capability, ATR NSUF seeks to form strategic partnerships with university facilities that add significant nuclear research capability to the ATR NSUF and are accessible to all ATR NSUF users. (author)

  3. Electrical properties of InP irradiated by fast neutrons of a nuclear reactor

    International Nuclear Information System (INIS)

    Kolin, N.G.; Merkurisov, D.I.; Solov'ev, S.P.

    2000-01-01

    Electrophysical properties of InP single crystalline samples with different initial concentration of charge carriers have been studied in relation to irradiation conditions with fast neutrons of a nuclear reactor and subsequent heat treatments within the temperature range of 20-900 deg C. It has been shown that changes of the properties depend on the initial doping level. The annealing in the temperature range mentioned above results in the elimination of radiation defects. This makes possible to use the nuclear doping method for InP samples. In this respect the contribution of intermediate neutron reactions to the whole effect of the InP nuclear doping is estimated to be ∼ 10% [ru

  4. Small nuclear reactors for desalination

    International Nuclear Information System (INIS)

    Goldsmith, K.

    1978-01-01

    Small nuclear reactors are considered to have an output of not more than 400MW thermal. Since they can produce steam at much higher conditions than needed by the brine heater of a multi-flash desalination unit, it may be economically advantageous to use small reactors for a dual-purpose installation of appropriate size, producing both electricity and desalted water, rather than for a single-purpose desalination plant only. Different combinations of dual-purpose arrangements are possible depending principally on the ratio of electricity to water output required. The costs of the installation as well as of the products are critically dependent on this ratio. For minimum investment costs, the components of the dual-purpose installation should be of a standardised design based on normal commercial power plant practice. This then imposes some restrictions on the plant arrangement but, on the other hand, it facilitates selection of the components. Depending on the electricity to water ratio to be achieved, the conventional part of the installation - essentially the turbines - will form a combination of back-pressure and condensing machines. Each ratio will probably lead to an optimum combination. In the economic evaluation of this arrangement, a distinction must be made between single-purpose and dual-purpose installations. The relationship between output and unit costs of electricity and water will be different for the two cases, but the relation can be expressed in general terms to provide guidelines for selecting the best dimensions for the plant. (author)

  5. Training and qualification of licensed reactor operators at General Public Utilities Nuclear Corporation

    International Nuclear Information System (INIS)

    Long, R.L.; Coe, R.P.

    1992-01-01

    Following the Three Mile Island-2 (TMI-2) accident in 1979, the utility responsible for managing the facility has looked closely at the training and qualification of its reactor operators. Performance-based operator training programmes are now in place, as required by the United States National Academy for Nuclear Training. Operators also participate directly in the development of a professional code of behaviour. (UK)

  6. Modification of reactor installation in the Genkai nuclear power plant No. 1 of Kyushu Electric Power Company, Inc

    International Nuclear Information System (INIS)

    1979-01-01

    The Nuclear Safety Commission recognized the adequacy concerning the inquiry which was offered from the Minister of International Trade and Industry on July 25, 1979, following the safety evaluation in the Ministry of International Trade and Industry, and decided to submit a report to the Minister of International Trade and Industry on July 26, 1979, about the modification of reactor installation in the Genkai nuclear power plant No. 1 of the Kyushu Electric Power Company, Inc. This is concerned to the application which was made from the president of the Kyushu Electric Power Company, Inc., to the Minister of International Trade and Industry on July 24, 1979. The content of the modification is to add a control circuit which is actuated by the signal of abnormal low pressure in a reactor to the circuit of actuating the emergency core cooling system of the plant. The influences on the safety protection system by the addition of the circuit transmitting safety injection signal and by the additions of an interlock circuit preventing the misoperation of pressurizer spray and of a block circuit of safety injection signal in case of the abnormal low pressure in a reactor were evaluated. The effects on the function and characteristics of the emergency core cooling system due to the addition of the control circuit were investigated, and it was recognized by the analysis that there is no effect in the pipe ruptures of both small and large scales. (Nakai, Y.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Carlson, Richard Barney [Idaho National Lab. (INL), Idaho Falls, ID (United States); Scoffield, Don [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bennett, Brion [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    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.

  8. Physics of nuclear reactors

    International Nuclear Information System (INIS)

    Baeten, Peter

    2006-01-01

    This course gives an introduction to Nuclear Reactor Physics. The first chapter explains the most important parameters and concepts in nuclear reactor physics such as fission, cross sections and the effective multiplication factor. Further on, in the second chapter, the flux distributions in a stationary reactor are derived from the diffusion equation. Reactor kinetics, reactor control and reactor dynamics (feedback effects) are described in the following three chapters. The course concludes with a short description of the different types of existing and future reactors. (author)

  9. Thermionic reactors for space nuclear power

    Science.gov (United States)

    Homeyer, W. G.; Merrill, M. H.; Holland, J. W.; Fisher, C. R.; Allen, D. T.

    1985-01-01

    Thermionic reactor designs for a variety of space power applications spanning the range from 5 kWe to 3 MWe are described. In all of these reactors, nuclear heat is converted directly to electrical energy in thermionic fuel elements (TFEs). A circulating reactor coolant carries heat from the core of TFEs directly to a heat rejection radiator system. The recent design of a thermionic reactor to meet the SP-100 requirements is emphasized. Design studies of reactors at other power levels show that the same TFE can be used over a broad range in power, and that design modifications can extend the range to many megawatts. The design of the SP-100 TFE is similar to that of TFEs operated successfully in test reactors, but with design improvements to extend the operating lifetime to seven years.

  10. The modular pebble bed nuclear reactor - the preferred new sustainable energy source for electricity, hydrogen and potable water production?

    International Nuclear Information System (INIS)

    Kemeny, L.G.

    2003-01-01

    This paper describes a joint project of Massachusetts Institute of technology, Nu-Tec Inc. and Proto Power. The elegant simplicity of graphite moderated pebble bed reactor is the basis for the 'generation four' nuclear power plants. High Temperature Gas Cooled (HTGC) nuclear power plant have the potential to become the preferred base load sustainable energy source for the new millennium. The great attraction of these helium cooled 'Generation Four' nuclear plant can be summarised as follows: Factory assembly line production; Modularity and ease of delivery to site; High temperature Brayton Cycle ideally suited for cogeneration of electricity, potable water and hydrogen; Capital and operating costs competitive with hydrocarbon plant; Design is inherently meltdown proof and proliferation resistant

  11. The advanced test reactor national scientific user facility advancing nuclear technology

    International Nuclear Information System (INIS)

    Allen, T.R.; Thelen, M.C.; Meyer, M.K.; Marshall, F.M.; Foster, J.; Benson, J.B.

    2009-01-01

    To help ensure the long-term viability of nuclear energy through a robust and sustained research and development effort, the U.S. Department of Energy (DOE) designated the Advanced Test Reactor and associated post-irradiation examination facilities a National Scientific User Facility (ATR NSUF), allowing broader access to nuclear energy researchers. The mission of the ATR NSUF is to provide access to world-class nuclear research facilities, thereby facilitating the advancement of nuclear science and technology. The ATR NSUF seeks to create an engaged academic and industrial user community that routinely conducts reactor-based research. Cost free access to the ATR and PIE facilities is granted based on technical merit to U.S. university-led experiment teams conducting non-proprietary research. Proposals are selected via independent technical peer review and relevance to DOE mission. Extensive publication of research results is expected as a condition for access. During FY 2008, the first full year of ATR NSUF operation, five university-led experiments were awarded access to the ATR and associated post-irradiation examination facilities. The ATR NSUF has awarded four new experiments in early FY 2009, and anticipates awarding additional experiments in the fall of 2009 as the results of the second 2009 proposal call. As the ATR NSUF program mature over the next two years, the capability to perform irradiation research of increasing complexity will become available. These capabilities include instrumented irradiation experiments and post-irradiation examinations on materials previously irradiated in U.S. reactor material test programs. The ATR critical facility will also be made available to researchers. An important component of the ATR NSUF an education program focused on the reactor-based tools available for resolving nuclear science and technology issues. The ATR NSUF provides education programs including a summer short course, internships, faculty-student team

  12. The Advanced Test Reactor National Scientific User Facility Advancing Nuclear Technology

    International Nuclear Information System (INIS)

    Allen, T.R.; Benson, J.B.; Foster, J.A.; Marshall, F.M.; Meyer, M.K.; Thelen, M.C.

    2009-01-01

    To help ensure the long-term viability of nuclear energy through a robust and sustained research and development effort, the U.S. Department of Energy (DOE) designated the Advanced Test Reactor and associated post-irradiation examination facilities a National Scientific User Facility (ATR NSUF), allowing broader access to nuclear energy researchers. The mission of the ATR NSUF is to provide access to world-class nuclear research facilities, thereby facilitating the advancement of nuclear science and technology. The ATR NSUF seeks to create an engaged academic and industrial user community that routinely conducts reactor-based research. Cost free access to the ATR and PIE facilities is granted based on technical merit to U.S. university-led experiment teams conducting non-proprietary research. Proposals are selected via independent technical peer review and relevance to DOE mission. Extensive publication of research results is expected as a condition for access. During FY 2008, the first full year of ATR NSUF operation, five university-led experiments were awarded access to the ATR and associated post-irradiation examination facilities. The ATR NSUF has awarded four new experiments in early FY 2009, and anticipates awarding additional experiments in the fall of 2009 as the results of the second 2009 proposal call. As the ATR NSUF program mature over the next two years, the capability to perform irradiation research of increasing complexity will become available. These capabilities include instrumented irradiation experiments and post-irradiation examinations on materials previously irradiated in U.S. reactor material test programs. The ATR critical facility will also be made available to researchers. An important component of the ATR NSUF an education program focused on the reactor-based tools available for resolving nuclear science and technology issues. The ATR NSUF provides education programs including a summer short course, internships, faculty-student team

  13. 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.

  14. Decommissioning a nuclear reactor

    International Nuclear Information System (INIS)

    Montoya, G.M.

    1991-01-01

    The process of decommissioning a facility such as a nuclear reactor or reprocessing plant presents many waste management options and concerns. Waste minimization is a primary consideration, along with protecting a personnel and the environment. Waste management is complicated in that both radioactive and chemical hazardous wastes must be dealt with. This paper presents the general decommissioning approach of a recent project at Los Alamos. Included are the following technical objectives: site characterization work that provided a thorough physical, chemical, and radiological assessment of the contamination at the site; demonstration of the safe and cost-effective dismantlement of a highly contaminated and activated nuclear-fuelded reactor; and techniques used in minimizing radioactive and hazardous waste. 12 figs

  15. Control for nuclear reactor

    International Nuclear Information System (INIS)

    Ash, E.B.; Bernath, L.; Facha, J.V.

    1980-01-01

    A nuclear reactor is provided with several hydraulically-supported spherical bodies having a high neutron absorption cross section, which fall by gravity into the core region of the reactor when the flow of supporting fluid is shut off. (auth)

  16. Laser-Ultrasonic Testing and its Applications to Nuclear Reactor Internals

    Science.gov (United States)

    Ochiai, M.; Miura, T.; Yamamoto, S.

    2008-02-01

    A new nondestructive testing technique for surface-breaking microcracks in nuclear reactor components based on laser-ultrasonics is developed. Surface acoustic wave generated by Q-switched Nd:YAG laser and detected by frequency-stabilized long pulse laser coupled with confocal Fabry-Perot interferometer is used to detect and size the cracks. A frequency-domain signal processing is developed to realize accurate sizing capability. The laser-ultrasonic testing allows the detection of surface-breaking microcrack having a depth of less than 0.1 mm, and the measurement of their depth with an accuracy of 0.2 mm when the depth exceeds 0.5 mm including stress corrosion cracking. The laser-ultrasonic testing system combined with laser peening system, which is another laser-based maintenance technology to improve surface stress, for inner surface of small diameter tube is developed. The generation laser in the laser-ultrasonic testing system can be identical to the laser source of the laser peening. As an example operation of the system, the system firstly works as the laser-ultrasonic testing mode and tests the inner surface of the tube. If no cracks are detected, the system then changes its work mode to the laser peening and improves surface stress to prevent crack initiation. The first nuclear industrial application of the laser-ultrasonic testing system combined with the laser peening was completed in Japanese nuclear power plant in December 2004.

  17. Severe accident testing of electrical penetration assemblies

    International Nuclear Information System (INIS)

    Clauss, D.B.

    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

  18. Determination of a test section parameters for Iris nuclear reactor pressurizer

    International Nuclear Information System (INIS)

    Silva, Mario A.B. da; Lira, Carlos A.B. de O.

    2009-01-01

    An integral, modular and medium size nuclear reactor, known as IRIS, is being developed by Westinghouse and by research centers. IRIS is characterized by having most of its components inside the pressure vessel, eliminating the probability of accidents. Due to its integral configuration, there is no spray system for boron homogenization, which may cause power transients. Thus, boron mixing must be investigated. The aim of this paper is to establish the conditions under which a test section has to be built for boron dispersion analysis inside IRIS reactor pressurizer. Through Fractional Scaling Analysis, which is a new methodology of similarity, the main parameters for a test section are obtained. By combining Fractional Scaling Analysis with local scaling for the densimetric Froude number and a previously established volumetric scale factor, the values of recirculation orifices, inlet water temperature, time scale factor and recirculation flow for the test section (model) are determined so that boron distribution is well represented in IRIS reactor pressurizer (prototype). Analytical solutions were used to validate the adopted methodology and when the results simulated in the model are compared to those that characterize the prototype, the agreement for both systems is absolute. The thermal power also influences boron distribution inside the test section. This power is determined by condensation laws in the vapor region and by suitable correlations for free convection. The fractions for rising inlet recirculation water enthalpy and vapor formation are also considered. (author)

  19. 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.

  20. DEVELOPMENT AND ADAPTATION OF VORTEX REALIZABLE MEASUREMENT SYSTEM FOR BENCHMARK TEST WITH LARGE SCALE MODEL OF NUCLEAR REACTOR

    Directory of Open Access Journals (Sweden)

    S. M. Dmitriev

    2017-01-01

    Full Text Available The last decades development of applied calculation methods of nuclear reactor thermal and hydraulic processes are marked by the rapid growth of the High Performance Computing (HPC, which contribute to the active introduction of Computational Fluid Dynamics (CFD. The use of such programs to justify technical and economic parameters and especially the safety of nuclear reactors requires comprehensive verification of mathematical models and CFD programs. The aim of the work was the development and adaptation of a measuring system having the characteristics necessary for its application in the verification test (experimental facility. It’s main objective is to study the processes of coolant flow mixing with different physical properties (for example, the concentration of dissolved impurities inside a large-scale reactor model. The basic method used for registration of the spatial concentration field in the mixing area is the method of spatial conductometry. In the course of the work, a measurement complex, including spatial conductometric sensors, a system of secondary converters and software, was created. Methods of calibration and normalization of measurement results are developed. Averaged concentration fields, nonstationary realizations of the measured local conductivity were obtained during the first experimental series, spectral and statistical analysis of the realizations were carried out.The acquired data are compared with pretest CFD-calculations performed in the ANSYS CFX program. A joint analysis of the obtained results made it possible to identify the main regularities of the process under study, and to demonstrate the capabilities of the designed measuring system to receive the experimental data of the «CFD-quality» required for verification.The carried out adaptation of spatial sensors allows to conduct a more extensive program of experimental tests, on the basis of which a databank and necessary generalizations will be created

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

    International Nuclear Information System (INIS)

    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

  2. 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.

  3. The nuclear heating calculation scheme for material testing in the future Jules Horowitz Reactor

    International Nuclear Information System (INIS)

    Huot, N.; Aggery, A.; Blanchet, D.; Courcelle, A.; Czernecki, S.; Di-Salvo, J.; Doederlein, C.; Serviere, H.; Willermoz, G.

    2004-01-01

    An innovative nuclear heating calculation scheme for materials testing carried out in in the future Jules Horowitz reactor (JHR) is described. A heterogeneous gamma source calculation is first performed at assembly level using the deterministic code APOLLO2. This is followed by a Monte Carlo gamma transport calculation in the whole core using the TRIPOLI4 code. The calculated gamma sources at the assembly level are applied in the whole core simulation using a weighting based on power distribution obtained from the neutronic core calculation. (authors)

  4. TU electric reactor model verification

    International Nuclear Information System (INIS)

    Willingham, C.E.; Killgore, M.R.

    1989-01-01

    Power reactor benchmark calculations using the code package CASMO-3/SIMULATE-3 have been performed for six cycles of Prairie Island Unit 1. The reload fuel designs for the selected cycles include gadolinia as a burnable absorber, natural uranium axial blankets, and increased water-to-fuel ratio. The calculated results for both low-power physics tests (boron end points, control rod worths, and isothermal temperature coefficients) and full-power operation (power distributions and boron letdown) are compared to measured plant data. These comparisons show that the TU Electric reactor physics models accurately predict important physics parameters for power reactors

  5. GE's advanced nuclear reactor designs

    International Nuclear Information System (INIS)

    Berglund, R.C.

    1993-01-01

    The excess of US electrical generating capacity which has existed for the past 15 years is coming to an end as we enter the 1990s. Environmental and energy security issues associated with fossil fuels are kindling renewed interest in the nuclear option. The importance of these issues are underscored by the National Energy Strategy (NES) which calls for actions which open-quotes are designed to ensure that the nuclear power option is available to utilities.close quotes Utilities, utility associations, and nuclear suppliers, under the leadership of the Nuclear Power Oversight Committee (NPOC), have jointly developed a 14-point strategic plan aimed at establishing a predictable regulatory environment, standardized and pre-licensed Advanced Light Water Reactor (ALWR) nuclear plants, resolving the long-term waste management issue, and other open-quotes enabling conditions.close quotes GE is participating in this national effort and GE's family of advanced nuclear power plants feature two reactor designs, developed on a common technology base, aimed at providing a new generation of nuclear plants to provide safe, clean, economical electricity to the world's utilities in the 1990s and beyond. Together, the large-size (1300 MWe) Advanced Boiling Water Reactor (ABWR) and the small-size (600 MWe) Simplified Boiling Water Reactor (SBWR) are innovative, near-term candidates for expanding electrical generating capacity in the US and worldwide. Both possess the features necessary to do so safety, reliably, and economically

  6. Change in plan for installation of nuclear reactor in Genkai Nuclear Power Plant of Kyushu Electric Power Co., Inc. (change in plan for No.3 and No.4 nuclear reactor facilities) (report)

    International Nuclear Information System (INIS)

    1987-01-01

    This report, compiled by the Nuclear Safety Commission to be submitted to the Minister of International Trade and Industry, deals with studies on a proposed change in the plan for the installation of nuclear reactors in the Genkai Nuclear Power Plant of Kyushu Electric Power Co., Inc. (change in the plan for the No.3 and No.4 nuclear reactor facilities). The conclusions of and principles for the examination and evaluation are described first. The studies carried out are focused on the safety of the facilities, and it is concluded that part of the proposed change is appropriate with respect to the required technical capability and that part of the change will not have adverse effects on the safety design of the facilities. The examination of the safety design of the reactor facilities cover the reactivity control, new material for the steam generator, design of chemical and volume control systems, design of liquid waste treatment facilities, integration of all confinement vessel spray rings, and design of the diesel power generator. It is confirmed that all of them can meet the safety requirements. Studies and analyses are also made of the emission of radiations to the surrounding environment, abnormal transient changes during operations, and possible accidents. (Nogami, K.)

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

    International Nuclear Information System (INIS)

    1975-05-01

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

  8. Comparison of nuclear irradiation parameters of fusion breeder materials in high flux fission test reactors and a fusion power demonstration reactor

    International Nuclear Information System (INIS)

    Fischer, U.; Herring, S.; Hogenbirk, A.; Leichtle, D.; Nagao, Y.; Pijlgroms, B.J.; Ying, A.

    2000-01-01

    Nuclear irradiation parameters relevant to displacement damage and burn-up of the breeder materials Li 2 O, Li 4 SiO 4 and Li 2 TiO 3 have been evaluated and compared for a fusion power demonstration reactor and the high flux fission test reactor (HFR), Petten, the advanced test reactor (ATR, INEL) and the Japanese material test reactor (JMTR, JAERI). Based on detailed nuclear reactor calculations with the MCNP Monte Carlo code and binary collision approximation (BCA) computer simulations of the displacement damage in the polyatomic lattices with MARLOWE, it has been investigated how well the considered HFRs can meet the requirements for a fusion power reactor relevant irradiation. It is shown that a breeder material irradiation in these fission test reactors is well suited in this regard when the neutron spectrum is well tailored and the 6 Li-enrichment is properly chosen. Requirements for the relevant nuclear irradiation parameters such as the displacement damage accumulation, the lithium burn-up and the damage production function W(T) can be met when taking into account these prerequisites. Irradiation times in the order of 2-3 full power years are necessary for the HFR to achieve the peak values of the considered fusion power Demo reactor blanket with regard to the burn-up and, at the same time, the dpa accumulation

  9. Surveillance tests for light-water cooled nuclear power reactor vessels in IMEF

    International Nuclear Information System (INIS)

    Choo, Yong-Sun; Ahn, Sang-Bok; Park, Dae-Gyu; Jung, Yang-Hong; Yoo, Byung-Ok; Oh, Wan-Ho; Baik, Seung-Je; Koo, Dae-Seo; Lee, Key-Soon

    1999-01-01

    The surveillance tests for light-water cooled nuclear power reactor vessels were established to monitor the radiation-induced changes in the mechanical properties of ferritic materials in the beltline according to US NRC 10 CFR 50 App. G, US NRC RG1.99-rev.2, ASTM E185-82 and E185-94 in Irradiated Materials Examination Facility(IMEF). The surveillance capsule was transported from NPPs pool sites to KAERI IMEF by using a shipping cask. The capsule was cut and dismantled by capsule cutting machine and milling machine in M2 hot cell. Charpy tests and tension tests were performed in M5a and M5b hot cells respectively. Especially the EPMA located at hot lab was used to analyze the Ni and Cu wt% composition of base metal and weld for predicting the adjusted reference temperature(ART). The established process and test results were summarized in this paper. (author)

  10. Forced vibration tests on the reactor building of a nuclear power station, 1

    International Nuclear Information System (INIS)

    Takeda, Toshikazu; Tsunoda, Tomohiko; Wakamatsu, Kunio; Kaneko, Masataka; Nakamura, Mitsuru; Kunoh, Toshio; Murahashi, Hisahiro

    1988-01-01

    Tsuruga Unit No.2 Nuclear Power Station of the Japan Atomic Power Company is the first PWR-type 4-loop plant constructed in Japan with a prestressed concrete containment vessel (PCCV). This report describes forced vibration tests carried out on the reactor building of this plant. The following were obtained as results: (1) The results of the forced vibration tests corresponded well on the whole with design values. (2) The vibration characteristics of the PCCV observed in the tests after prestressing are no different from the ones before prestressing. This shows that the vibration properties of the PCCV are practically independent of prestressing loads. (3) A seismic response analysis of the design basis earthquake was made on the design model reflecting the test results. The seismic safety of the plant was confirmed by this analysis. (author)

  11. South Africa's nuclear model: A small and innovative reactor is seen as the model for new electricity plants. The project is nearing the starting blocks

    International Nuclear Information System (INIS)

    Ferreira, Tom

    2004-01-01

    Although nuclear power generation has by far the best safety and environmental record of any technology in general use, it has for many years been unable to make any meaningful inroads into the wall of negative perceptions that have arisen against it. But sentiments are changing rapidly on a global scale. The flare-up of oil prices is a sobering reminder of the volatility in the energy market, the exhaustibility of fossil fuels and the urgent need for stable, reliable, non-polluting sources of electrical power that are indispensable to a modern industrial economy. Today, new types of nuclear plants are prized, and South Africa is moving ahead. The State energy provider, Eskom, is internationally regarded as the leader in the field of the Pebble Bed Modular Reactor (PBMR) technology, a 'new generation' nuclear power plant. A decision on the PBMR project's future is on the near horizon. Should approvals be received in the coming months to proceed to the project's next phase, construction of the PBMR demonstration plant will start in 2006, in which case the reactor will start in 2010 and handed over to the client, Eskom, in 2011. Eskom has conditionally undertaken to purchase the first commercial units. Pebble bed reactors are small, about one-sixth the size of most current nuclear plants. Multiple PBMRs can share a common control center and occupy an area of no more than three football fields. More specifically, the PBMR is a helium-cooled, graphite moderated high temperature reactor (HTR). The concept is based on experience in the UK, United States and particularly Germany where prototype reactors were operated successfully between the late 1960s and 1980s. Although it is not the only high-temperature, gas-cooled nuclear reactor being developed in the world, the South African project is internationally regarded as a front-runner. The South African PBMR includes unique and patented technological innovations which make it particularly competitive. The Chief Executive

  12. Safety device for nuclear reactor

    International Nuclear Information System (INIS)

    Jacquelin, Roland.

    1977-01-01

    This invention relates to a safety device for a nuclear reactor, particularly a liquid metal (generally sodium) cooled fast reactor. This safety device includes an absorbing element with a support head connected by a disconnectable connector formed by the armature of an electromagnet at the end of an axially mobile vertical control rod. This connection is so designed that in the event of it becoming disconnected, the absorbing element gravity slides in a passage through the reactor core into an open container [fr

  13. Nuclear reactor theory

    International Nuclear Information System (INIS)

    Sekimoto, Hiroshi

    2007-09-01

    This textbook is composed of two parts. Part 1 'Elements of Nuclear Reactor Theory' is composed of only elements but the main resource for the lecture of nuclear reactor theory, and should be studied as common knowledge. Much space is therefore devoted to the history of nuclear energy production and to nuclear physics, and the material focuses on the principles of energy production in nuclear reactors. However, considering the heavy workload of students, these subjects are presented concisely, allowing students to read quickly through this textbook. (J.P.N.)

  14. Nuclear reactor safety in the USA

    International Nuclear Information System (INIS)

    Vigil, J.C.

    1983-01-01

    Nuclear reactor safety in the USA has emphasized a defense-in-depth approach to protecting the public from reactor accidents. This approach was severely tested by the Three Mile Island accident and was found to be effective in safeguarding the public health and safety. However, the economic impact of the TMI accident was very large. Consequently, more attention is now being given to plant protection as well as public-health protection in reactor-safety studies. Sophisticated computer simulations at Los Alamos are making major contributions in this area. In terms of public risk, nuclear power plants compare favorably with other large-scale alternatives to electricity generation. Unfortunately, there is a large gulf between the real risks of nuclear power and the present public perception of these risks

  15. Nuclear power reactor technology

    International Nuclear Information System (INIS)

    1978-09-01

    Risoe National Laboratory was established more than twenty years ago with research and development of nuclear reactor technology as its main objective. The Laboratory has by now accumulated many years of experience in a number of areas vital to nuclear reactor technology. The work and experience of, and services offered by the Laboratory within the following fields are described: Health physics site supervision; Treatment of low and medium level radioactive waste; Core performance evaluation; Transient analysis; Accident analysis; Fuel management; Fuel element design, fabrication and performance evaluation; Non-destructive testing of nuclear fuel; Theoretical and experimental structural analysis; Reliability analysis; Site evaluation. Environmental risk and hazard calculation; Review and analysis of safety documentation. Risoe has already given much assistance to the authorities, utilities and industries in such fields, carrying out work on both light and heavy water reactors. The Laboratory now offers its services to others as a consultant, in education and training of staff, in planning, in qualitative and quantitative analysis, and for the development and specification of fabrication techniques. (author)

  16. Mission operations for unmanned nuclear electric propulsion outer planet exploration with a thermionic reactor spacecraft.

    Science.gov (United States)

    Spera, R. J.; Prickett, W. Z.; Garate, J. A.; Firth, W. L.

    1971-01-01

    Mission operations are presented for comet rendezvous and outer planet exploration NEP spacecraft employing in-core thermionic reactors for electric power generation. The selected reference missions are the Comet Halley rendezvous and a Jupiter orbiter at 5.9 planet radii, the orbit of the moon Io. The characteristics of the baseline multi-mission NEP spacecraft are presented and its performance in other outer planet missions, such as Saturn and Uranus orbiters and a Neptune flyby, are discussed. Candidate mission operations are defined from spacecraft assembly to mission completion. Pre-launch operations are identified. Shuttle launch and subsequent injection to earth escape by the Centaur D-1T are discussed, as well as power plant startup and the heliocentric mission phases. The sequence and type of operations are basically identical for all missions investigated.

  17. Nuclear future: thinking for building. Proceedings of the 12. Brazilian national meeting on reactor physics and thermal hydraulics; 8. General congress on nuclear energy; 5. Brazilian national meeting on nuclear applications

    International Nuclear Information System (INIS)

    2000-01-01

    These proceedings, for the first time, present jointly the 12. Brazilian national meeting on reactor physics and thermal hydraulics (12 ENFIR), 8. General congress on nuclear energy (8. CGEN), and 5. Brazilian national meeting on nuclear applications (5. ENAN). The main theme of discussion was: 'Nuclear Future: thinking for building'. The papers have analysed the progresses of peaceful utilization of nuclear technology and its forecasting for the beginning of the new millennium. The construction of Angra-3 nuclear power plant have been discussed

  18. Towards nuclear fusion reactors

    International Nuclear Information System (INIS)

    1993-11-01

    The results of nuclear fusion researches in JAERI are summarized. In this report, following themes are collected: the concept of fusion reactor (including ITER), fusion reactor safety, plasma confinement, fusion reactor equipment, and so on. Includes glossary. (J.P.N.)

  19. Refuelling nuclear reactors

    International Nuclear Information System (INIS)

    Stacey, J.; Webb, J.; White, W.P.; McLaren, N.H.

    1981-01-01

    An improved nuclear reactor refuelling machine is described which can be left in the reactor vault to reduce the off-load refuelling time for the reactor. The system comprises a gripper device rangeable within a tubular chute, the gripper device being movable by a pantograph. (U.K.)

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

    International Nuclear Information System (INIS)

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

    1993-08-01

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

  1. Water cooled nuclear reactor

    International Nuclear Information System (INIS)

    1975-01-01

    A description is given of a cooling water intake collector for a nuclear reactor. It includes multiple sub-collectors extending out in a generally parallel manner to each other, each one having a first end and a second one separated along their length, and multiple water outlets for connecting each one to a corresponding pressure tube of the reactor. A first end tube and a second one connect the sub-collector tubes together to their first and second ends respectively. It also includes multiple collector tubes extending transversely by crossing over the sub-collector tubes and separated from each other in the direction of these tubes. Each collector tubes has a water intake for connecting to a water pump and multiple connecting tubes separated over its length and connecting each one to the corresponding sub-collector [fr

  2. [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.

  3. Electric capacitance tomography and two-phase flow for the nuclear reactor safety analysis

    International Nuclear Information System (INIS)

    Lee, Jae Young

    2008-01-01

    Recently electric capacitance tomography has been developed to be used in the analysis of two-phase flow. Although its electric field is not focused as the hard ray tomography such as the X-ray or gamma ray, its convenience of easy access to the system and easy maintenance due to no requirement of radiation shielding benefits us in its application in the two-phase flow study, one of important area in the nuclear safety analysis. In the present paper, the practical technologies in the electric capacitance tomography are represented in both parts of hardware and software. In the software part, both forward problem and inverse problem are discussed and the method of regularization. In the hardware part, the brief discussion of the electronics circuits is made which provides femto farad resolution with a reasonable speed (150 frame/sec for 16 electrodes). Some representative ideal cases are studied to demonstrate its potential capability for the two-phase flow analysis. Also, some variations of the tomography such as axial tomography, and three dimensional tomography are discussed. It was found that the present ECT is expected to become a useful tool to understand the complicated three dimensional two-phase flow which may be an important feature to be equipped by the safety analysis codes. (author)

  4. Economics of nuclear electricity

    International Nuclear Information System (INIS)

    Frederick, G.

    1997-01-01

    On the sites of Tihange and Doel in Belgium, a total of seven nuclear generating units with an aggregate installed power of 5807 MWe are operated. Construction of another unit at Doel was postponed indefinitely in 1988 after the Chernobyl accident. Electrabel holds a 25% interest in the Chooz B-1 and B-2 nuclear generating units under construction in France near the Belgian border. In terms of gross installed nuclear generating capacity worldwide, Belgium holds twelfth place; when ranked according to the contribution to public electricity supply, the country holds third place with a 57% share. Before decisions are taken about future nuclear power plants, above all the fuel costs of gas-fired cogeneration plants and the capital costs of nuclear power plants must be weighed. Current evaluation of all costs shows the use of nuclear power for electricity generation to be ten percent more expensive than that of natural gas. However, those responsible in the power supply industry feel that this short-term competitive situation is only one factor out of many others, such as safety issues, diversification in sourcing and deliveries, climatic influences, and employment. The development and construction of advanced reactors will result in the desired cost reduction and lead to a new era of nuclear power, also in Europe. (orig.) [de

  5. Nuclear reactor system for ABWR

    International Nuclear Information System (INIS)

    Miyano, Hiroshi; Kitagawa, Koji

    1997-01-01

    Various tests and measurements were performed during the pre-operational test run of Unit No. 6 of The Tokyo Electric Power Co., Inc.'s Kashiwazaki-Kariwa Nuclear Power Station, the first advanced boiling water reactor (ABWR) unit in the world, and the design and performance adequacy of the ABWR were confirmed. The realization of the ABWR in Japan took about 20 years. It was decided that technologies for the reactor internal pump (RIP) and the fine-motion control rod drive (FMCRD), which had been applied in Europe, would be incorporated in the ABWR aiming at simplification of its structure and operation. These main components were evaluated, modified and verified in consideration of the unique Japanese environment, such as seismic conditions, through a joint study program with Japanese utilities as well as an improvement and standardization program in cooperation with the government. In addition to incorporating RIP and FMCRD technologies, the ABWR also has improved features in terms of the design of the reactor pressure vessel and internals, as well as automated servicing equipment for the RIP, FMCRD, and primary containment vessel. (author)

  6. 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

  7. Nuclear reactor trip system

    International Nuclear Information System (INIS)

    Cook, B.M.

    1982-01-01

    Each parameter of the processes of a nuclear reactor and components operatively associated with it is monitored by a set of four like sensors. A trip system normally operates on a ''two out four'' configuration; i.e., to trip the reactor it is necessary that at least two sensors of a set sense an off-normal parameter. This assumes that all sensors are in normal operating condition. However, when a sensor is in test or is subject to maintenance or is defective or disabled, the ''two out of four''configuration would be reduced to a ''one out of three'' configuration because the affected sensor is taken out of service. This would expose the system to the possibility that a single sensor failure, which may be spurious, will cause a trip of the reactor. To prevent this, it is necessary that the affected sensor be bypassed. If only one sensor is bypassed, the system operates on a ''two out of three'' configuration. With two sensors bypassed, the sensing of an off-normal parameter by a third sensor trips the reactor. The by-pass circuit also disables the circuit coupling the by-passed sensor to the trip circuit. (author)

  8. Concerning partial revision of regulations on installation, operation, etc., of nuclear reactor, etc., for test and research

    International Nuclear Information System (INIS)

    1989-01-01

    To enforce the rules relating to nuclear material protection at nuclear power facilities as covered by the Nuclear Reactor Control Law, which was revised in May last year, orders should be issued by the Prime Minister's Office (or Ministry of International Trade and Industry) to specify the following matters: (1) measures to be carried out by the operators of nuclear facilities to ensure the protection of specially designated nuclear fuel materials, (2) procedures for the application for permission as covered by nuclear material protection rules, and (3) requirements for managers in charge of nuclear material protection. The new regulations should cover the following: (1) rules relating to the business of refining of nuclear fuels, and raw materials for nuclear substances, (2) rules relating to the business of processing of nuclear fuels, (3) rules relating to the installation, operation, etc., of nuclear reactor, etc., for test and research, (4) rules relating to the business of reprocessing of spent fules, (5) rules relating to the business of management of nuclear fuels or waste contaminated with nuclear fuels, and (6) rules relating to the application of nuclear fuels. (N.K)

  9. Nuclear reactor internals arrangement

    International Nuclear Information System (INIS)

    Frisch, E.; Andrews, H.N.

    1976-01-01

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

  10. Indian advanced nuclear reactors

    International Nuclear Information System (INIS)

    Saha, D.; Sinha, R.K.

    2005-01-01

    For sustainable development of nuclear energy, a number of important issues like safety, waste management, economics etc. are to be addressed. To do this, a number of advanced reactor designs as well as fuel cycle technologies are being pursued worldwide. The advanced reactors being developed in India are the AHWR and the CHTR. Both the reactors use thorium based fuel and have many passive features. This paper describes the Indian advanced reactors and gives a brief account of the international initiatives for the sustainable development of nuclear energy. (author)

  11. Development of helium turbines associated with high temperature nuclear reactors for electric power plants

    International Nuclear Information System (INIS)

    Chaboseau, J.

    1975-01-01

    First, the system is defined and found perfectly adapted, in the industrial meaning of the term, to the utilization with high temperature nuclear reactors. Subsequently, a realistic determination of its original feature and that of the turbine is tempted: these features are few, particularly if the existence of numerous, already utilized, direct cycle devices is considered. Then, the influence of main characteristic parameters is evaluated, i.e. the choice of thermodynamic cycle, the utilization of helium, the principle of the layout and the rating of the power station. It appears to be sure that the engineering designs are of great importance for the turbine designer. For the latter, however, the technical aspects of the required developments will be of the conventional type. The character and the importance of these developments will mainly depend on how the system is spread in industry. It seems possible that the reasonable choice of the data of the first generating station allows to reduce the preliminary developments to a minimum amount while ensuring the subsequent evolution of the techniques. The inherent research of perfectness of the system justifies to anticipate long-term developments which are pursued in parallel with the commissioning of subsequent generations of power stations [fr

  12. Reactors. Nuclear propulsion ships

    International Nuclear Information System (INIS)

    Fribourg, Ch.

    2001-01-01

    This article has for object the development of nuclear-powered ships and the conception of the nuclear-powered ship. The technology of the naval propulsion P.W.R. type reactor is described in the article B.N.3 141 'Nuclear Boilers ships'. (N.C.)

  13. Inquiry relating to modifications of reactor installation in Ikata No. 1 and 2 nuclear power plants of Shikoku Electric Power Company, Inc

    International Nuclear Information System (INIS)

    1979-01-01

    Application was made to the Minister of International Trade and Industry for the license relating to the modifications of reactor installation in the Ikata No. 1 and 2 nuclear power plants of the Shikoku Electric Power Company, Inc., on February 13, 1979, from the president of the company. After the safety evaluation was finished by the Ministry of International Trade and Industry, inquiry was conducted to the Atomic Energy Safety Commission (AESC) on June 15, 1979, from the Minister of International Trade and Industry. The investigation and discussion were commenced by the AESC on June 19, 1979. The modifications of the reactor installation are the increase of new fuel storage capacity from about 1/3 to about 2/3 of in-core fuel for No. 1 plant and the modification of driving mechanism from the roller nut type to the magnetic jack type for the control rod cluster for adjusting power distribution in the No. 2 plant. The contents of the safety examination for each item written above are presented. The prevention of criticality is carefully practiced for the new fuel storage by putting fuel assemblies in stainless steel can type racks and locating the fuel assemblies at the proper distance. Relating to the driving mechanism for the control rod cluster adjusting power distribution, the driving speed is not modified and the reliability is kept by carrying out the continuous operation test and the electric power black out test as the demonstration test. The magnetic jack type mechanism has the locking device to prevent reactor tripping at the time of electric power black out, and the cluster is held at the location where the cluster existed at the time of black out. (Nakai, Y.)

  14. Laboratory instrumentation modernization at the WPI Nuclear Reactor Facility

    International Nuclear Information System (INIS)

    1995-01-01

    With partial funding from the Department of Energy (DOE) University Reactor Instrumentation Program several laboratory instruments utilized by students and researchers at the WPI Nuclear Reactor Facility have been upgraded or replaced. Designed and built by General Electric in 1959, the open pool nuclear training reactor at WPI was one of the first such facilities in the nation located on a university campus. Devoted to undergraduate use, the reactor and its related facilities have been since used to train two generations of nuclear engineers and scientists for the nuclear industry. The low power output of the reactor and an ergonomic facility design make it an ideal tool for undergraduate nuclear engineering education and other training. The reactor, its control system, and the associate laboratory equipment are all located in the same room. Over the years, several important milestones have taken place at the WPI reactor. In 1969, the reactor power level was upgraded from 1 kW to 10 kW. The reactor's Nuclear Regulatory Commission operating license was renewed for 20 years in 1983. In 1988, under DOE Grant No. DE-FG07-86ER75271, the reactor was converted to low-enriched uranium fuel. In 1992, again with partial funding from DOE (Grant No. DE-FG02-90ER12982), the original control console was replaced

  15. General Nuclear Medicine

    Science.gov (United States)

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

  16. Nuclear reactor power supply

    International Nuclear Information System (INIS)

    Cook, B.M.

    1982-01-01

    The redundant signals from the sensor assemblies measuring the process parameters of a nuclear reactor power supply are transmitted each in its turn to a protection system which operates to actuate the protection apparatus for signals indicating off-process conditions. Each sensor assembly includes a number of like sensors measuring the same parameters. The sets of process signals derived from the sensor assemblies are each in its turn transmitted from the protection system to the control system which impresses control signals on the reactor or its components to counteract the tendency for conditions to drift off-normal status requiring operation of the protection system. A parameter signal selector is interposed between the protection system and the control system. This selector prevents a parameter signal of a set of signals, which differs from the other parameters signals of the set by more than twice the allowable variation of the sensors which produce the set, from passing to the control system. The selectors include a pair of signal selection units, one unit sending selected process signals to primary control channels and the other sending selected process signals to back-up control channels. Test signals are periodically impressed by a test unit on a selected pair of a selected unit and control channels. When test signals are so impressed the selected control channel is disabled from transmitting control signals to the reactor and/or its associated components. This arrangement eliminates the possibility that a single component failure which may be spurious will cause an inadvertent trip of the reactor during test

  17. Testing FLUKA on neutron activation of Si and Ge at nuclear research reactor using gamma spectroscopy

    Science.gov (United States)

    Bazo, J.; Rojas, J. M.; Best, S.; Bruna, R.; Endress, E.; Mendoza, P.; Poma, V.; Gago, A. M.

    2018-03-01

    Samples of two characteristic semiconductor sensor materials, silicon and germanium, have been irradiated with neutrons produced at the RP-10 Nuclear Research Reactor at 4.5 MW. Their radionuclides photon spectra have been measured with high resolution gamma spectroscopy, quantifying four radioisotopes (28Al, 29Al for Si and 75Ge and 77Ge for Ge). We have compared the radionuclides production and their emission spectrum data with Monte Carlo simulation results from FLUKA. Thus we have tested FLUKA's low energy neutron library (ENDF/B-VIIR) and decay photon scoring with respect to the activation of these semiconductors. We conclude that FLUKA is capable of predicting relative photon peak amplitudes, with gamma intensities greater than 1%, of produced radionuclides with an average uncertainty of 13%. This work allows us to estimate the corresponding systematic error on neutron activation simulation studies of these sensor materials.

  18. Nuclear reactor physics course for reactor operators

    International Nuclear Information System (INIS)

    Baeten, P.

    2006-01-01

    The education and training of nuclear reactor operators is important to guarantee the safe operation of present and future nuclear reactors. Therefore, a course on basic 'Nuclear reactor physics' in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The aim of the basic course on 'Nuclear Reactor Physics for reactor operators' is to provide the reactor operators with a basic understanding of the main concepts relevant to nuclear reactors. Seen the education level of the participants, mathematical derivations are simplified and reduced to a minimum, but not completely eliminated

  19. 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)

  20. Space nuclear reactor electric power. January 1972-May 1990 (A Bibliography from the International Aerospace Abstracts data base). Report for January 1972-May 1990

    International Nuclear Information System (INIS)

    1990-05-01

    This bibliography contains citations concerning studies and conceptual designs of nuclear space power reactors to generate electric power for space missions. The citations cover the technology, safety aspects, and policy considerations. (This updated bibliography contains 280 citations, 172 of which are new entries to the previous edition.)

  1. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    Hindle, E.D.

    1981-01-01

    An array of rods comprising zirconium alloy sheathed nuclear fuel pellets assembled to form a fuel element for a pressurised water reactor is claimed. The helium gas pressure within each rod differs substantially from that of its closest neighbours

  2. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    Hindle, E.D.

    1984-01-01

    The fuel elements for a pressurised water reactor comprise arrays of rods of zirconium alloy sheathed nuclear fuel pellets. The helium gas pressure within each rod differs substantially from that of its closest neighbours

  3. Nuclear reactor core catcher

    International Nuclear Information System (INIS)

    1977-01-01

    A nuclear reactor core catcher is described for containing debris resulting from an accident causing core meltdown and which incorporates a method of cooling the debris by the circulation of a liquid coolant. (U.K.)

  4. Seals in nuclear reactors

    International Nuclear Information System (INIS)

    1979-01-01

    The aim of this invention is the provision of improved seals for reactor vessels in which fuel assemblies are located together with inlets and outlets for the circulation of a coolant. The object is to provide a seal arrangement for the rotatable plugs of nuclear reactor closure heads which has good sealing capacities over a wide gap during operation of the reactor but which also permits uninhibited rotation of the plugs for maintenance. (U.K.)

  5. Results from integral tests of single reformer tubes under simulated nuclear reactor conditions

    International Nuclear Information System (INIS)

    Decken, C.B. von der; Fedders, H.; Harth, R.; Hoehlein, B.; Riensche, E.

    1980-01-01

    The possibility of supplying high temperature heat from a HTGR for process application is being investigated at some places in the world. In all programmes or projects existing with respect to this application, the endothermic steam reforming of methane is one main step in the transmission of heat produced by nuclear fission to different chemical processes. The KFA is involved in the two German projects PNP - Prototypanlage Nukleare Prozesswaerme (Prototype-plant Nuclear Process-heat), and NFE -Nukleare Fernenergie (Long Distance Energy Transport). In a HTGR, helium generally serves as reactor coolant. It transports the heat from the core to the different components which take over this heat for various purposes. In case of arranging a steam reformer in the helium circuit, it is necessary for economic reasons to reach very high temperatures. In the two German projects mentioned above, the helium temperature at HTGR core outlet is determined to 950 0 C. Thus the main design data for a steam reformer supplied by heat from a HTGR are maximum helium temperature 950 0 C, helium pressure 40 bar. By an extensive utilization of the available advanced conventional steam reforming technology, the helium heated steam reformer design is using normal steam reforming tubes arranged in compact bundles

  6. Concerning installation of reactor at Noto nuclear power plant of Hokuriku Electric Power Co., Inc. (reply to inquiry)

    International Nuclear Information System (INIS)

    1988-01-01

    In response to an inquiry on the conformity of the title issue to the applicable law concerning nuclear material, nuclear fuel and nuclear reactor, the Nuclear Safety Commission ordered the Nuclear Reactor Safety Expert Group to make a study, made an examination after receiving a report from the Group, and submit the findings to the Minister of International Trade and Industry. The study on the site conditions covered the site location, geology, effects or earthquakes, meteorology, hydrology and social environment. The study on the safety design of the reactor covered the design of the facilities as a whole, anti-earthquake design, reactor, instrumentation, reactor shutdown system, reactivity control, protection, cooling system, reactor vessel, fuel handling, fuel storage, and radiation control. The study also included exposure evaluation, abnormal transient state analysis and accident analysis. It is concluded that the relevant company has a technical capability required for the installation and proper operation of the reactor in question, and that the reactor will have adequate safety with no possibility of causing a disaster. (Nogami, K.)

  7. Nuclear reactor simulator

    International Nuclear Information System (INIS)

    Baptista, Vinicius Damas

    1996-01-01

    The Nuclear Reactor Simulator was projected to help the basic training in the formation of the Nuclear Power Plants operators. It gives the trainee the opportunity to see the nuclear reactor dynamics. It's specially indicated to be used as the support tool to NPPT (Nuclear Power Preparatory Training) from NUS Corporation. The software was developed to Intel platform (80 x 86, Pentium and compatible ones) working under the Windows operational system from Microsoft. The program language used in development was Object Pascal and the compiler used was Delphi from Borland. During the development, computer algorithms were used, based in numeric methods, to the resolution of the differential equations involved in the process. (author)

  8. Nuclear electricity - a progress report

    International Nuclear Information System (INIS)

    England, G.

    1980-01-01

    A survey of the progress of nuclear power over the past three years reveals three major features: (i) the continued operation of the first generation of commercial nuclear power stations, based on the Magnox gas-cooled reactor; (ii) the introduction and operation of the first of the second-generation stations, based on the advanced gas-cooled reactor (AGR); and (iii) the commitment of two successive Governments to a flexible thermal reactor strategy. Each of these features is considered and a number of related issues, including the safety record and cost savings to the electricity consumer, are discussed. (author)

  9. Neutron noise in nuclear reactors

    International Nuclear Information System (INIS)

    Blaquiere, A.; Pachowska, R.

    1961-06-01

    The power of a nuclear reactor, in the operating conditions, presents fluctuations due to various causes. This random behaviour can be included in the study of 'noises'. Among other sources of noise, we analyse hereafter the fluctuations due: a) to the discontinuous emissions of neutrons from an independent source; b) to the multiplication of neutrons inside the reactor. The method which we present makes use of the analogies between the rules governing a nuclear reactor in operation and a number of radio-electrical systems, in particular the feed-back loops. The reactor can be characterized by its 'passing band' and is described as a system submitted to a sequence of random pulses. In non linear operating condition, the effect of neutron noise is defined by means of a non-linear functional, this theory is thus related to previous works the references of which are given at the end of the present report. This leads us in particular in the case of nuclear reactors to some results given by A. Blaquiere in the case of radio-electrical loops. (author) [fr

  10. Nuclear reactor PBMR and cogeneration

    International Nuclear Information System (INIS)

    Ramirez S, J. R.; Alonso V, G.

    2013-10-01

    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)

  11. MANAGEMENT OF RESEARCH AND TEST REACTOR ALUMINUM SPENT NUCLEAR FUEL - A TECHNOLOGY ASSESSMENT

    Energy Technology Data Exchange (ETDEWEB)

    Vinson, D.

    2010-07-11

    The Department of Energy's Environmental Management (DOE-EM) Program is responsible for the receipt and storage of aluminum research reactor spent nuclear fuel or used fuel until ultimate disposition. Aluminum research reactor used fuel is currently being stored or is anticipated to be returned to the U.S. and stored at DOE-EM storage facilities at the Savannah River Site and the Idaho Nuclear Technology and Engineering Center. This paper assesses the technologies and the options for safe transportation/receipt and interim storage of aluminum research reactor spent fuel and reviews the comprehensive strategy for its management. The U.S. Department of Energy uses the Appendix A, Spent Nuclear Fuel Acceptance Criteria, to identify the physical, chemical, and isotopic characteristics of spent nuclear fuel to be returned to the United States under the Foreign Research Reactor Spent Nuclear Fuel Acceptance Program. The fuel is further evaluated for acceptance through assessments of the fuel at the foreign sites that include corrosion damage and handleability. Transport involves use of commercial shipping casks with defined leakage rates that can provide containment of the fuel, some of which are breached. Options for safe storage include wet storage and dry storage. Both options must fully address potential degradation of the aluminum during the storage period. This paper focuses on the various options for safe transport and storage with respect to technology maturity and application.

  12. Thermionic nuclear reactor systems

    International Nuclear Information System (INIS)

    Kennel, E.B.

    1986-01-01

    Thermionic nuclear reactors can be expected to be candidate space power supplies for power demands ranging from about ten kilowatts to several megawatts. The conventional ''ignited mode'' thermionic fuel element (TFE) is the basis for most reactor designs to date. Laboratory converters have been built and tested with efficiencies in the range of 7-12% for over 10,000 hours. Even longer lifetimes are projected. More advanced capabilities are potentially achievable in other modes of operation, such as the self-pulsed or unignited diode. Coupled with modest improvements in fuel and emitter material performance, the efficiency of an advanced thermionic conversion system can be extended to the 15-20% range. Advanced thermionic power systems are expected to be compatible with other advanced features such as: (1) Intrinsic subcritically under accident conditions, ensuring 100% safety upon launch abort; (2) Intrinsic low radiation levels during reactor shutdown, allowing manned servicing and/or rendezvous; (3) DC to DC power conditioning using lightweight power MOSFETS; and (4) AC output using pulsed converters

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

    International Nuclear Information System (INIS)

    Gomez, A.

    2008-01-01

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

  14. Alteration of installation of nuclear reactors (alteration of No. 1 and No. 2 reactor facilities) in Shimane Nuclear Power Station, Chugoku Electric Power Co. , Inc. (report)

    Energy Technology Data Exchange (ETDEWEB)

    1987-07-01

    The Nuclear Safety Commission reported to the Minister of International Trade and Industry after prudent deliberation on this alteration problem which had been inquired on August 28, 1986. It was recognized that the technical capability of the applicant is appropriate, as the result of deliberation. It was judged that the safety after this alteration of the installation of the reactor facilities can be ensured. As the result of examining new type 8 x 8 zirconium liner fuel, there was no problem in its mechanical design, the analysis of dynamical characteristics, and the analysis of abnormal transient change and accident in operation. As to the change of the average degree of enrichment of replacement fuel, the thermonuclear design of the reactor core was adequate. In the incineration of spent resin and filter sludge, the effect of radioactive substances to the environment was negligible. The safety after abolishing the auxiliary protection function against exhaust radioactivity is ensured with a rare gas holdup equipment. The soundness of fuel and the soundness of reactor coolant pressure boundary are maintained in abnormal transient change and accident. (Kako, I.).

  15. Alteration of installation of nuclear reactors (alteration of No.1 and No.2 reactor facilities) in Shimane Nuclear Power Station, Chugoku Electric Power Co., Inc. (report)

    International Nuclear Information System (INIS)

    1987-01-01

    The Nuclear Safety Commission reported to the Minister of International Trade and Industry after prudent deliberation on this alteration problem which had been inquired on August 28, 1986. It was recognized that the technical capability of the applicant is appropriate, as the result of deliberation. It was judged that the safety after this alteration of the installation of the reactor facilities can be ensured. As the result of examining new type 8 x 8 zirconium liner fuel, there was no problem in its mechanical design, the analysis of dynamical characteristics, and the analysis of abnormal transient change and accident in operation. As to the change of the average degree of enrichment of replacement fuel, the thermonuclear design of the reactor core was adequate. In the incineration of spent resin and filter sludge, the effect of radioactive substances to the environment was negligible. The safety after abolishing the auxiliary protection function against exhaust radioactivity is ensured with a rare gas holdup equipment. The soundness of fuel and the soundness of reactor coolant pressure boundary are maintained in abnormal transient change and accident. (Kako, I.)

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

    International Nuclear Information System (INIS)

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

    2012-02-01

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

  17. Technical evaluation report on the monitoring of electric power to the reactor-protection system for the Pilgrim Nuclear Power Station

    International Nuclear Information System (INIS)

    Selan, J.C.

    1982-01-01

    This report documents the technical evaluation of the monitoring of electric power to the reactor protection system (RPS) at the Pilgrim Nuclear Power Station. The evaluation is to determine if the proposed design modification will protect the RPS from abnormal voltage and frequency conditions which could be supplied from the power supplies and will meet certain requirements set forth by the Nuclear Regulatory Commission. The proposed design modifications will protect the RPS from sustained abnormal voltage and frequency conditions from the supplying sources

  18. Expert systems for the analysis of transients on nuclear reactors: crisis analysis, sextant, a general purpose physical analyser

    International Nuclear Information System (INIS)

    Barbet, N.; Dumas, M.; Mihelich, G.; Souchet, Y.; Thomas, J.B.

    1987-04-01

    Two developments of expert systems intended to work on line to the analysis of nuclear reactor transients are reported. During an hypothetical crisis occurring in a nuclear facility, a staff of the Institute for Protection and Nuclear Safety (IPSN) has to assess the risk to local population. The expert system is intended to work as an assistant to the staff. At the present time, it deals with the availability of the safety systems of the plant (e.g. ECCS), depending on the functional state of the support systems. A next step is to take into account the physical transient of the reactor (mass and energy balance, pressure, flows). In order to reach this goal as in the development of other similar expert systems, a physical analyser is required. This is the aim of SEXTANT, which combines several knowledge bases concerning measurements, models and qualitative behaviour of the plant with a mechanism of conjecture-refutation and a set of simplified models matching the current physical state. A prototype is under assessment by dealing with integral test facility transients. Both expert systems require powerful shells for their development. SPIRAL is such a toolkit for the development of expert systems devoted to the computer aided management of complex processes

  19. Method of 16N generation for test of radiation controlled channels at nuclear power stations with water-cooled reactors

    International Nuclear Information System (INIS)

    Khryachkov, V.A.; Bondarenko, I.P.; Dvornikov, P.A.; Zhuravlev, B.V.; Kovtun, S.N.; Khromyleva, T.A.; Pavlov, A.V.; Roshchin, N.G.

    2012-01-01

    The preferences of nuclear reaction use for radiation control channels test in water-cooled power reactors have been analyzed in the paper. The new measurements for more accurate determination of reaction cross section energy dependence have been carried out. A set of new methods for background reducing and improvement of events determination reliability has also been developed [ru

  20. HOMOGENEOUS NUCLEAR POWER REACTOR

    Science.gov (United States)

    King, L.D.P.

    1959-09-01

    A homogeneous nuclear power reactor utilizing forced circulation of the liquid fuel is described. The reactor does not require fuel handling outside of the reactor vessel during any normal operation including complete shutdown to room temperature, the reactor being selfregulating under extreme operating conditions and controlled by the thermal expansion of the liquid fuel. The liquid fuel utilized is a uranium, phosphoric acid, and water solution which requires no gus exhaust system or independent gas recombining system, thereby eliminating the handling of radioiytic gas.

  1. Nuclear reactor safety systems

    International Nuclear Information System (INIS)

    Ball, R.M.; Roberts, R.C.

    1980-01-01

    A safety system for shutting down a nuclear reactor under overload conditions is described. The system includes a series of parallel-connected computer memory type look-up tables each of which receives data on a particular reactor parameter and in each of which a precalculated functional value for that parameter is stored indicative of the percentage of maximum reactor load that the parameter contributes. The various functional values corresponding to the actual measured parameters are added together to provide a control signal used to shut down the reactor under overload conditions. (U.K.)

  2. Decommissioning of the research nuclear reactor WWR-S Magurele - Bucharest. General presentation of the project

    International Nuclear Information System (INIS)

    Dragulescu, Emilian; Dragusin, Mitica; Popa, Victor; Boicu, Alin; Tuca, Carmen; Iorga, Ioan; Vrabie, Ionut; Mustata, Carmen

    2003-01-01

    A decommissioning project was worked out concerning the nuclear facility research reactor WWR-S Magurele-Bucharest to remove the radioactive and hazardous materials and so to exclude any risk for human health and environment. The project involves the four phases named assessment, development, operations and closeout. There are two major parts to the assesment phase: preliminary characterisation and the review and decision-making process. Characterisation is needed to develop project baseline data, which should include sufficient chemical, physical, and radiological characterisation to meet planning needs. Based on the conclusions of these studies, possible decommissioning alternative will be analyzed and: the best alternative chosen, final goal identified, risk assessments are evaluated. Also, taken into account are: regulations supporting assessment, land use considerations, financial concerns, disposal availability, public involvement, technology developments. After a decommissioning alternative was chosen, detailed engineering will begin following appropriate regulatory guidance. The plan will include characterisation information, namely: review of decommissioning alternatives; justification for the selected alternative; provision for regulatory compliance; predictions of personnel exposure, radioactive waste volume, and cost. Other activities are: scheduling, preparation for decommissioning operations; coordination, documentation, characterization report, feasibility studies, Decommissioning Plan, project daily report, radiological survey, airborne sampling records, termination survey of the site. The operations imply: identification and sequencing the operations on contaminated materials, storing on site the wastes, awaiting processing or disposal, and packaging of materials for transport to processing or disposal facilities.The key operations are: worker protection, health and safety program, review of planing work, work area assessment, work area controls

  3. Energy from nuclear reactors

    International Nuclear Information System (INIS)

    Hospe, J.

    1977-01-01

    This VDI-Nachrichten series has the target to provide a technical-objective basis for the discussion of the pros and cons of nuclear power. The first part deals with LWR-type reactors which so far have prevailed in nuclear power generation. (orig.) [de

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

    International Nuclear Information System (INIS)

    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

  5. Comparison of Direct and Indirect Gas Reactor Brayton Systems for Nuclear Electric Space Propulsion

    International Nuclear Information System (INIS)

    M Postlehwait; P DiLorenzo; S Belanger; J Ashcroft

    2005-01-01

    Gas reactor systems are being considered as candidates for use in generating power for the Prometheus-1 spacecraft, along with other NASA missions as part of the Prometheus program. Gas reactors offer a benign coolant, which increases core and structural materials options. However, the gas coolant has inferior thermal transport properties, relative to other coolant candidates such as liquid metals. This leads to concerns for providing effective heat transfer and for minimizing pressure drop within the reactor core. In direct gas Brayton systems, i.e. those with one or more Brayton turbines in the reactor cooling loop, the ability to provide effective core cooling and low pressure drop is further constrained by the need for a low pressure, high molecular weight gas, typically a mixture of helium and xenon. Use of separate primary and secondary gas loops, one for the reactor and one or more for the Brayton system(s) separated by heat exchanger(s), allows for independent optimization of the pressure and gas composition of each loop. The reactor loop can use higher pressure pure helium, which provides improved heat transfer and heat transport properties, while the Brayton loop can utilize lower pressure He-Xe. However, this approach requires a separate primary gas circulator and also requires gas to gas heat exchangers. This paper focuses on the trade-offs between the direct gas reactor Brayton system and the indirect gas Brayton system. It discusses heat exchanger arrangement and materials options and projects heat exchanger mass based on heat transfer area and structural design needs. Analysis indicates that these heat exchangers add considerable mass, but result in reactor cooling and system resiliency improvements

  6. A generalized adjoint framework for sensitivity and global error estimation in time-dependent nuclear reactor simulations

    International Nuclear Information System (INIS)

    Stripling, H.F.; Anitescu, M.; Adams, M.L.

    2013-01-01

    Highlights: ► We develop an abstract framework for computing the adjoint to the neutron/nuclide burnup equations posed as a system of differential algebraic equations. ► We validate use of the adjoint for computing both sensitivity to uncertain inputs and for estimating global time discretization error. ► Flexibility of the framework is leveraged to add heat transfer physics and compute its adjoint without a reformulation of the adjoint system. ► Such flexibility is crucial for high performance computing applications. -- Abstract: We develop a general framework for computing the adjoint variable to nuclear engineering problems governed by a set of differential–algebraic equations (DAEs). The nuclear engineering community has a rich history of developing and applying adjoints for sensitivity calculations; many such formulations, however, are specific to a certain set of equations, variables, or solution techniques. Any change or addition to the physics model would require a reformulation of the adjoint problem and substantial difficulties in its software implementation. In this work we propose an abstract framework that allows for the modification and expansion of the governing equations, leverages the existing theory of adjoint formulation for DAEs, and results in adjoint equations that can be used to efficiently compute sensitivities for parametric uncertainty quantification. Moreover, as we justify theoretically and demonstrate numerically, the same framework can be used to estimate global time discretization error. We first motivate the framework and show that the coupled Bateman and transport equations, which govern the time-dependent neutronic behavior of a nuclear reactor, may be formulated as a DAE system with a power constraint. We then use a variational approach to develop the parameter-dependent adjoint framework and apply existing theory to give formulations for sensitivity and global time discretization error estimates using the adjoint

  7. Comments on nuclear reactor safety in Ontario

    International Nuclear Information System (INIS)

    1987-08-01

    The Chalk River Technicians and Technologists Union representing 500 technical employees at the Chalk River Nuclear Laboratories of AECL submit comments on nuclear reactor safety to the Ontario Nuclear Safety Review. Issues identified by the Review Commissioner are addressed from the perspective of both a labour organization and experience in the nuclear R and D field. In general, Local 1568 believes Ontario's CANDU nuclear reactors are not only safe but also essential to the continued economic prosperity of the province

  8. Nuclear research reactors in the world. June 1988 ed.

    International Nuclear Information System (INIS)

    1988-01-01

    This is the third edition of Reference Data Series No. 3, Nuclear Research Reactors in the World, which replaces the Agency's publications Power and Research Reactors in Member States and Research Reactors in Member States. This booklet contains general information, as of the end of June 1988, on research reactors in operation, under construction, planned, and shut down. The information is collected by the Agency through questionnaires sent to the Member States through the designated national correspondents. All data on research reactors, training reactors, test reactors, prototype reactors and critical assemblies are stored in the IAEA Research Reactor Data Base (RRDB) system. This system contains all the information and data previously published in the Agency's publication Power and Research Reactors in Member States as well as additional information. 12 figs, 19 tabs

  9. Nuclear reactor pressure vessel integrity insurance by crack arrestability evaluation using load from CVN tests

    International Nuclear Information System (INIS)

    Fabry, A.

    1997-01-01

    The present work is undertaken in the framework of nuclear reactor pressure vessel (RPV) surveillance and aims at revisiting the crack arrest approach to structural integrity insurance. This approach, performed under normal plant operation conditions, can also offer an attractive alternative to the crack initiation philosophy promoted for accidental analysis. To this end, an accidental conservative, cost effective and robust methodology is forwarded and demonstrated: it makes use of the crack arrest information contained in the instrumented Charpy V-notch impact test and/or in the shear fracture appearance of broken samples. Particular attention is paid to the appraisal of uncertainties and the related safety margin. The resulting capability is placed in perspective with the state-of-the-art crack initiation methodology based on the slow bend testing of recracked specimens, presently under standardization world-wide. The investigation leads to highlight three conceptual weaknesses of current enfgineering and regulatory practices. Improved crack arrestability evaluation emerges as an optimal approach to insure safe PWR operation up to design end-of-life and beyond

  10. Nuclear reactor pressure vessel integrity insurance by crack arrestability evaluation using load from CVN tests

    Energy Technology Data Exchange (ETDEWEB)

    Fabry, A.

    1997-10-15

    The present work is undertaken in the framework of nuclear reactor pressure vessel (RPV) surveillance and aims at revisiting the crack arrest approach to structural integrity insurance. This approach, performed under normal plant operation conditions, can also offer an attractive alternative to the crack initiation philosophy promoted for accidental analysis. To this end, an accidental conservative, cost effective and robust methodology is forwarded and demonstrated: it makes use of the crack arrest information contained in the instrumented Charpy V-notch impact test and/or in the shear fracture appearance of broken samples. Particular attention is paid to the appraisal of uncertainties and the related safety margin. The resulting capability is placed in perspective with the state-of-the-art crack initiation methodology based on the slow bend testing of recracked specimens, presently under standardization world-wide. The investigation leads to highlight three conceptual weaknesses of current enfgineering and regulatory practices. Improved crack arrestability evaluation emerges as an optimal approach to insure safe PWR operation up to design end-of-life and beyond.

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

    International Nuclear Information System (INIS)

    1995-05-01

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

  12. Nuclear Power Plant Thermocouple Sensor-Fault Detection and Classification Using Deep Learning and Generalized Likelihood Ratio Test

    Science.gov (United States)

    Mandal, Shyamapada; Santhi, B.; Sridhar, S.; Vinolia, K.; Swaminathan, P.

    2017-06-01

    In this paper, an online fault detection and classification method is proposed for thermocouples used in nuclear power plants. In the proposed method, the fault data are detected by the classification method, which classifies the fault data from the normal data. Deep belief network (DBN), a technique for deep learning, is applied to classify the fault data. The DBN has a multilayer feature extraction scheme, which is highly sensitive to a small variation of data. Since the classification method is unable to detect the faulty sensor; therefore, a technique is proposed to identify the faulty sensor from the fault data. Finally, the composite statistical hypothesis test, namely generalized likelihood ratio test, is applied to compute the fault pattern of the faulty sensor signal based on the magnitude of the fault. The performance of the proposed method is validated by field data obtained from thermocouple sensors of the fast breeder test reactor.

  13. Test reactors in the world

    International Nuclear Information System (INIS)

    Corella, M.R.; Gomez Alonso, M.

    1983-01-01

    INFCE work on research reactor core conversion from HEU to LEU, attracted a raising interest on this type of nuclear reactors. In this context, the present work shows a compilation of worldwide research and test nuclear reactors, now in operation, under construction, or planned, as well as decommissioned reactors (tables A to F). Brief descriptions of these reactors are included in tables G to L. In table M a summary view of reactors with power level between 10 and 30 MWt is shown. Attention is focused on that power range, as it has been considered in very preliminar studies for a new research reactor. Almost all data have been obtained from current available bibliography. (author)

  14. Nuclear reactor safety research in Idaho

    International Nuclear Information System (INIS)

    Zeile, H.J.

    1983-01-01

    Detailed information about the performance of nuclear reactor systems, and especially about the nuclear fuel, is vital in determining the consequences of a reactor accident. Fission products released from the fuel during accidents are the ultimate safety concern to the general public living in the vicinity of a nuclear reactor plant. Safety research conducted at the Idaho National Engineering Laboratory (INEL) in support of the U.S. Nuclear Regulatory Commission (NRC) has provided the NRC with detailed data relating to most of the postulated nuclear reactor accidents. Engineers and scientists at the INEL are now in the process of gathering data related to the most severe nuclear reactor accident - the core melt accident. This paper describes the focus of the nuclear reactor safety research at the INEL. The key results expected from the severe core damage safety research program are discussed

  15. Pump selection and application in a pressurized water reactor electric generating plant

    International Nuclear Information System (INIS)

    Kitch, D.M.

    1985-01-01

    Various pump applications utilized in a nuclear pressurized water reactor electric generating plant are described. Emphasis is on pumps installed in the auxiliary systems of the primary nuclear steam supply system. Hydraulic and mechanical details, the ASME Code (Nuclear Design), materials, mechanical seals, shaft design, seismic qualification, and testing are addressed

  16. Present status of space nuclear reactor

    International Nuclear Information System (INIS)

    Kaneko, Yoshihiko

    1996-01-01

    USA and former USSR led space development, and had the experience of launching nuclear reactor satellites. In USA, the research and development of space nuclear reactor were advanced mainly by NASA, and in 1965, the nuclear reactor for power source ''SNAP-10A'' was launched and put on the orbit around the earth. Thereafter, the reactor was started up, and the verifying test at 500 We was successfully carried out. Also for developing the reactor for thermal propulsion, NERVA/ROVER project was done till 1973, and the technological basis was established. The space Exploration Initiative for sending mankind to other solar system planets than the earth is the essential point of the future projects. In former USSR, the ground experiment of the reactor for 800 We power source ''Romashka'', the development of the reactor for 10 kWe power source ''Topaz-1 and 2'', the flight of the artificial satellites, Cosmos 954 and Cosmos 1900, on which nuclear reactors were mounted, and the operation of 33 ocean-monitoring satellites ''RORSAT'' using small fast reactors were carried out. The mission of space development and the nuclear reactors as power source, the engineering of space nuclear reactors, the present status and the trend of space nuclear reactor development, and the investigation by the UN working group on the safety problem of space nuclear reactors are described. (K.I.)

  17. 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)

  18. Nuclear reactors and fuel cycle

    International Nuclear Information System (INIS)

    2014-01-01

    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)

  19. Spent nuclear fuel discharges from US reactors 1992

    International Nuclear Information System (INIS)

    1994-01-01

    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

  20. Physically - engineering problems of the Salaspils Nuclear reactor: Solutions and their topicality

    International Nuclear Information System (INIS)

    Mozgirs, Z.V.

    2005-01-01

    The paper generalizes technical solutions of physically-engineering problems of the Salaspils nuclear research reactor, experience of its modernization and exploitation. New equipment and the related technical solutions have been tested at the Salaspils reactor during its operation time and are now recommended for further use at nuclear reactors. (author)

  1. Nuclear future: thinking for building. Proceedings of the 5. Brazilian national meeting on nuclear applications; 8. General congress on nuclear energy; 12. Brazilian national meeting on reactor physics and thermal hydraulics

    International Nuclear Information System (INIS)

    2000-01-01

    These proceedings, for the first time, present jointly the 12. Brazilian national meeting on reactor physics and thermal hydraulics (12. ENFIR), the 8. General congress on nuclear energy (8. CGEN), and the 5. Brazilian national meeting on nuclear applications (5. ENAN). The main theme of discussion was: 'Nuclear Future: thinking for building'. The papers have analysed the progresses of peaceful utilization of nuclear technology and its forecasting for the beginning of the new millennium. The construction of Angra-3 nuclear power plant have been discussed

  2. 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.

  3. A cermet fuel reactor for nuclear thermal propulsion

    International Nuclear Information System (INIS)

    Kruger, G.

    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 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

  4. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    Butterfield, C.E.; Waite, E.

    1982-01-01

    A nuclear reactor fuel element comprising a column of vibration compacted fuel which is retained in consolidated condition by a thimble shaped plug. The plug is wedged into gripping engagement with the wall of the sheath by a wedge. The wedge material has a lower coefficient of expansion than the sheath material so that at reactor operating temperature the retainer can relax sufficient to accommodate thermal expansion of the column of fuel. (author)

  5. Nuclear reactor monitoring system

    International Nuclear Information System (INIS)

    Drummond, C.N.; Bybee, R.T.; Mason, F.L.; Worsham, H.J.

    1976-01-01

    The invention pertains to an improved monitoring system for the neutron flux in a nuclear reactor. It is proposed to combine neutron flux detectors, a thermoelement, and a background radiation detector in one measuring unit. The spatial arrangement of these elements is fixed with great exactness; they are enclosed by an elastic cover and are brought into position in the reactor with the aid of a bent tube. The arrangement has a low failure rate and is easy to maintain. (HP) [de

  6. Technology of nuclear reactors

    International Nuclear Information System (INIS)

    Ravelet, F.

    2016-01-01

    This academic report for graduation in engineering first presents operation principles of a nuclear reactor core. It presents core components, atomic nuclei, the notions of transmutation and radioactivity, quantities used to characterize ionizing radiations, the nuclear fission, statistical aspects of fission and differences between fast and slow neutrons, a comparison between various heat transfer fluids, the uranium enrichment process, and different types of reactor (boiling water, natural uranium and heavy water, pressurized water, and fourth generation). Then, after having recalled the French installed power, the author proposes an analysis of a typical 900 MWe nuclear power plant: primary circuit, reactor, fuel, spent fuel, pressurizer and primary pump, secondary circuit, aspects related to control-command, regulation, safety and exploitation. The last part proposes a modelling of the thermodynamic cycle of a pressurized water plant by using an equivalent Carnot cycle, a Rankine cycle, and a two-phase expansion cycle with drying-overheating

  7. Process for testing noise emission from containers or pipelines made of steel, particularly for nuclear reactor plants

    International Nuclear Information System (INIS)

    Votava, E.; Stipsits, G.; Sommer, R.

    1982-01-01

    In a process for noise emission testing of steel containers or pipelines, particularly for testing primary circuit components of nuclear reactor plants, measuring sensors and/or associated electronic amplifiers are used, which are tuned for receiving the frequency band of the sound emission spectrum above a limiting frequency f G , but are limited or non-resonant for frequency bands less than f G . (orig./HP) [de

  8. 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.

  9. Nuclear reactor buildings

    International Nuclear Information System (INIS)

    Nagashima, Shoji; Kato, Ryoichi.

    1985-01-01

    Purpose: To reduce the cost of reactor buildings and satisfy the severe seismic demands in tank type FBR type reactors. Constitution: In usual nuclear reactor buildings of a flat bottom embedding structure, the flat bottom is entirely embedded into the rock below the soils down to the deck level of the nuclear reactor. As a result, although the weight of the seismic structure can be decreased, the amount of excavating the cavity is significantly increased to inevitably increase the plant construction cost. Cross-like intersecting foundation mats are embedded to the building rock into a thickness capable withstanding to earthquakes while maintaining the arrangement of equipments around the reactor core in the nuclear buildings required by the system design, such as vertical relationship between the equipments, fuel exchange systems and sponteneous drainings. Since the rock is hard and less deformable, the rigidity of the walls and the support structures of the reactor buildings can be increased by the embedding into the rock substrate and floor responsivity can be reduced. This enables to reduce the cost and increasing the seismic proofness. (Kamimura, M.)

  10. Simulator for materials testing reactors

    International Nuclear Information System (INIS)

    Takemoto, Noriyuki; Sugaya, Naoto; Ohtsuka, Kaoru; Hanakawa, Hiroki; Onuma, Yuichi; Hosokawa, Jinsaku; Hori, Naohiko; Kaminaga, Masanori; Tamura, Kazuo; Hotta, Kohji; Ishitsuka, Tatsuo

    2013-06-01

    A real-time simulator for both reactor and irradiation facilities of a materials testing reactor, “Simulator of Materials Testing Reactors”, was developed for understanding reactor behavior and operational training in order to utilize it for nuclear human resource development and to promote partnership with developing countries which have a plan to introduce nuclear power plant. The simulator is designed based on the JMTR (Japan Materials Testing Reactor), and it simulates operation, irradiation tests and various kinds of anticipated operational transients and accident conditions caused by the reactor and irradiation facilities. The development of the simulator was sponsored by the Japanese government as one of the specialized projects of advanced research infrastructure in order to promote basic as well as applied researches. This report summarizes the simulation components, hardware specification and operation procedure of the simulator. (author)

  11. Prospect of realizing nuclear fusion reactors

    International Nuclear Information System (INIS)

    1989-01-01

    This Report describes the results of the research work on nuclear fusion, which CRIEPI has carried out for about ten years from the standpoint of electric power utilities, potential user of its energy. The principal points are; (a) economic analysis (calculation of costs) based on Japanese analysis procedures and database of commercial fusion reactors, including fusion-fission hybrid reactors, and (b) conceptual design of two types of hybrid reactors, that is, fission-fuel producing DMHR (Demonstration Molten-Salt Hybrid Reactor) and electric-power producing THPR (Tokamak Hybrid Power Reactor). The Report consists of the following chapters: 1. Introduction. 2. Conceptual Design of Hybrid Reactors. 3. Economic Analysis of Commercial Fusion Reactors. 4. Basic Studies Applicable Also to Nuclear Fusion Technology. 5. List of Published Reports and Papers; 6. Conclusion. Appendices. (author)

  12. Nuclear reactor safety system

    International Nuclear Information System (INIS)

    Ball, R.M.; Roberts, R.C.

    1983-01-01

    The invention provides a safety system for a nuclear reactor which uses a parallel combination of computer type look-up tables each of which receives data on a particular parameter (from transducers located in the reactor system) and each of which produces the functional counterpart of that particular parameter. The various functional counterparts are then added together to form a control signal for shutting down the reactor. The functional counterparts are developed by analysis of experimental thermal and hydraulic data, which are used to form expressions that define safe conditions

  13. Management and storage of spent nuclear fuel at research and test reactors. Proceedings of an advisory group meeting

    International Nuclear Information System (INIS)

    1996-08-01

    Irradiated fuel from research and test reactors has been stored at various facilities for several decades. As these facilities age and approach or exceed their original design lifetimes, there is mounting concern about closure of the fuel cycle and about the integrity of ageing fuels from the materials point of view as well as some concern about the loss of self-protection of the fuels as their activity decays. It is clear that an international effort is necessary to give these problems sufficient exposure and to ensure that work continues on appropriate solutions. The future of nuclear research, with its many benefits to mankind, is in jeopardy in some countries, especially countries without nuclear power programmes, because effective solutions for extended interim storage and final disposition of spent research reactor fuels are not yet available. An advisory Group meeting was convened in Vienna to consider a Database on the Management and Storage of Spent Nuclear Fuel from Research and Test Reactors. Sixteen experts from sixteen different countries participated in the Advisory Group meeting and presented country reports, which together represent an overview of the technologies used in spent fuel management and storage at research and test reactors world-wide. The sixteen country reports together with the database summary are presented in this publication. Refs, figs, tabs

  14. Continuous parameter determination of irradiated nuclear fuels in the test-reactor

    International Nuclear Information System (INIS)

    Bevilacqua, A.; Junod, E.; Mas, P.; Perdreau, R.

    1977-01-01

    During the irradiation tests of nuclear fuels, the flux level may often be variable by shifting the loops in a high neutron-gradient. So integral fluence measurements are no longer sufficient. The self-powered neutron detectors allow to finely scan instantaneous fluxes. More than 100 such SPN detectors are used on the experiments irradiated in the SILOE reactor. The treatment of the large amount of information is following. A first minicomputer scans all the measurement lines at a variable frequence (10 min to 1 hr) and writes rough voltage values on a magnetic disk. A second computer does a sorting of these values for each set of SPND corresponding to an experiment. At the present time, the main treatment is performed by batch processing by some FORTRAN codes to get time-evolving quantities, such as effective flux, fission power, burn-up, fission product activities, etc. The future development of the system will allow some of these calculations to be performed directly on the second computer in such a manner to control the movements of the loops automatically in view of a given irradiation program

  15. A nuclear reactor for district heating

    International Nuclear Information System (INIS)

    Bancroft, A.R.; Fenton, N.

    1989-07-01

    Global energy requirements are expected to double over the next 40 years. In the northern hemisphere, many countries consume in excess of 25 percent of their primary energy supply for building heating. Satisfying this need, within the constraints now being acknowledged for sustainable global development, provides an important opportunity for district heating. Fuel-use flexibility, energy and resource conservation, and reduced atmospheric pollution from acid gases and greenhouse gases, are important features offered by district heating systems. Among the major fuel options, only hydro-electricity and nuclear heat completely avoid emissions of combustion gases. To fill the need for an economical nuclear heat source, Atomic Energy of Canada Limited has designed a 10 MW plant that is suitable as a heat source within a network or as the main supply to large individual users. Producing hot water at temperatures below 100 degrees C, it incorporates a small pool-type reactor based on AECL's successful SLOWPOKE Research Reactor. A 2 MW prototype for the commercial unit is now being tested at the Whiteshell Nuclear Research Establishment in Manitoba. With capital costs of $7 million (Canadian), unit energy costs are projected to be $0.02/kWh for a 10 MW unit operating in a heating grid over a 30-year period. By keeping the reactor power low and the water temperature below 100 degrees C, much of the complexity of the large nuclear power plants can be avoided, thus allowing these small, safe nuclear heating systems to be economically viable

  16. Nuclear reactor control column

    International Nuclear Information System (INIS)

    Bachovchin, D.M.

    1982-01-01

    The nuclear reactor control column comprises a column disposed within the nuclear reactor core having a variable cross-section hollow channel and containing balls whose vertical location is determined by the flow of the reactor coolant through the column. The control column is divided into three basic sections wherein each of the sections has a different cross-sectional area. The uppermost section of the control column has the greatest crosssectional area, the intermediate section of the control column has the smallest cross-sectional area, and the lowermost section of the control column has the intermediate cross-sectional area. In this manner, the area of the uppermost section can be established such that when the reactor coolant is flowing under normal conditions therethrough, the absorber balls will be lifted and suspended in a fluidized bed manner in the upper section. However, when the reactor coolant flow falls below a predetermined value, the absorber balls will fall through the intermediate section and into the lowermost section, thereby reducing the reactivity of the reactor core and shutting down the reactor

  17. Requirements of coolants in nuclear reactors

    International Nuclear Information System (INIS)

    Abass, O. A. M.

    2014-11-01

    This study discussed the purposes and types of coolants in nuclear reactors to generate electricity. The major systems and components associated with nuclear reactors are cooling system. There are two major cooling systems utilized to convert the heat generated in the fuel into electrical power. The primary system transfers the heat from the fuel to the steam generator, where the secondary system begins. The steam formed in the steam generator is transferred by the secondary system to the main turbine generator, where it s converted into electricity after passing through the low pressure turbine. There are various coolants used in nuclear reactors-light water, heavy water and liquid metal. The two major types of water-cooled reactors are pressurized water reactors (PWR) and boiling water reactors (BWR) but pressurized water reactors are more in the world. Also discusses this study the reactors and impact of the major nuclear accidents, in the April 1986 disaster at the Chernobyl nuclear power plant in Ukraine was the product operators, and in the March 2011 at the Fukushima nuclear power plant in Japan was the product of earthquake of magnitude 9.0, the accidents caused the largest uncontrolled radioactive release into the environment.(Author)

  18. Inquiry relating to modification of reactor installation of Hamaoka No. 2 nuclear power plant of Chubu Electric Power Company, Inc

    International Nuclear Information System (INIS)

    1979-01-01

    Application was made to the Minister of International Trade and Industry for the license relating to the modification of reactor installation of the Hamaoka No. 2 nuclear power plant, Chubu Electric Power Company, Inc., on February 8, 1979, from the president of the company. After the safety evaluation in the Ministry of International Trade and Industry was finished, inquiry was conducted to the head of the Atomic Energy Safety Commission (AESC) on May 25, 1979, from the Minister of International Trade and Industry. The investigation and discussion were commenced by the AESC on May 28, 1979. The modification of the reactor installation is the increase of spent fuel storage capacity from about 220% of in-core fuel at present to about 325%. The fundamental philosophy of the safety evaluation includes the following items; 1) the storage capacity of spent fuel is adequate, 2) the design is such that the criticality is prevented under any assumed condition, 3) the sufficient cooling capacity is kept for decay heat removal, 4) and others required for the safety. The contents of the safety examination for each philosophical item written above are presented. The increased spent fuel storage capacity is equivalent to the quantity produced in about eight years. The prevention of criticality in the spent fuel storage is carefully practiced by putting fuel assemblies in the stainless steel racks with large neutron absorption cross section and locating spent fuel assemblies at the proper distances. The effective multiplication factor is less than 0.95 at the most severe arrangement in the fuel pool. The water temperature in the pool is less than 65 deg C at about 325% core storage by operating the spent fuel pool water cooling system. The spent fuel storage racks are designed as the A class aseismatic structure. (Nakai, Y.)

  19. Advanced nuclear reactors and their simulators

    International Nuclear Information System (INIS)

    Chaushevski, Anton; Boshevski, Tome

    2003-01-01

    Population growth, economy development and improvement life standard impact on continually energy needs as well as electricity. Fossil fuels have limited reserves, instability market prices and destroying environmental impacts. The hydro energy capacities highly depend on geographic and climate conditions. The nuclear fission is significant factor for covering electricity needs in this century. Reasonable capital costs, low fuel and operating expenses, environmental acceptable are some of the facts that makes the nuclear energy an attractive option especially for the developing countries. The simulators for nuclear reactors are an additional software tool in order to understand, study research and analyze the processes in nuclear reactors. (Original)

  20. Superconducting magnets for model ship propulsion and for material tests of a nuclear fusion reactor

    International Nuclear Information System (INIS)

    Horiuchi, T.; Matsumoto, K.; Monju, Y.; Tatara, I.; Hamada, M.

    1982-01-01

    Nuclear fusion reactors, magnetically levitated trains, and MHD generators, etc., all need a very high magnetic field; which in order to be attained a means the application of superconductors is inevitable. This paper describes the development of ''CRYOZITT'', a superconductor featuring high current density and high mechanical strength. CRYOZITT has already been used in the manufacture of two race-track shaped superconducting magnets, and delivered to highly satisfied customers. (author)

  1. 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.

  2. Nuclear Power in Countries with Limited Electrical Grid Capacities: The Case of Armenia. A Report of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO)

    International Nuclear Information System (INIS)

    2015-12-01

    This publication addresses issues relating to nuclear power deployment faced by countries with electrical grids of limited capacity and stability. In particular, technology issues and related institutional measures as well as some technical and economic options for managing spent fuel and radioactive waste applicable in these circumstances are addressed. It aims to assist States implementing a nuclear power programme in the development of a comprehensive approach to the long term management of spent nuclear fuel and radioactive waste that is technically sound, environmentally responsible, economically feasible and acceptable to all stakeholders. Armenia was selected as a case study and the data obtained from the studies performed led to general recommendations which could be applicable to some other countries with similar economies and grid characteristics

  3. Nuclear power plant with boiling water reactor VK-300 for district heating and electricity supply

    International Nuclear Information System (INIS)

    Kuznetsov, Y.N.; Lisitza, F.D.; Romenkov, A.A.; Tokarev, Y.I.

    1998-01-01

    The paper considers specific design features of a pressure vessel boiling water reactor with coolant natural circulation and three-step in-vessel steam separation (at draught tube outlet of the upcomer, within zone of overflow from the upcomer to downcomer and in cyclon-type separators). Design description and analytical study results are presented for the passive core cooling system in the case of loss of preferred power and rupture in primary circuit pipeline. Specific features of a primary containment (safeguard vessel) are given for an underground NPP sited in a rock ground. (author)

  4. Fuzzy controller for real time supervision of nuclear power reactor

    International Nuclear Information System (INIS)

    Bala Subramanian, R.

    2012-01-01

    Generally nuclear energy provides about 60% of the whole electricity production. A modulation of the nuclear power plants must be able to respond to the demand on the network. The pressurized water nuclear reactor has to yield correctly a load set point. Fundamentally, two parameters are concerned in leading this task to a successful conclusion: the power axial-offset and the control rods position. The focus of this study is the automation of the control of the power axial-offset by adding soluble boron and by minimizing the volume flows through the water pump. It is also important to take into consideration the liquid waste volume. Water or boron is injected into the reactor primary circuit. At the present time this task is still performed manually by an operator, for all previous attempts to automate it failed. That device, sketchily described in the paper, gave rise to the development of a real-time fuzzy controller for the power axial-offset and the control rods insertion in a pressurized water reactor (PWR). The fuzzy controller, which is the main subject of the paper, expresses more naturally the human expertise, thus avoiding the previous issue of empirical tunings. It was implemented in simulation using Matlab-Simulink on a Sun workstation. Two realistic tests discussed show that the fuzzy controller runs as efficiently as an expert operator does

  5. Nuclear power reactor safety

    International Nuclear Information System (INIS)

    Pon, G.A.

    1976-10-01

    This report is based on the Atomic Energy of Canada Limited submission to the Royal Commission on Electric Power Planning on the safety of CANDU reactors. It discusses normal operating conditions, postulated accident conditions, and safety systems. The release of radioactivity under normal and accident conditions is compared to the limits set by the Atomic Energy Control Regulations. (author)

  6. The roles played by the Canadian General Electric Company's Atomic Power Department in Canada's nuclear power program: work, organization and success in APD, 1955-1995

    International Nuclear Information System (INIS)

    Cantello, G.W.

    2003-01-01

    This thesis explores the roles played by the Canadian General Electric Company's Atomic Power Department (APD) in Canada's distinctive nuclear power program. From the establishment of APD in 1955 until the completion of the KANUPP project in Pakistan in 1972, the company's strategy encompassed the design, manufacture, and commissioning of entire nuclear power projects in Canada and abroad. APD then developed a specialized role in the design and supply of complete nuclear fuel handling systems, nuclear fuel bundles, and service work, that sustained a thriving workplace. Five key factors are identified as the reasons behind the long and successful history of the department: (1) Strong, capable and efficient management from the start, (2) Flexible organizational structure, (3) Extremely competent design group, (4) Excellent manufacturing, test, commissioning and service capabilities, (5) Correctly identifying, at the right time, the best fields in which to specialize. (author)

  7. 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

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

    International Nuclear Information System (INIS)

    Ishay, Liel; Bieder, Ulrich; Ziskind, Gennady; Rashkovan, Alex

    2015-01-01

    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

  9. Nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    Sasaki, Y.; Tashima, J.

    1975-01-01

    A description is given of nuclear reactor fuel assemblies arranged in the form of a lattice wherein there is attached to the interface of one of two adjacent fuel assemblies a plate spring having a concave portion curved toward said interface and to the interface of the other fuel assembly a plate spring having a convex portion curved away from said interface

  10. Forced vibration tests and simulation analyses of a nuclear reactor building. Part 2: simulation analyses

    International Nuclear Information System (INIS)

    Kuno, M.; Nakagawa, S.; Momma, T.; Naito, Y.; Niwa, M.; Motohashi, S.

    1995-01-01

    Forced vibration tests of a BWR-type reactor building. Hamaoka Unit 4, were performed. Valuable data on the dynamic characteristics of the soil-structure interaction system were obtained through the tests. Simulation analyses of the fundamental dynamic characteristics of the soil-structure system were conducted, using a basic lumped mass soil-structure model (lattice model), and strong correlation with the measured data was obtained. Furthermore, detailed simulation models were employed to investigate the effects of simultaneously induced vertical response and response of the adjacent turbine building on the lateral response of the reactor building. (author). 4 refs., 11 figs

  11. TIBER II/ETR [Engineering Test Reactor] nuclear shielding and optional tritium breeding system: An overview

    International Nuclear Information System (INIS)

    Lee, J.D.; Sawan, M.

    1987-01-01

    TIBER II, the Tokamak Ignition/Burn Experimental Reactor II, is a design concept developed as the US candidate for an International Engineering Test Reactor (ETR). An important objective of this design is to minimize cost by minimizing major radius while providing a wall loading greater than 1.0 MW/m2 and a total fluence greater than 3.0 MWY/m2 needed for blanket module testing. The shielding required for the superconducting TF coils is an important element in setting TIBER II's 3.0m major radius. 6 refs., 1 fig., 1 tab

  12. Materials for nuclear reactors

    International Nuclear Information System (INIS)

    Banerjee, S.; Kamath, H.S.

    2005-01-01

    The improved performance of present generation nuclear reactors and the realization of advanced reactor concepts, both, require development of better materials. Physical metallurgy/materials science principles which have been exploited in meeting the exacting requirements of nuclear reactor materials (fuels and structural materials), are outlined citing a few specific examples. While the incentive for improvement of traditional fuels (e.g., UO 2 fuel) is primarily for increasing the average core burn up, the development of advanced fuels (e.g., MOX, mixed carbide, nitride, silicide and dispersion fuels) are directed towards better utilization of fissile and fertile inventories through adaptation of innovative fuel cycles. As the burn up of UO 2 fuel reaches higher levels, a more detailed and quantitative understanding of the phenomena such as fission gas release, fuel restructuring induced by radiation and thermal gradients and pellet-clad interaction is being achieved. Development of zirconium based alloys for both cladding and pressure tube applications is discussed with reference to their physical metallurgy, fabrication techniques and in-reactor degradation mechanisms. The issue of radiation embrittlement of reactor pressure vessels (RPVs) is covered drawing a comparison between the western and eastern specifications of RPV steels. The search for new materials which can stand higher rates of atomic displacement due to radiation has led to the development of swelling resistant austenitic and ferritic stainless steels for fast reactor applications as exemplified by the development of the D-9 steel for Indian fast breeder reactor. The presentation will conclude by listing various materials related phenomena, which have a strong bearing on the successful development of future nuclear energy systems. (author)

  13. Nuclear reactor assembly

    International Nuclear Information System (INIS)

    Dorner, H.; Scholz, M.; Jungmann, A.

    1975-01-01

    A nuclear reactor assembly includes a reactor pressure tank having a substantially cylindrical side wall surrounded by the wall of a cylindrical cavity formed by a biological shield. A rotative cylindrical wall is interposed between the walls and has means for rotating it from outside of the shield, and a probe is carried by the rotative wall for monitoring the pressure tank's wall. The probe is vertically movable relative to the rotative cylindrical wall, so that by the probe's vertical movement and rotation of the rotative cylinder, the reactor's wall can be very extensively monitored. If the reactor pressure tank's wall fails, it is contained by the rotative wall which is backed-up by the shield cavity wall. (Official Gazette)

  14. 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).

  15. Environmental Assessment for Authorizing the Puerto Rico Electric Power Authority (PREPA) to allow Public Access to the Boiling Nuclear Superheat (BONUS) Reactor Building, Rincon, Puerto Rico

    International Nuclear Information System (INIS)

    2003-01-01

    The U.S. Department of Energy (DOE) proposes to consent to a proposal by the Puerto Rico Electric Power Authority (PREPA) to allow public access to the Boiling Nuclear Superheat (BONUS) reactor building located near Rincon, Puerto Rico for use as a museum. PREPA, the owner of the BONUS facility, has determined that the historical significance of this facility, as one of only two reactors of this design ever constructed in the world, warrants preservation in a museum, and that this museum would provide economic benefits to the local community through increased tourism. Therefore, PREPA is proposing development of the BONUS facility as a museum

  16. CANDU nuclear reactor technology

    International Nuclear Information System (INIS)

    Kakaria, B. K.

    1994-01-01

    AECL has over 40 years of experience in the nuclear field. Over the past 20 years, this unique Canadian nuclear technology has made a worldwide presence, In addition to 22 CANDU reactors in Canada, there are also two in India, one in Pakistan, one in Argentina, four in Korea and five in Romania. CANDU advancements are based on evolutionary plant improvements. They consist of system performance improvements, design technology improvements and research and development in support of advanced nuclear power. Given the good performance of CANOU plants, it is important that this CANDU operating experience be incorporated into new and repeat designs

  17. Moderator for nuclear reactor

    International Nuclear Information System (INIS)

    Milgram, M.S.; Dunn, J.T.; Hart, R.S.

    1995-01-01

    This invention relates to a moderator for a nuclear reactor and more specifically, to a composite moderator. A moderator is designed to slow down, or thermalize, neutrons which are released during nuclear reactions in the reactor fuel. Pure or almost pure materials like light water, heavy water, beryllium or graphite are used singly as moderators at present. All these materials, are used widely. Graphite has a good mechanical strength at high temperatures encountered in the nuclear core and therefore is used as both the moderator and core structural material. It also exhibits a low neutron-capture cross section and high neutron scattering cross section. However, graphite is susceptible to attach by carbon dioxide and/or oxygen where applicable, and releases stress energy under certain circumstances, although under normal operating conditions these reactions can be controlled. (author). 1 tab

  18. Power Nuclear Reactors: technology and innovation for development in future

    International Nuclear Information System (INIS)

    Suarez Antola, R.

    2009-01-01

    The conference is about some historicals task of the fission technology as well as many types of Nuclear Reactors. Enrichment of fuel, wastes, research reactors and power reactors, a brief advertisment about Uruguay electric siystem and power generation, energetic worldwide, proliferation, safety reactors, incidents, accidents, Three-Mile Island accident, Chernobil accident, damages, risks, classification and description of Power reactors steam generation, nuclear reactor cooling systems, future view

  19. Test facility for auxiliary cooling system (ACS) of fast breeder reactor for Power Reactor and Nuclear Fuel Development Corporation (PNC)

    International Nuclear Information System (INIS)

    1983-01-01

    In preparation of constructing ''Monju'', a prototype fast breeder reactor, PNC has been pushing forward its research and development projects and the ACS was constructed under these projects. The auxiliary cooling system is an important engineered safety feature, and is used for safe removal of heat from the reactor at the shutdown. The ACS serves as a means of testing and assessing the auxiliary cooling system for the ''Monju'' and is designed and manufactured to have one fifth capacity of the Monju. The air heat exchanger and the ACS system was designed to withstand higher temperature range of the conventional design code (MITI-501), and finned tubes were applied for effective heat removal. Preheating system was designed to heat up the whole system over 200 0 C within 20 hours to prevent sodium from freezing. Basic performance of ACS was verified satisfactorily by a series of performance tests, such as start up test, flow rate measurement and preheating test before delivery. The experience from designing and construction of ACS and data obtained by these tests will be very instructive for designing and construction of the ''Monju''. (author)

  20. Nuclear reactor instrumentation at research reactor renewal

    International Nuclear Information System (INIS)

    Baers, B.; Pellionisz, P.

    1981-10-01

    The paper overviews the state-of-the-art of research reactor renewals. As a case study the instrumentation reconstruction of the Finnish 250 kW TRIGA reactor is described, with particular emphasis on the nuclear control instrumentation and equipment which has been developed and manufactured by the Central Research Institute for Physics, Budapest. Beside the presentation of the nuclear instrument family developed primarily for research reactor reconstructions, the quality assurance policy conducted during the manufacturing process is also discussed. (author)

  1. TU Electric reactor physics model verification: Power reactor benchmark

    International Nuclear Information System (INIS)

    Willingham, C.E.; Killgore, M.R.

    1988-01-01

    Power reactor benchmark calculations using the advanced code package CASMO-3/SIMULATE-3 have been performed for six cycles of Prairie Island Unit 1. The reload fuel designs for the selected cycles included gadolinia as a burnable absorber, natural uranium axial blankets and increased water-to-fuel ratio. The calculated results for both startup reactor physics tests (boron endpoints, control rod worths, and isothermal temperature coefficients) and full power depletion results were compared to measured plant data. These comparisons show that the TU Electric reactor physics models accurately predict important measured parameters for power reactors

  2. 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

  3. Obtaining 64Cu in a nuclear reactor from a Zn matrix: Preliminary tests

    International Nuclear Information System (INIS)

    Aguirre, Andrea; Bedregal, Patricia; Montoya, Eduardo; Cohen, Marcos Isaac

    2014-01-01

    The design and feasibility of a method for obtaining 6C u in a nuclear reactor, from the 64 Zn(n,p) 64 Cu threshold reaction of zinc, induced by the fast component of the neutron spectrum, is presented. The product obtained will be used in positron emission tomography (PET). The preliminary experiments were performed using the RP-10 research reactor at a power of 3.5 MW, followed by a radiochemical separation by solvent extraction using a chloroform solution of dithizone. The radioisotope has been identified and quantified through the full energy peak of 1345.77 keV, using a high resolution gamma spectrometry system. The preliminary yield achieved demonstrates the feasibility of the proposed method. (authors).

  4. Characteristics of potential repository wastes: Volume 4, Appendix 4A, Nuclear reactors at educational institutions of the United States; Appendix 4B, Data sheets for nuclear reactors at educational institutions; Appendix 4C, Supplemental data for Fort St. Vrain spent fuel; Appendix 4D, Supplemental data for Peach Bottom 1 spent fuel; Appendix 4E, Supplemental data for Fast Flux Test Facility

    International Nuclear Information System (INIS)

    1992-07-01

    Volume 4 contains the following appendices: nuclear reactors at educational institutions in the United States; data sheets for nuclear reactors at educational institutions in the United States(operational reactors and shut-down reactors); supplemental data for Fort St. Vrain spent fuel; supplemental data for Peach Bottom 1 spent fuel; and supplemental data for Fast Flux Test Facility

  5. An overview of future sustainable nuclear power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Poullikkas, Andreas [Electricity Authority of Cyprus, P.O. Box 24506, 1399 Nicosia (Cyprus)

    2013-07-01

    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 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 tend to have closed

  6. Conceptual design of an electrical power module for the tokamak fusion test reactor

    International Nuclear Information System (INIS)

    Jassby, D.L.; Bullis, R.; Sedgeley, D.; Caldwell, C.S.; Pettus, W.G.; Schluderberg, D.C.

    1979-01-01

    The TFTR Engineering Test Station (ETS) can support blanket modules with a fusion-neutron view area of 0.5 m/sup 2/. If the TFTR magnetic systems and beam injectors can operate with pulse lengths of 5 s, once every 300 s, the time-averaged neutron power incident on a module will be 1.5 kW, which can be enhanced by a suitable blanket energy multiplier. A preliminary conceptual design of a dual-loop steam-generating power system that can be housed in the ETS has been carried out. The optimal heat transfer fluid in the primary loop is an organic liquid, which allows an operating temperature of 700/degree/F at low pressure. The primary coolant must be preheated electrically to operating temperature. A ballast tank levels the temperature at the steam generator, so that the secondary loop is in steady-state operation. With a natural-uranium blanket multiplier, the time-averaged net electrical power is 1.2 kW(e). 8 refs

  7. Verifying seismic design of nuclear reactors by testing. Volume 2: appendix, theoretical discussions

    International Nuclear Information System (INIS)

    1979-01-01

    Theoretical discussions on seismic design testing are presented under the following appendix headings: system functions, pulse optimization program, system identification, and motion response calculations from inertance measurements of a nuclear power plant

  8. 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.

  9. Nuclear reactor constructions

    International Nuclear Information System (INIS)

    Aspden, G.J.

    1980-01-01

    A nuclear reactor construction comprising a reactor core submerged in a pool of liquid metal coolant in a primary vessel which is suspended from the roof structure of a containment vault. Control rods supported from the roof structure are insertable in the core which is carried on a support structure from the wall of the primary vessel. To prevent excessive relaxation of the support structure whereby the control rods would be displaced relative to the core, the support structure incorporates a normally inactive secondary structure designed to become effective in bracing the primary structure against further relaxation beyond a predetermined limit. (author)

  10. Nuclear reactor installation

    International Nuclear Information System (INIS)

    Keller, W.

    1976-01-01

    A nuclear reactor installation includes a pressurized-water coolant reactor vessel and a concrete biological shield surrounding this vessel. The shield forms a space between it and the vessel large enough to permit rapid escape of the pressurized-water coolant therefrom in the event the vessel ruptures. Struts extend radially between the vessel and shield for a distance permitting normal radial thermal movement of the vessel, while containing the vessel in the event it ruptures, the struts being interspaced from each other to permit rapid escape of the pressurized-water coolant from the space between the shield and the vessel

  11. Australia's new nuclear reactor

    International Nuclear Information System (INIS)

    Kemeny, L.

    2007-01-01

    On 19 and 20 April 2007, the Australian Nuclear Science and Technology Organisation (ANSTO) celebrated the recent commissioning of its new, world-class, OPAL (Open Pool Australian Lightwater) research reactor at the Lucas Heights. On the 19th, scientists, business leaders and academics were introduced to the reactor and its technical capacity for the manufacture of radiopharmaceuticals, its material science applications, its environmental services and its neutron scattering facilities for business applications. The formal OPAL opening function took place that evening and, on the 20th, Prime Minister John Howard visited ANSTO to be briefed about OPAL and to be shown the work being carried out at Lucas Heights

  12. Refueling of nuclear reactor

    International Nuclear Information System (INIS)

    Kaufmann, J.W.; Swidwa, K.J.; Hornak, L.P.

    1989-01-01

    This patent describes an apparatus for refueling a nuclear reactor, the reactor being disposed for refueling under water in a pit in a containment, the apparatus including a bridge to be mounted moveably over the pit on the containment, first means connected to the bridge, for moving the bridge forward and backward on the containment over the pit along a first path, a first pulse generator, connected to the moving means, responsive to the movement of the bridge, for producing pulses, means, connected to the generator,for counting the pulses, the count of the pulses being dependent on the distance of the movement of the bridge

  13. Equipment for testing a group of nuclear reactor fuel elements for damage to the cans

    International Nuclear Information System (INIS)

    Mohm, F.

    1977-01-01

    Equipment is described for use in sodium cooled nuclear reactors, with which the fuel elements consisting of bundles of fuel and fertile rods can be examined for damage to the cans. Fission poducts occurring in the liquid coolant act as indicators. The coolant is sucked via pipelines which penetrate into the elements into a collecting container, and a special pipeline is available for every element of a group, where the highest points of individual pipelines at different hydrostatic heads are taken to the collecting container. This permits the checking of one line at a time due to pressure changes. (UWI) [de

  14. 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.

  15. 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.

  16. The Birth of Nuclear-Generated Electricity

    International Nuclear Information System (INIS)

    Claflin, D.J. POC

    1999-01-01

    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

  17. Technical evaluation report on the monitoring of electric power to the reactor protection system for the Nine Mile Point Nuclear Station, Unit 1 (Docket No. 50-220)

    International Nuclear Information System (INIS)

    Selan, J.C.

    1984-01-01

    This report documents the technical evaluation of the monitoring of electric power to the reactor protection system (RPS) at the Nine Mile Point Nuclear Station, Unit 1. The evaluation is to determine if the proposed design modification will protect the RPS from abnormal voltage and frequency conditions which could be supplied from the power supplies and will meet certain requirements set forth by the Nuclear Regulatory Commission. The proposed design modifications will protect the RPS from sustained abnormal voltage and frequency conditions from the supplying sources

  18. Sextant: an expert system for transient analysis of nuclear reactors and integral test facilities

    International Nuclear Information System (INIS)

    Barbet, N.; Dumas, M.; Mihelich, G.

    1987-01-01

    Expert systems provide a new way of dealing with the computer-aided management of nuclear plants by combining several knowledge bases and reasoning modes together with a set of numerical models for real-time analysis of transients. New development tools are required together with metaknowledge bases handling temporal hypothetical reasoning and planning. They have to be efficient and robust because during a transient, neither measurements nor models, nor scenarios are hold as absolute references. SEXTANT is a general purpose physical analyzer intended to provide a pattern and avoid duplication of general tools and knowledge bases for similar applications. It combines several knowledge bases concerning measurements, models and qualitative behavior of PWR with a mechanism of conjecture-refutation and a set of simplified models matching the current physical state. A prototype is under assessment by dealing with integral test facility transients. For its development, SEXTANT requires a powerful shell. SPIRAL is such a toolkit, oriented towards online analysis of complex processes and already used in several applications

  19. Licensing of nuclear reactor operators

    International Nuclear Information System (INIS)

    1979-09-01

    Recommendations are presented for the licensing of nuclear reactor operators in units licensed according to the legislation in effect. They apply to all physical persons designated by the Operating Organization of the nuclear reactor or reactors to execute any of the following functional activities: a) to manipulate the controls of a definite reactor b) to direct the authorized activities of the reactor operators licesed according to the present recommendations. (F.E.) [pt

  20. Nuclear electricity. 5. ed.

    International Nuclear Information System (INIS)

    Hore-Lacy, I.

    1999-01-01

    This new edition is updated, expanded and in a larger format than its predecessors. Also it is a joint undertaking with Canada, rather than simply an Australian perspective. It has a much expanded chapter on nuclear wastes and reprocessing, as well as more on the advanced reactors which are now coming into service. It also discusses nuclear reactor safety and safeguards issues. A final chapter looks at Australian and Canadian uranium mining

  1. Nuclear reactor containment device

    International Nuclear Information System (INIS)

    Ichiki, Tadaharu.

    1980-01-01

    Purpose: To reduce the volume of a containment shell and decrease the size of a containment equipment for BWR type reactors by connecting the containment shell and a suppression pool with slanted vent tubes to thereby shorten the vent tubes. Constitution: A pressure vessel containing a reactor core is installed at the center of a building and a containment vessel for the nuclear reactor that contains the pressure vessel forms a cabin. To a building situated below the containment shell, is provided a suppression chamber in which cooling water is charged to form a suppression pool. The suppression pool is communicated with vent tubes that pass through the partition wall of the containment vessel. The vent tubes are slanted and their lower openings are immersed in coolants. Therefore, if accident is resulted and fluid at high temperature and high pressure is jetted from the pressure vessel, the jetting fluid is injected and condensated in the cooling water. (Moriyama, K.)

  2. Nuclear reactor container

    International Nuclear Information System (INIS)

    Ishiyama, Takenori.

    1989-01-01

    This invention concerns a nuclear reactor container in which heat is removed from a container by external water injection. Heat is removed from the container by immersing the lower portion of the container into water and scattering spary water from above. Thus, the container can be cooled by the spray water falling down along the outer wall of the container to condensate and cool vapors filled in the container upon occurrence of accidents. Further, since the inside of the container can be cooled also during usual operation, it can also serve as a dry well cooler. Accordingly, heat is removed from the reactor container upon occurrence of accidents by the automatic operation of a spray device corresponding to the change of the internal temperature and the pressure in the reactor container. Further, since all of these devices are disposed out of container, maintenance is also facilitated. (I.S.)

  3. The nuclear reactor systems

    International Nuclear Information System (INIS)

    Bacher, P.

    2008-01-01

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

  4. Nuclear reactor refueling system

    International Nuclear Information System (INIS)

    Wade, E.E.

    1978-01-01

    A system for transferring fuel assemblies between a nuclear reactor core and a fuel storage area while the fuel assembies remain completely submerged in a continuous body of coolant is described. The system comprises an in-vessel fuel transfer machine located inside the reactor vessel and an ex-vessel fuel transfer machine located in a fuel storage tank. The in-vessel fuel transfer machine comprises two independently rotatable frames with a pivotable fuel transfer apparatus disposed on the lower rotatable frame. The ex-vessel fuel transfer machine comprises one frame with a pivotable fuel transfer apparatus disposed thereon. The pivotable apparatuses are capable of being aligned with each other to transfer a fuel assembly between the reactor vessel and fuel storage tank while the fuel assembly remains completely submerged in a continuous body of coolant. 9 claims, 7 figures

  5. Heat dissipating nuclear reactor

    Science.gov (United States)

    Hunsbedt, A.; Lazarus, J.D.

    1985-11-21

    Disclosed is a nuclear reactor containment adapted to retain and cool core debris in the unlikely event of a core meltdown and subsequent breach in the reactor vessel. The reactor vessel is seated in a cavity which has a thick metal sidewall that is integral with a thick metal basemat at the bottom of the cavity. The basemat extends beyond the perimeter of the cavity sidewall. Underneath the basemat is a porous bed with water pipes and steam pipes running into it. Water is introduced into the bed and converted into steam which is vented to the atmosphere. A plurality of metal pilings in the form of H-beams extend from the metal base plate downwardly and outwardly into the earth.

  6. Treatment of opinions, etc. in the public hearing on the alteration of reactor installation (addition of Unit 2) in the Shimane Nuclear Power Station of The Chugoku Electric Power Company, Inc

    International Nuclear Information System (INIS)

    1983-01-01

    The Nuclear Safety Commission has acknowledged the governmental policy, and further decided on the treatment of the opinions expressed by the local people in the public hearing held in May, 1983, in Shimane Prefecture on the addition of Unit 2 to the Shimane Nuclear Power Station, Chugoku Electric Power Co., Inc. The NSC has directed the Committee on Examination of Reactor Safety to take into consideration the opinions in its later examination. The opinions expressed by the local people in the form of question are given as follows: siting conditions (earthquake, ground, weather, etc.), the safety design for reactor installation (general aspect, aseismatic design, core design, ECCS, the teaching of TMI accident, etc.), radioactive wastes, radiation exposure, site evaluation. (Mori, K.)

  7. Nuclear reactor operator licensing

    International Nuclear Information System (INIS)

    Bursey, R.J.

    1978-01-01

    The Atomic Energy Act of 1954, which was amended in 1974 by the Energy Reorganization Act, established the requirement that individuals who had the responsibility of operating the reactors in nuclear power plants must be licensed. Section 107 of the act states ''the Commission shall (1) prescribe uniform conditions for licensing individuals; (2) determine the qualifications of such individuals; and (3) issue licenses to such individuals in such form as the Commission may prescribe.'' The article discusses the types of licenses, the selection and training of individuals, and the administration of the Nuclear Regulatory Commission licensing examinations

  8. Development and testing of nuclear graphite for the German pebble-bed high temperature reactor

    International Nuclear Information System (INIS)

    Haag, G.; Delle, W.; Nickel, H.; Theymann, W.; Wilhelmi, G.

    1987-01-01

    Several types of high temperature reactors have been developed in the Federal Republic of Germany. They are all based on spherical fuel elements being surrounded by graphite as reflector material. As an example, HTR-500 developed by the Hochtemperatur Reaktorbau GmbH is shown. The core consists of the top reflector, the side reflector with inner and outer parts, the bottom reflector and the core support columns. The most serious problem with respect to fast neutron radiation damage had to be solved for the materials of those parts near the pebble bed. Regarding the temperature profile in the core, the top reflector is at 300 deg C, and as cooling gas flows from the top downward, the temperature of the inner side reflector rises to about 700 deg C at the bottom. Fortunately, the highest fast neutron load accumulated during the life time of a reactor corresponds to the lowest temperature. This makes graphite components easier to survive neutron exposure without being mechanically damaged, although the maximum fast neutron fluence is as high as 4 x 10 22 /cm 2 at about 400 deg C. HTR graphite components are divided into four classes according to loading. The raw materials for nuclear graphite, the development of pitch coke nuclear graphite, the irradiation behavior of ATR-2E and ASR-IRS and others are reported. (Kako, I.)

  9. Nuclear power reactor physics

    International Nuclear Information System (INIS)

    Barjon, Robert

    1975-01-01

    The purpose of this book is to explain the physical working conditions of nuclear reactors for the benefit of non-specialized engineers and engineering students. One of the leading ideas of this course is to distinguish between two fundamentally different concepts: - a science which could be called neutrodynamics (as distinct from neutron physics which covers the knowledge of the neutron considered as an elementary particle and the study of its interactions with nuclei); the aim of this science is to study the interaction of the neutron gas with real material media; the introduction will however be restricted to its simplified expression, the theory and equation of diffusion; - a special application: reactor physics, which is introduced when the diffusing and absorbing material medium is also multiplying. For this reason the chapter on fission is used to introduce this section. In practice the section on reactor physics is much longer than that devoted to neutrodynamics and it is developed in what seemed to be the most relevant direction: nuclear power reactors. Every effort was made to meet the following three requirements: to define the physical bases of neutron interaction with different materials, to give a correct mathematical treatment within the limit of necessary simplifying hypotheses clearly explained; to propose, whenever possible, numerical applications in order to fix orders of magnitude [fr

  10. In France, 37 nuclear reactors out of 58 can be stopped right away tomorrow without electricity outage. How is it possible? Mathematical demonstration of an electric scenario 'right-away Tomorrow'

    International Nuclear Information System (INIS)

    Houpert, Sylvain

    2013-12-01

    According to the 'Des demain' (right-away tomorrow), this document indicates data related to nuclear electric power in France (production, export, self-consumption, loss due to Fessenheim shut down, to the production of other electricity production plants (from coal, gas, oil, biomass, wind, sun). From these data, it states that France could operate with 1 nuclear reactor out of 3 (whereas Japan has stopped 98 pc of its reactors after the Fukushima accident). This result and its possible consequences are then discussed, and the 'Des demain' scenario is then presented in terms of electric power production for the 20 years to come, CO 2 emissions for the next 20 years, energy transition and renewable energies in France in 2012

  11. Shields for nuclear reactors

    International Nuclear Information System (INIS)

    Aspden, G.J.

    1984-01-01

    The patent concerns shields for nuclear reactors. The roof shield comprises a normally fixed radial outer portion, a radial inner portion rotatable about a vertical axis, and a connection between the inner and outer portions. In the event of hypothecal core disruption conditions, a cantilever system on the inner wall allows the upward movement of the inner wall, in order to prevent loss of containment. (UK)

  12. Compact nuclear reactor

    International Nuclear Information System (INIS)

    Juric, S.I.

    1975-01-01

    A compact nuclear reactor of the pressurized-water variety is described which has two separate parts separably engageable for ease of inspection, maintenance and repair. One of the parts is a pressure vessel having an active core and the other of the parts is a closure adapted on its lower surface with an integral steam generator. An integral pump, external pressurizer and control rods are provided which communicate with the active core when engaged to form a total unit. (U.S.)

  13. Nuclear reactor instrumentation

    International Nuclear Information System (INIS)

    Duncombe, E.; McGonigal, G.

    1975-01-01

    A liquid metal cooled nuclear reactor is described which has an equal number of fuel sub-assemblies and sensing instruments. Each instrument senses temperature and rate of coolant flow of a coolant derived from a group of three sub-assemblies so that an abnormal value for one sub-assembly will be indicated on three instruments thereby providing for redundancy of up to two of the three instruments. The abnormal value may be a precurser to unstable boiling of coolant

  14. Directory of Nuclear Research Reactors 1994

    International Nuclear Information System (INIS)

    1995-08-01

    The Directory of Nuclear Research Reactors is an output of the Agency's computerized Research Reactor Data Base (RRDB). It contains administrative, technical and utilization information on research reactors known to the Agency at the end of December 1994. The data base converted from mainframe to PC is written in Clipper 5.0 and the publication generation system uses Excel 4. The information was collected by the Agency through questionnaires sent to research reactor owners. All data on research reactors, training reactors, test reactors, prototype reactors and critical assemblies are stored in the RRDB. This system contains all the information and data previously published in the Agency's publication, Directory of Nuclear Research Reactor, as well as updated information

  15. Subchannel analysis in nuclear reactors

    International Nuclear Information System (INIS)

    Ninokata, H.; Aritomi, M.

    1992-01-01

    This book contains 10 informative papers, presented at the International Seminar on Subchannel Analysis 1992 (ISSCA '92), organized by the Institute of Applied Energy, in collaboration with Atomic Energy Society of Japan, Tokyo Electric Power Company, Kansai Electric Power Company, Nuclear Power Engineering Corporation and the Japan Atomic Energy Research Institute, and held at the TIS-Green Forum, Tokyo, Japan, 30 October 1992. The seminar ISSCA '92 was intended to review the current state-of-the-arts of the method being applied to advanced nuclear reactors including Advanced BWRs, Advanced PWRs and LMRs, and to identify the problems to be solved, improvements to be made, and the needs of R and Ds that were required from the new fuel bundles design. The critical review was to focus on the performances of currently available subchannel analysis codes with regard to heat transfer and fluid flows in various types of nuclear reactor bundles under both steady-state and transient operating conditions, CHF, boiling transition (BT) or dryout behaviors and post BT. The behaviors of physical modeling and numerical methods in these extreme conditions were discussed and the methods critically evaluated in comparison with experiments. (author) (J.P.N.)

  16. Nozzle for electric dispersion reactor

    Science.gov (United States)

    Sisson, W.G.; Basaran, O.A.; Harris, M.T.

    1995-11-07

    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. 4 figs.

  17. The International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO): General description and implications for the research reactor infrastructure needed for R and D

    International Nuclear Information System (INIS)

    Sokolov, Yury A.

    2005-01-01

    The substantial growth in 21st century energy supplies needed to meet sustainable development goals has been emphasized by UNCSD, WSSD, IPCC and others. This will be driven by continuing population growth, economic development and aspiration to provide access to modern energy systems to the 1,6 billion people now without such access, the growth demand on limiting greenhouse gas emissions, and reducing the risk of climate change. A key factor to the future of nuclear power is the degree to which innovative nuclear technologies can be developed to meet challenges of economic competitiveness, safety, waste and proliferation concerns. There are two major international initiatives in the area of innovative nuclear technology: the IAEA's International Project on Innovative Nuclear Reactors and Fuel Cycle (INPRO) and the Generation IV International Forum. With INPRO some scenarios of future energy needs were identified and the methodology for holistic assessment of the innovative nuclear energy systems (INS), which can be developed to meet these scenarios, was developed.. The current status of the INPRO project and details of the INPRO methodology will be reported. The research needs identified due to Agency's activities on innovative nuclear system development assume the use of research reactors. The areas crucial for the development of INS which critically dependent of the RR experiments and following requirements addressed to the RR will be discussed. These areas include the development of advanced fuel and core materials for proposed innovative power reactor concepts. (author)

  18. Selection of nuclear reactor coolant materials

    International Nuclear Information System (INIS)

    Shi Lisheng; Wang Bairong

    2012-01-01

    Nuclear material is nuclear material or materials used in nuclear industry, the general term, it is the material basis for the construction of nuclear power, but also a leader in nuclear energy development, the two interdependent and mutually reinforcing. At the same time, nuclear materials research, development and application of the depth and breadth of science and technology reflects a nation and the level of the nuclear power industry. Coolant also known as heat-carrier agent, is an important part of the heart nuclear reactor, its role is to secure as much as possible to the economic output in the form fission energy to heat the reactor to be used: the same time cooling the core, is controlled by the various structural components allowable temperature. This paper described the definition of nuclear reactor coolant and characteristics, and then addressed the requirements of the coolant material, and finally were introduced several useful properties of the coolant and chemical control. (authors)

  19. Nuclear reactor cavity streaming shield

    International Nuclear Information System (INIS)

    Klotz, R.J.; Stephen, D.W.

    1978-01-01

    The upper portion of a nuclear reactor vessel supported in a concrete reactor cavity has a structure mounted below the top of the vessel between the outer vessel wall and the reactor cavity wall which contains hydrogenous material which will attenuate radiation streaming upward between vessel and the reactor cavity wall while preventing pressure buildup during a loss of coolant accident

  20. 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)

  1. Nuclear research reactors in Brazil

    International Nuclear Information System (INIS)

    Cota, Anna Paula Leite; Mesquita, Amir Zacarias

    2011-01-01

    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)

  2. Management of Spent Nuclear Fuel from Nuclear Power Plant Reactor

    International Nuclear Information System (INIS)

    Wati, Nurokhim

    2008-01-01

    Management of spent nuclear fuel from Nuclear Power Plant (NPP) reactor had been studied to anticipate program of NPP operation in Indonesia. In this paper the quantity of generated spent nuclear fuel (SNF) is predicted based on the national electrical demand, power grade and type of reactor. Data was estimated using Pressurized Water Reactor (PWR) NPP type 1.000 MWe and the SNF management overview base on the experiences of some countries that have NPP. There are four strategy nuclear fuel cycle which can be developed i.e: direct disposal, reprocessing, DUPlC (Direct Use of Spent PWR Fuel In Candu) and wait and see. There are four alternative for SNF management i.e : storage at the reactor building (AR), away from reactor (AFR) using wet centralized storage, dry centralized storage AFR and prepare for reprocessing facility. For the Indonesian case, centralized facility of the wet type is recommended for PWR or BWR spent fuel. (author)

  3. Economic aspects of ecological risk due to nuclear and coal-fired electricity production (general comparison, related to the USSR)

    International Nuclear Information System (INIS)

    Novikov, V.; Wahlstroem, B.; Demin, V.; Lebedev, O.; Ignatiyev, V.

    1991-05-01

    The content of this paper is reflected in the chapter headings: (1) Introduction (2) Environmental problems relating to the coal-fired power plants (3) Costs of environmental protection for coal-fired power plants (4) Comparison of economic efficiency of nuclear and coal-fired power plants (5) Cost of environmental protection for normal operation of a nuclear power plant and its fuel facilities (6) Accidental risk from off-reactor nuclear fuel cycle facilities (7) Conclusion. (Quittner)

  4. Broad-Application Test Reactor

    International Nuclear Information System (INIS)

    Motloch, C.G.

    1992-05-01

    This report is about a new, safe, and operationally efficient DOE reactor of nuclear research and testing proposed for the early to mid- 21st Century. Dubbed the Broad-Application Test Reactor (BATR), the proposed facility incorporates a multiple-application, multiple-mission design to support DOE programs such as naval reactors and space power and propulsion, as well as research in medical, science, isotope, and electronics arenas. DOE research reactors are aging, and implementing major replacement projects requires long lead times. Primary design drivers include safety, low risk, minimum operation cost, mission flexibility, waste minimization, and long life. Scientists and engineers at the Idaho National Engineering Laboratory are evaluating possible fuel forms, structural materials, reactor geometries, coolants, and moderators

  5. Nuclear reactor sealing system

    International Nuclear Information System (INIS)

    McEdwards, J.A.

    1983-01-01

    A liquid metal-cooled nuclear reactor sealing system is disclosed. The nuclear reactor includes a vessel sealed at its upper end by a closure head. The closure head comprises at least two components, one of which is rotatable; and the two components define an annulus therebetween. The sealing system includes at least a first and second inflatable seal disposed in series in an upper portion of the annulus. The system further includes a dip seal extending into a body of insulation located adjacent a bottom portion of the closure head. The dip seal comprises a trough formed by a lower portion of one of the components, and a seal blade pendently supported from the other component and extending downwardly into the trough. A body of liquid metal is contained in the trough which submerges a portion of the seal blade. The seal blade is provided with at least one aperture located above the body of liquid metal for providing fluid communication between the annulus intermediate the dip seal and the inflatable seals, and a body of cover gas located inside the vessel. There also is provided means for introducing a purge gas into the annulus intermediate the inflatable seals and the seal blade. The purge gas is introduced in an amount sufficient to substantially reduce diffusion of radioactive cover gas or sodium vapor up to the inflatable seals. The purge gas mixes with the cover gas in the reactor vessel where it can be withdrawn from the vessel for treatment and recycle to the vessel

  6. Tests of experimental fuel elements by the method of nuclear-thermal pulse loadings in 'HYDRA' reactor

    International Nuclear Information System (INIS)

    Nastoyashchaya, O.V.; Lebedev, Yu. M.; Chechurov, A.M.; Khvostionov, Ye

    1997-01-01

    The results of tests of experimental fuel elements with uranium dioxide fuel composition embedded in Al and Zr matrix with the enrichment from 90% to 36% in respect to U-235 performed at the pulse 'HYDRA' reactor are presented in this paper. Testing is performed in the frame-work of extensive research program studying the behavior of fuel elements (FE) of research and mini nuclear power systems in case of practically immediate energy release in the fuel taking place during the RIA-type accidents. Duration of the neutron pulse when testing in 'HYDRA' reactor is from 7 to 20 ms. The methods of diagnostics of the state of FE prior to and after testing in the reactor are developed and verified. Mathematical model describing temperature fields inside the FE in the process of testing. and accounting for non-uniformity of fuel composition has been developed in order to summarize experimental results. Experimental data on the limiting values of the energy density leading to deformation and degradation of FE depending on the type of fuel composition have been obtained and the mechanisms for the development of these processes have been determined. The nature of physical-chemical processes taking place in the fuel composition and fuel cladding depending on material composition under different levels of energy deposition is demonstrated. The data on hydrogen generation and radioactive product release out of fuel after failure of FE are presented. (author)

  7. RA Research nuclear reactor Part 1, RA Reactor operation and maintenance in 1987

    International Nuclear Information System (INIS)

    Sotic, O.; Martinc, R.; Cupac, S.; Sulem, B.; Badrljica, R.; Majstorovic, D.; Sanovic, V.

    1987-01-01

    RA research reacto was not operated due to the prohibition issued in 1984 by the Government of Serbia. Three major tasks were finished in order to fulfill the licensing regulations about safety of nuclear facilities which is the condition for obtaining permanent operation licence. These projects involved construction of the emergency cooling system, reconstruction of the existing special ventilation system, and renewal of the system for electric power supply of the reactor systems. Renewal of the RA reactor instrumentation system was initiated. Design project was done by the Russian Atomenergoeksport, and is foreseen to be completed by the end of 1988. The RA reactor safety report was finished in 1987. This annual report includes 8 annexes concerning reactor operation, activities of services and financial issues, and three special annexes: report on testing the emergency cooling system, report on renewal of the RA reactor and design specifications for reactor renewal and reconstruction [sr

  8. The high flux reactor Petten, A multi-purpose research and test facility for the future of nuclear energy

    International Nuclear Information System (INIS)

    Bergmans, H.; Duijves, K.; Conrad, R.; Markgraf, J.F.W.; May, R.; Moss, R.L.; Sordon, G.; Tartaglia, G.P.

    1996-01-01

    The High Flux Reactor (HFR) at Petten, is owned by the European Commission (EC) and managed by the Institute for Advanced Materials (IAM) of the Joint Research Centre (JRC) of the EC. Its operation has been entrusted since 1962 to the Netherlands Energy Research Foundation (ECN). The HFR is one of the most powerful multi-purpose research and test reactors in the world. Together with the ECN hot cells at Petten, it has provided since three decades an integral and full complement of irradiation and examination services as required by current and future research and development for nuclear energy, industry and research organizations. Since 1963, the HFR has recognized record of consistent, reliable and high availability of more than 250 days of operation per year. The HFR has 20 in-core and 12 poolside irradiation positions, plus 12 beam tubes. With a variety of dedicated irradiation devices, and with its long-standing experience in executing small and large irradiation projects, the HFR is particularly suited for fuel, materials and components testing for all reactor lines, including thermonuclear fusion reactors. In addition, processing with neutrons and gamma rays, neutron-based research and inspection services are employed by industry and research, such as activation analysis, boron neutron capture therapy, neutron radiography and neutron diffraction. Moreover, in recent years, HFRs' mission has been broadened within the area of radioisotopes production, where, within a few years, the HFR has attained the European leadership in production volume

  9. Nuclear Reactor Sharing Program

    International Nuclear Information System (INIS)

    1994-01-01

    The Ohio State University Research Reactor (OSURR) is licensed to operate at a maximum power level of 500 kW. A pool-type reactor using flat-plate, low enriched fuel elements, the OSURR provides several experimental facilities including two 6-inch i.d. beam ports, a graphite thermal column, several graphite-isotope-irradiation elements, a pneumatic transfer system (Rabbit), various dry tubes, and a Central Irradiation Facility (CIF). The core arrangement and accessibility facilitates research programs involving material activation or core parameter studies. The OSURR control room is large enough to accommodate laboratory groups which can use control instrumentation for monitoring of experiments. The control instrumentation is relatively simple, without a large amount of duplication. This facilitates opportunities for hands-on experience in reactor operation by nuclear engineering students making reactor parameter measurements. For neutron activation analysis and analyses of natural environmental radioactivity, the NRL maintains the gamma ray spectroscopy system (GRSS). It is comprised of two PC-based 8192-channel multichannel analyzers (MCAs) with all the required software for quantitative analysis. A 3 double-prime x 3 double-prime NaI(Tl), a 14 percent Ge(Li), and a High Purity Germanium detector are currently available for use with the spectroscopy system

  10. Automated ultrasonic testing of nuclear reactor welds and overlays in pre-service and in-service inspections

    International Nuclear Information System (INIS)

    Sladky, J.

    1988-01-01

    Since 1982, automatic pre-service and in-service checks are being made of welded joints and overlays on pressure vessels of WWER-440 nuclear reactors in Czechoslovakia. This is being done using the SKODA REACTORTEST TRC facility which is used for checking peripheral welded joints on the pressure vessel, neck joints, overlays in other selected areas of the cylindrical section of the pressure vessel, on radius transitions of the pressure vessel and of necks, and on the cylindrical part of necks, and also for checking the base material in selected parts of the pressure vessel and the base material of the neck extension piece. The tests are of two types, namely tests of peripheral welds and overlays of the cylindrical parts of the pressure vessel, and tests of the necks. Different ultrasonic probe holders are used for the tests, with totally different design. Ultrasonic probes which were initially used were of foreign make while at present, those of Czechoslovak make are used. For each pressure vessel a set of ultrasonic probes is used which should suffice for the life of the vessel. Experience gained so far is being used in work on the project of a new device for testing nuclear reactor presure vessels from the inside. (Z.M.)

  11. Calorimeter measures high nuclear heating rates and their gradients across a reactor test hole

    Science.gov (United States)

    Burwell, D.; Coombe, J. R.; Mc Bride, J.

    1970-01-01

    Pedestal-type calorimeter measures gamma-ray heating rates from 0.5 to 7.0 watts per gram of aluminum. Nuclear heating rate is a function of cylinder temperature change, measured by four chromel-alumel thermocouples attached to the calorimeter, and known thermoconductivity of the tested material.

  12. Fast Reactors and Nuclear Nonproliferation

    International Nuclear Information System (INIS)

    Avrorina, E.N.; Chebeskovb, A.N.

    2013-01-01

    Conclusion remarks: 1. Fast reactor start-up with U-Pu fuel: – dependent on thermal reactors, – no needs in U enrichment, – needs in SNF reprocessing, – Pu is a little suitable for NED, – practically impossible gun-type NED, – difficulties for implosion-type NED: necessary tests, advanced technologies, etc. – Pu in blankets is similar to WPu by isotopic composition, – Use of blanket for production isotopes (e.g. 233 U), – Combined reprocessing of SNF: altogether blanket and core, – Blanket elimination: decrease in Pu production – No pure Pu separation. 2. Fast reactor start-up with U fuel: - Needs in both U enrichment and SNF reprocessing, - Independent of thermal reactors, - Good Pu bred in the core let alone blankets, - NED of simple gun-type design, - Increase of needs in SWU, - Increased demands in U supply. 3. Fast reactors for export: - Uranium shortage, - To replace thermal reactors in future, - No blankets (depends on the country, though), - Fuel supply and SNF take back, - International centers for rendering services of NFC. Time has come to remove from FRs and their NFC the label unfairly identifying them as the most dangerous installations of nuclear power from the standpoint of being a proliferation problem

  13. Reactor core of nuclear reactor

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  14. Determination of the neutron flux for the Yankee Rowe experiment in the Ford Nuclear Reactor

    International Nuclear Information System (INIS)

    Cacciapouti, R.J.; Petrusha, L.

    1994-01-01

    Yankee Atomic Electric Company undertook a Test Irradiation Program at the Ford Nuclear Reactor of the University of Michigan. The program was implemented to characterize the irradiation response of representative Yankee Rowe reactor vessel beltline plate materials. The program was also intended to remove uncertainties in the existing reactor vessel fluence and damage predictions on the Yankee Rowe reactor vessel steel. Since this is the first in-core experiment of this type for the Ford Nuclear Reactor, the measurement of the reaction rate and the estimate of the fluence are presented

  15. Seals in nuclear reactors

    International Nuclear Information System (INIS)

    1979-01-01

    The seals described are for use in a nuclear reactor where there are fuel assemblies in a vessel, an inlet and an outlet for circulating a coolant in heat transfer relationship with the fuel assemblies and a closure head on the vessel in a tight fluid relationship. The closure head comprises rotatable plugs which have mechanical seals disposed in the annulus around each plug while allowing free rotation of the plug when the seal is not actuated. The seal is usually an elastomer or copper. A means of actuating the seal is attached for drawing it vertically into the annulus for sealing. When the reactor coolant is liquid sodium, contact with oxygen must be avoided and argon cover gas fills the space between the bottom of the closure head and the coolant liquid level and the annuli in the closure head. (U.K.)

  16. Nuclear reactor container

    International Nuclear Information System (INIS)

    Yamaki, Rika; Kawabe, Ryuhei.

    1989-01-01

    A venturi scrubber is connected to a nuclear reactor container. Gases containing radioactive aerosols in the container are introduced into the venturi scrubber in the form of a high speed stream under the pressure of the container. The radioactive aerosols are captured by inertia collision due to the velocity difference between the high speed gas stream and water droplets. In the case of the present invention, since the high pressure of the reactor container generated upon accident is utilized, compressor, etc. is no more required, thereby enabling to reduce the size of the aerosol removing device. Further, since no external power is used, the radioactive aerosols can be removed with no starting failure upon accidents. (T.M.)

  17. Nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    1975-01-01

    A description is given of a nuclear reactor fuel assembly comprising a cluster of fuel elements supported by transversal grids so that their axes are parallel to and at a distance from each other, in order to establish interstices for the axial flow of a coolant. At least one of the interstices is occupied by an axial duct reserved for an auxiliary cooling fluid and is fitted with side holes through which the auxiliary cooling fluid is sprayed into the cluster. Deflectors extend as from a transversal grid in a position opposite the holes to deflect the cooling fluid jet towards those parts of the fuel elements that are not accessible to the auxiliary coolant. This assembly is intended for reactors cooled by light or heavy water [fr

  18. 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…

  19. Nuclear future: thinking for building. Proceedings of the 5. Brazilian national meeting on nuclear applications; 8. General congress on nuclear energy; 12. Brazilian national meeting on reactor physics and thermal hydraulics; Futuro nuclear: refletindo para construir. Anais do 5. Encontro nacional de aplicacoes nucleares; 8. Congresso geral de energia nuclear; 12. Encontro nacional de fisica de reatores e termo-hidraulica

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-07-01

    These proceedings, for the first time, present jointly the 12. Brazilian national meeting on reactor physics and thermal hydraulics (12. ENFIR), the 8. General congress on nuclear energy (8. CGEN), and the 5. Brazilian national meeting on nuclear applications (5. ENAN). The main theme of discussion was: 'Nuclear Future: thinking for building'. The papers have analysed the progresses of peaceful utilization of nuclear technology and its forecasting for the beginning of the new millennium. The construction of Angra-3 nuclear power plant have been discussed.

  20. Nuclear future: thinking for building. Proceedings of the 12. Brazilian national meeting on reactor physics and thermal hydraulics; 8. General congress on nuclear energy; 5. Brazilian national meeting on nuclear applications; Futuro nuclear: refletindo para construir. Anais do 12. Encontro nacional de fisica de reatores e termo-hidraulica; 8. Congresso geral de energia nuclear; 5. Encontro nacional de aplicacoes nucleares

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-07-01

    These proceedings, for the first time, present jointly the 12. Brazilian national meeting on reactor physics and thermal hydraulics (12 ENFIR), 8. General congress on nuclear energy (8. CGEN), and 5. Brazilian national meeting on nuclear applications (5. ENAN). The main theme of discussion was: 'Nuclear Future: thinking for building'. The papers have analysed the progresses of peaceful utilization of nuclear technology and its forecasting for the beginning of the new millennium. The construction of Angra-3 nuclear power plant have been discussed.

  1. Fast reactors in nuclear power

    Energy Technology Data Exchange (ETDEWEB)

    Kazachkovskii, O

    1981-02-01

    The possible applications are discussed of fast reactor nuclear power plants. Basic differences are explained in fast and thermal reactors, mainly with a view to nuclear fuel utilization. Discussed in more detail are the problems of nuclear fuel reproduction and the nost important technical problems of fast reactors. Flow charts are shown of heat transfer for fast reactors BN-350 (loop design) and BN-600 (integral coolant circuit design). Main specifications are given for demonstration and power fast reactors in operation, under construction and in project-stage.

  2. Modification of reactor installation in the Takahama nuclear power plants No.1 and No.2 of Kansai Electric Power Company, Inc

    International Nuclear Information System (INIS)

    1979-01-01

    The Nuclear Safety Commission recognized the adequacy concerning the inquiry which was offered from the Minister of International Trade and Industry on July 24, 1979, following the safety evaluation in the Ministry of International Trade and Industry, and decided to submit a report to the Minister of International Trade and Industry on July 26, 1979, about the modification of reactor installation in the Takahama nuclear power plants No. 1 and No. 2 of the Kansai Electric Power Company, Inc. This is concerned to the application which was made from the president of the Kansai Electric Power Company, Inc., to the Minister of International Trade and Industry on July 23, 1979. The content of the modification is to add a control circuit which is actuated by the signal of abnormal low pressure in a reactor to the circuit of actuating the emergency core cooling system of the plant. The influences on the safety protection system by the addition of the circuit for transmitting safety injection signal and by the additions of an interlock circuit preventing the misoperation of pressurizer spray and of a block circuit of safety injection signal in case of the abnormal low pressure in a reactor were evaluated. The effects on the function and characteristics of the emergency core cooling system due to the addition of the control circuit were investigated, and it was recognized by the analysis that there is no effect in the pipe ruptures of both small and large scales. (Nakai, Y.)

  3. Advanced nuclear reactor and nuclear fusion power generation

    International Nuclear Information System (INIS)

    2000-04-01

    This book comprised of two issues. The first one is a advanced nuclear reactor which describes nuclear fuel cycle and advanced nuclear reactor like liquid-metal reactor, advanced converter, HTR and extra advanced nuclear reactors. The second one is nuclear fusion for generation energy, which explains practical conditions for nuclear fusion, principle of multiple magnetic field, current situation of research on nuclear fusion, conception for nuclear fusion reactor and economics on nuclear fusion reactor.

  4. Nuclear reactor plant

    International Nuclear Information System (INIS)

    Schabert, H.P.; Laurer, E.

    1977-01-01

    The invention is concerned with a quick-closing valve on the main-steam pipe of a nuclear reactor plant. The quick-closing valve serves as isolating valve and as safety valve permitting depressurization in case of an accident. For normal operation a tube-shaped gate valve is provided as valve disc, enclosing an auxiliary valve disc to be used in case of accidents and which is opened at increased pressure to provide a smaller flow cross-section. The design features are described in detail. (RW) [de

  5. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    Hindle, E. D.

    1984-01-01

    An array of rods is assembled to form a fuel element for a pressurized water reactor, the rods comprising zirconium alloy sheathed nuclear fuel pellets and containing helium. The helium gas pressure is selected for each rod so that it differs substantially from the helium gas pressure in its closest neighbors. In a preferred arrangement the rods are arranged in a square lattice and the helium gas pressure alternates between a relatively high value and a relatively low value so that each rod has as its closest neighbors up to four rods containing helium gas at the other pressure value

  6. Nuclear reactor fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Hindle, E. D.

    1984-10-16

    An array of rods is assembled to form a fuel element for a pressurized water reactor, the rods comprising zirconium alloy sheathed nuclear fuel pellets and containing helium. The helium gas pressure is selected for each rod so that it differs substantially from the helium gas pressure in its closest neighbors. In a preferred arrangement the rods are arranged in a square lattice and the helium gas pressure alternates between a relatively high value and a relatively low value so that each rod has as its closest neighbors up to four rods containing helium gas at the other pressure value.

  7. Nuclear reactor control assembly

    International Nuclear Information System (INIS)

    Negron, S.B.

    1991-01-01

    This patent describes an assembly for providing global power control in a nuclear reactor having the core split into two halves. It comprises a disk assembly formed from at least two disks each machined with an identical surface hole pattern such that rotation of one disk relative to the other causes the hole pattern to open or close, the disk assembly being positioned substantially at the longitudinal center of and coaxial with the core halves; and means for rotating at least one of the disks relative to the other

  8. Nuclear reactor facility

    International Nuclear Information System (INIS)

    Wampole, N.C.

    1978-01-01

    In order to improve the performance of manitenance and inspections it is proposed for a nuclear reactor facility with a primary circuit containing liquid metal to provide a thermally insulated chamber, within which are placed a number of components of the primary circuit, as e.g. valves, recirculation pump, heat exchangers. The isolated placement permit controlled preheating on one hand, but prevents undesirable heating of adjacent load-bearing elements on the other. The chamber is provided with heating devices and, on the outside, with cooling devices; it is of advantage to fill it with an inert gas. (UWI) 891 HP [de

  9. Advanced reactor development for non-electric applications

    International Nuclear Information System (INIS)

    Chang, M.H.; Kim, S.H.

    1996-01-01

    Advance in the nuclear reactor technology achieved through nuclear power programs carried out in the world has led nuclear communities to direct its attention to a better and peaceful utilization of nuclear energy in addition to that for power generation. The efforts for non-electric application of nuclear energy has been pursued in a limited number of countries in the world for their special needs. However, those needs and the associated efforts contributed largely to the development and practical realization of advanced reactors characterized by highly improved reactor safety and reliability by deploying the most up-to-date safety technologies. Due mainly to the special purpose of utilization, economic reasons and ease in implementation of new advanced technologies, small and medium reactors have become a major stream in the reactor developments for non-electric applications. The purpose of this paper is to provide, to the interested nuclear society, the overview of the development status and design characteristics of selected advanced nuclear reactors previously developed and/or currently under development specially for non-electric applications. Major design technologies employed in those reactors to enhance the reactor safety and reliability are reviewed to present the underlying principles of the design. Along with the overview, this paper also introduces a development program and major design characteristics of an advanced integral reactor (SMART) for co-generation purpose currently under conceptual development in Korea. (author)

  10. RB research nuclear reactor, Annual report for 1989, I - III

    International Nuclear Information System (INIS)

    Stefanovic, D.; Pesic, M.; Hadimahmutovic, N.; Vranic, S.; Petronijevic, M.; Jevremovic, M.; Ilic, I.

    1989-12-01

    This report is made of three parts. Part one contains a short description of the reactor, reactor operation, incidents, status of reactor equipment and components (nuclear fuel, heavy water, reactor vessel, heavy water circulation system, electronic, electric and mechanical equipment, auxiliary systems and Vax-8250 computer). It includes dosimetry and radiation protection data, personnel and financial data. Second part of this report in concerned with maintenance of reactor components and instrumentation. Part three includes data about reactor utilization during 1989

  11. Nuclear reactor container

    International Nuclear Information System (INIS)

    Kawabe, Ryuhei; Yamaki, Rika.

    1989-01-01

    Aerosol filters considered so far for nuclear reactor containers in conventional BWR type nuclear power plants make the facility larger and involve a risk of clogging. In view of the above, in the present invention, the diameter of a flow channel of gases entering from a bent pipe to a suppression pool is made smaller thereby decreasing the diameter of gas bubbles in the supperssional pool. Since this reduces the force of surface tension, the diameter of resulted gas bubbles is made remarkably smaller as compared with the case where the gases are released from the lower end of the bent pipe. Since the absorption velocity of bubble-entrained aerosols into water is in proportion to the square of the bubble diameter, the absorption efficiency can be increased remarkably by reducing the diameter of the gas bubbles. Accordingly, it is possible to improve the efficiency of eliminating radioactivity of released gases. (K.M.)

  12. 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.)

  13. 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 ...

  14. Linear regression and sensitivity analysis in nuclear reactor design

    International Nuclear Information System (INIS)

    Kumar, Akansha; Tsvetkov, Pavel V.; McClarren, Ryan G.

    2015-01-01

    Highlights: • Presented a benchmark for the applicability of linear regression to complex systems. • Applied linear regression to a nuclear reactor power system. • Performed neutronics, thermal–hydraulics, and energy conversion using Brayton’s cycle for the design of a GCFBR. • Performed detailed sensitivity analysis to a set of parameters in a nuclear reactor power system. • Modeled and developed reactor design using MCNP, regression using R, and thermal–hydraulics in Java. - Abstract: The paper presents a general strategy applicable for sensitivity analysis (SA), and uncertainity quantification analysis (UA) of parameters related to a nuclear reactor design. This work also validates the use of linear regression (LR) for predictive analysis in a nuclear reactor design. The analysis helps to determine the parameters on which a LR model can be fit for predictive analysis. For those parameters, a regression surface is created based on trial data and predictions are made using this surface. A general strategy of SA to determine and identify the influential parameters those affect the operation of the reactor is mentioned. Identification of design parameters and validation of linearity assumption for the application of LR of reactor design based on a set of tests is performed. The testing methods used to determine the behavior of the parameters can be used as a general strategy for UA, and SA of nuclear reactor models, and thermal hydraulics calculations. A design of a gas cooled fast breeder reactor (GCFBR), with thermal–hydraulics, and energy transfer has been used for the demonstration of this method. MCNP6 is used to simulate the GCFBR design, and perform the necessary criticality calculations. Java is used to build and run input samples, and to extract data from the output files of MCNP6, and R is used to perform regression analysis and other multivariate variance, and analysis of the collinearity of data

  15. Evaluation of deformation and fracture characteristics of nuclear reactor materials using ball indentation test technique

    International Nuclear Information System (INIS)

    Byun, T. S.; Hong, J. H.; Lee, B. S.; Park, D. G.; Kim, J. H.; Oh, Y. J.; Yoon, J. H.; Chi, S. H.; Kuk, I. H.; Kwon, D. I.; Lee, J. H.

    1998-05-01

    The present report describes the automated ball indentation test techniques and the results of their applications. The ball indentation test technique is an innovative method for evaluating the key mechanical properties from the indentation load-depth data. In the 1st chapter, the existing technique for evaluating basic deformation (tensile) properties is described in detail, and also the application result of the technique is presented. The through-thickness variations of mechanical properties in SA 508 C1.3 reactor pressure vessel steels were measured using an automated ball indentation (ABI) technique. In the 2nd chapter, a method under development, which is similar to that in the 1st chapter, is new method is based on the theoretical solutions rather than experimental relationships. The result of the application showed that the stress-strain curves of various metals were successfully determined with the method. In the 3rd chapter, a new theoretical model was proposed to estimate the fracture toughness of ferritic steels in the transition temperature region. The key concept of the model is that the indention energy to a critical load is related to the fracture energy of the material. The theory was applied to the reactor pressure vessel (RPV) base and weld metals. (author). 24 refs., 3 tabs., 6 figs

  16. Development of a method of testing irradiation devices by gamma scanning inside a research nuclear reactor

    International Nuclear Information System (INIS)

    Michel, Francois.

    1975-01-01

    A tridimensional experiment of spectrometry of an irradiation device located inside the reactor Siloe at a place shielded against spurious radiations, is exposed. The automatic scanning system that was developed, makes it possible to fully analyze in about 24 hours, the irradiation device (fuel pin, coolant and structures). The process combined with a 'pre-processing' program allows first partial results to be simultaneously obtained, more refined results being achieved during the next week, using the 'heavy processing'. The irradiation of the device is only interrupted during the compelling duty shutdown of the reactor, the evolution of the device during the successive irradiation cycles being thus followed up without pertubing said evolution. The reproducibility was studied at a local stage for testing the collimation of the detection unit (1% accuracy) and for testing the whole set 'processing measurement and computation' (5% accuracy). The sensitivity has been illustrated by detecting and measuring local singularities inside fuel (pellets), determining the detection efficiency dependence on the radial distribution of fission products and measuring nucleides inside the coolant flow (limiting value 10 -7 ). The accuracy of the method is evaluated at 5% for relative measurement of an experimental device during its in-pile lifetime and as 10% for quantitative absolute measurements [fr

  17. Economic aspects of electricity and industrial heat generating reactors

    International Nuclear Information System (INIS)

    Gaussens, J.; Moulle, N.; Dutheil, F.

    1964-01-01

    The economic advantage of electricity-generating nuclear stations decreases when their size decreases. However, when a counter-pressure turbine is joined on to a reactor and the residual heat can be properly used, it can be shown that fairly low capacity nuclear equipment may compete with conventional equipment under certain realistic enough conditions. The aim of this paper is to define these special conditions under which nuclear energy can be profitable. They are connected with the location and the general economic environment of the station, the pattern of the electricity and heat demands it must meet, the level of fuel and specific capital costs, nuclear and conventional. These conditions entail certain technical and economic specifications for the reactors used in this way otherwise they are unlikely to be competitive. In addition, these results are referred to the potential steam and electricity market, which leads us to examine certain uses for the heat generated by double purpose power stations; for example, to supply combined industrial plants, various types of town heating and for removal of salt from sea water. (authors) [fr

  18. Nuclear reactor installation

    International Nuclear Information System (INIS)

    Jungmann, A.

    1976-01-01

    A nuclear reactor metal pressure vessel is surrounded by a concrete wall forming an annular space around the vessel. Thermal insulation is in this space and surrounds the vessel, and a coolant-conductive layer is also in this space surrounding the thermal insulation, coolant forced through this layer reducing the thermal stress on the concrete wall. The coolant-conductive layer is formed by concrete blocks laid together and having coolant passages, these blocks being small enough individually to permit them to be cast from concrete at the reactor installation, the thermal insulation being formed by much larger sheet-metal clad concrete segments. Mortar is injected between the interfaces of the coolant-conductive layer and concrete wall and the interfaces between the fluid-conductive layer and the insulation, a layer of slippery sheet material being interposed between the insulation and the mortar. When the pressure vessel is thermally expanded by reactor operation, the annular space between it and the concrete wall is completely filled by these components so that zero-excursion rupture safeguard is provided for the vessel. 4 claims, 1 figure

  19. Nuclear reactor building

    International Nuclear Information System (INIS)

    Oshima, Nobuaki.

    1991-01-01

    The secondary container in a nuclear reactor building is made of a transparent structure having a shielding performance such as lead glass, by which the inside of the secondary container can be seen without undergoing radiation exposure. In addition, an operator transportation facility capable of carrying about 5 to 10 operators at one time is disposed, and the side of the facility on the secondary container is constituted with a transparent material such as glass, to provide a structure capable of observing the inside of the secondary container. The ventilation and air conditioning in the operator's transportation facility is in communication with the atmosphere of a not-controlled area. Accordingly, operators at the outside of the reactor building can reach the operator's transportation facility without taking and procedures for entering the controlled area and without undergoing radiation exposure. The inside of the secondary container in the reactor building can be seen from various directions through the transparent structure having the shielding performance. (N.H.)

  20. Helium heater design for the helium direct cycle component test facility. [for gas-cooled nuclear reactor power plant

    Science.gov (United States)

    Larson, V. R.; Gunn, S. V.; Lee, J. C.

    1975-01-01

    The paper describes a helium heater to be used to conduct non-nuclear demonstration tests of the complete power conversion loop for a direct-cycle gas-cooled nuclear reactor power plant. Requirements for the heater include: heating the helium to a 1500 F temperature, operating at a 1000 psia helium pressure, providing a thermal response capability and helium volume similar to that of the nuclear reactor, and a total heater system helium pressure drop of not more than 15 psi. The unique compact heater system design proposed consists of 18 heater modules; air preheaters, compressors, and compressor drive systems; an integral control system; piping; and auxiliary equipment. The heater modules incorporate the dual-concentric-tube 'Variflux' heat exchanger design which provides a controlled heat flux along the entire length of the tube element. The heater design as proposed will meet all system requirements. The heater uses pressurized combustion (50 psia) to provide intensive heat transfer, and to minimize furnace volume and heat storage mass.

  1. Ground testing of an SP-100 prototypic reactor

    International Nuclear Information System (INIS)

    Motwani, K.; Pflasterer, G.R.; Upton, H.; Lazarus, J.D.; Gluck, R.

    1988-01-01

    SP-100 is a space power system which is being developed by GE to meet future space electrical power requirements. The ground testing of an SP-100 prototypic reactor system will be conducted at the Westinghouse Hanford Company site located at Richland, Washington. The objective of this test is to demonstrate the performance of a full scale prototypic reactor system, including the reactor, control system and flight shield. The ground test system is designed to simulate the flight operating conditions while meeting all the necessary nuclear safety requirements in a gravity environment. The goal of the reactor ground test system is to establish confidence in the design maturity of the SP-100 space reactor power system and resolve the technical issues necessary for the development of a flight mission design

  2. Virtual nuclear reactor for education of nuclear reactor physics

    International Nuclear Information System (INIS)

    Tsuji, Masashi; Narabayashi, Takashi; Shimazu, Youichiro

    2008-01-01

    As one of projects that were programmed in the cultivation program for human resources in nuclear engineering sponsored by the Ministry of Economy, Trade and Industry, the development of a virtual reactor for education of nuclear reactor physics started in 2007. The purpose of the virtual nuclear reactor is to make nuclear reactor physics easily understood with aid of visualization. In the first year of this project, the neutron slowing down process was visualized. The data needed for visualization are provided by Monte Carlo calculations; The flights of the respective neutrons generated by nuclear fissions are traced through a reactor core until they disappear by neutron absorption or slow down to a thermal energy. With this visualization and an attached supplement textbook, it is expected that the learners can learn more clearly the physical implication of neutron slowing process that is mathematically described by the Boltzmann neutron transport equation. (author)

  3. The safety of Ontario's nuclear reactors

    International Nuclear Information System (INIS)

    1980-06-01

    A Select Committee of the Legislature of Ontario was established to examine the affairs of Ontario Hydro, the provincial electrical utility. Extensive public hearings were held on several topics including the safety of nuclear power reactors operating in Ontario. The Committee found that these reactors are acceptably safe. Many of the 24 recommendations in this report deal with the licensing process and public access to information. (O.T.)

  4. Design of radiation shields in nuclear reactor core

    International Nuclear Information System (INIS)

    Mousavi Shirazi, A.; Daneshvar, Sh.; Aghanajafi, C.; Jahanfarnia, Gh.; Rahgoshay, M.

    2008-01-01

    This article consists of designing radiation shields in the core of nuclear reactors to control and restrain the harmful nuclear radiations in the nuclear reactor cores. The radiation shields protect the loss of energy. caused by nuclear radiation in a nuclear reactor core and consequently, they cause to increase the efficiency of the reactor and decrease the risk of being under harmful radiations for the staff. In order to design these shields, by making advantages of the O ppenheim Electrical Network m ethod, the structure of the shields are physically simulated and by obtaining a special algorithm, the amount of optimized energy caused by nuclear radiations, is calculated

  5. Real time simulator for material testing reactor

    Energy Technology Data Exchange (ETDEWEB)

    Takemoto, Noriyuki; Imaizumi, Tomomi; Izumo, Hironobu; Hori, Naohiko; Suzuki, Masahide [Japan Atomic Energy Agency, Oarai Research and Development Center, Oarai, Ibaraki (Japan); Ishitsuka, Tatsuo; Tamura, Kazuo [ITOCHU Techno-Solutions Corp., Tokyo (Japan)

    2012-03-15

    Japan Atomic Energy Agency (JAEA) is now developing a real time simulator for a material testing reactor based on Japan Materials Testing Reactor (JMTR). The simulator treats reactor core system, primary and secondary cooling system, electricity system and irradiation facility systems. Possible simulations are normal reactor operation, unusual transient operation and accidental operation. The developed simulator also contains tool to revise/add facility in it for the future development. (author)

  6. Real time simulator for material testing reactor

    International Nuclear Information System (INIS)

    Takemoto, Noriyuki; Imaizumi, Tomomi; Izumo, Hironobu; Hori, Naohiko; Suzuki, Masahide; Ishitsuka, Tatsuo; Tamura, Kazuo

    2012-01-01

    Japan Atomic Energy Agency (JAEA) is now developing a real time simulator for a material testing reactor based on Japan Materials Testing Reactor (JMTR). The simulator treats reactor core system, primary and secondary cooling system, electricity system and irradiation facility systems. Possible simulations are normal reactor operation, unusual transient operation and accidental operation. The developed simulator also contains tool to revise/add facility in it for the future development. (author)

  7. Nuclear energy products except the electric power

    International Nuclear Information System (INIS)

    2004-01-01

    Technically the fission reactors, on service or under construction, can produce other products than the electric power. Meanwhile, these applications are known since the beginning of the reactors exploitation, they never have been developed industrially. This report examines the necessary technical characteristics for using the nuclear systems on non electric power applications with an economical efficiency. What are the markets for these products? What are the strategical challenges to favor the development of non electric power applications of the nuclear energy? (A.L.B.)

  8. Nuclear reaction data and nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Paver, N [University of Trieste (Italy); Herman, M [International Atomic Energy Agency, Vienna (Austria); Gandini, A [ENEA, Rome (Italy)

    2001-12-15

    These two volumes contain the lecture notes of the workshop 'Nuclear Reaction Data and Nuclear Reactors: Physics, Design and Safety', which was held at the Abdus Salam ICTP in the Spring of 2000. The workshop consisted of five weeks of lecture courses followed by practical computer exercises on nuclear data treatment and design of nuclear power systems. The spectrum of topics is wide enough to timely cover the state-of-the-art and the perspectives of this broad field. The first two weeks were devoted to nuclear reaction models and nuclear data evaluation. Nuclear data processing for applications to reactor calculations was the subject of the third week. On the last two weeks reactor physics and on-going projects in nuclear power generation, waste disposal and safety were presented.

  9. Progress of the United States foreign research reactor spent nuclear fuel acceptance program. Reduced enrichment for research and test reactors conference 2002

    International Nuclear Information System (INIS)

    Clapper, Maureen

    2002-01-01

    Foreign Research Reactor Spent nuclear fuel Acceptance Program is actively working with research reactors to accept eligible material before the Acceptance Policy proper expires in 2006. Reactors/governments wishing to participate should contact US immediately if they have not done so already. Program operations are changing to adapt to new challenges. We continue to promote the importance of this Program to senior management in the Department of Energy

  10. General discrimination technique to determine between earthquake and nuclear test with seismic data

    International Nuclear Information System (INIS)

    Bashillah Baharuddin; Alawiah Musa; Roslan Mohd Ali

    2007-01-01

    The Comprehensive Nuclear-Test-Ban Treaty (CTBT) was developed to ban of any nuclear weapon test explosion moreover will restrict the development and qualitative improvement of nuclear weapons and end the development of advanced new types of these weapons. The Treaty provides for a comprehensive global verification regime, which includes an International Monitoring System (IMS). The IMS comprises a network of 321 monitoring stations and 16 radionuclide laboratories that monitor the Earth for evidence of nuclear explosions, which cover underground, underwater and atmosphere environments. Presently, Malaysia receives seismic, infrasound, hydroacoustic and radionuclide data from the International Data Centre (IDC) of the CTBT. In order to maximise the use of the data for the purposes of the CTBT, the Malaysian Nuclear Agency is developing capability to analyse the data in order to detect nuclear weapon test, with an initial focus on the seismic data. Through the CTBT IMS, seismic data is constantly being obtained to monitor and detect nuclear explosions. However, in the process, other natural and man-made activities that generate seismic waves, especially earthquakes and large man-made explosions, are also detectable through the IMS, and need to be differentiated and discriminated before any nuclear explosions can be identified. The detection capability by using seismological methods was proven through simulated explosion tests at selected nuclear weapon test sites. This is supported by data previously collected from a total of 2089 nuclear weapon tests that have been carried out globally, out of which 1567 were underground, 514 in the atmosphere, including outer space, and 8 underwater. The discrimination of seismic data to detect nuclear explosions from natural earthquake and explosions can be undertaken through the identification of the epicentre location, hypocentre depth, magnitude and short-period discrimination of the seismic events. (Author)

  11. Improved nuclear reactor construction with bottom supported reactor vessel

    International Nuclear Information System (INIS)

    Sharbaugh, J.E.

    1987-01-01

    An improved liquid metal nuclear reactor construction has a reactor core and a generally cylindrical reactor vessel for holding liquid metal coolant and housing the core within the pool. A generally cylindrical concrete containment structure surrounds the reactor vessel and a central support pedestal is anchored to the containment structure base mat and supports the bottom wall of the reactor vessel and the reactor core. The periphery of the reactor vessel bore is supported by an annular structure which allows thermal expansion but not seismic motion of the vessel, and a bed of thermally insulating material uniformly supports the vessel base whilst allowing expansion thereof. A guard ring prevents lateral seismic motion of the upper end of the reactor vessel. The periphery of the core is supported by an annular structure supported by the vessel base and keyed to the vessel wall so as to be able to expand but not undergo seismic motion. A deck is supported on the containment structure above the reactor vessel open top by annular bellows, the deck carrying the reactor control rods such that heating of the reactor vessel results in upward expansion against the control rods. (author)

  12. IEEE No. 323, IEEE trial-use standard: General guide for qualifying Class I electric equipment for nuclear power generating stations

    International Nuclear Information System (INIS)

    Anon.

    1992-01-01

    This document describes the basic requirements for the qualification of Class I electric equipment. This is equipment which is essential to the safe shutdown and isolation of the reactor or whose failure or damage could result in significant release of radioactive material. The purpose of this document is to provide guidance for demonstrating the qualifications of electrical equipment as required in the IEEE Std 279 -- Criteria for Nuclear Power Generating Station Protection Systems, and IEEE Std 308 -- Criteria for Class 1E Electric Systems for Nuclear Power Generating Stations. The qualification methods described may be used in conjunction with the Guides for qualifying specific types of equipment, (see Foreword), for updating qualification following modifications or for qualifying equipment for which no applicable Guide exists

  13. Design and testing of integrated circuits for reactor protection channels

    International Nuclear Information System (INIS)

    Battle, R.E.; Vandermolen, R.I.; Jagadish, U.; Swail, B.K.; Naser, J.

    1995-01-01

    Custom and semicustom application-specific integrated circuit design and testing methods are investigated for use in research and commercial nuclear reactor safety systems. The Electric Power Research Institute and Oak Ridge National Laboratory are working together through a cooperative research and development agreement to apply modern technology to a nuclear reactor protection system. The purpose of this project is to demonstrate to the nuclear industry an alternative approach for new or upgrade reactor protection and safety system signal processing and voting logic. Motivation for this project stems from (1) the difficulty of proving that software-based protection systems are adequately reliable, (2) the obsolescence of the original equipment, and (3) the improved performance of digital processing. A demonstration model for protection system of PWR reactor has been designed and built

  14. Nuclear data requirements for fusion reactor nucleonics

    International Nuclear Information System (INIS)

    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

  15. Guidelines for nuclear reactor equipments safety-analysis

    International Nuclear Information System (INIS)

    1978-01-01

    The safety analysis in approving the applications for nuclear reactor constructions (or alterations) is performed by the Committee on Examination of Reactor Safety in accordance with various guidelines prescribed by the Atomic Energy Commission. In addition, the above Committee set forth its own regulations for the safety analysis on common problems among various types of nuclear reactors. This book has collected and edited those guidelines and regulations. It has two parts: Part I includes the guidelines issued to date by the Atomic Energy Commission: and Part II - regulations of the Committee. Part I has collected 8 categories of guidelines which relate to following matters: nuclear reactor sites analysis guidelines and standards for their applications; standard exposure dose of plutonium; nuclear ship operation guidelines; safety design analysis guidelines for light-water type, electricity generating nuclear reactor equipments; safety evaluation guidelines for emergency reactor core cooling system of light-water type power reactors; guidelines for exposure dose target values around light-water type electricity generating nuclear reactor equipments, and guidelines for evaluation of above target values; and meteorological guidelines for the safety analysis of electricity generating nuclear reactor equipments. Part II includes regulations of the Committee concerning - the fuel assembly used in boiling-water type and in pressurized-water type reactors; techniques of reactor core heat designs, etc. in boiling-water reactors; and others

  16. Nuclear reactor strategies

    International Nuclear Information System (INIS)

    Konno, H.; Srinivasan, T.N.

    1975-01-01

    Reference is made to a linear programming model considered by Hafele and Manne ('Strategies for a Transition from Fossil to Nuclear Fuels'. 11ASA Research Report RR-74-7) in which the sum of discounted costs of meeting demand for electrical and non-electrical energy over a horizon of 75 years divided into 25 periods of 3 years is minimised subject to constraints, inter alia, on the total availability of fossil fuel and low cost ($15/lb) natural uranium. The sensitivity of the Hafele-Manne results are explored with respect to changes in some crucial parameters and assumptions namely; variation in discount rate, variation in current costs of operation of HTRB, variation in costs and availability of natural uranium, market penetration constraints, changing capital costs, price responsive demands, petroleum prices, and minimisation of PETG consumption with constraints on the sum of discounted costs. (U.K.)

  17. Nuclear reactors safety issues

    International Nuclear Information System (INIS)

    Barre, Francois; Seiler, Nathalie

    2008-01-01

    Full text of publication follows: Since the seventies, economic incentives have led the utilities to drive a permanent evolution of the light water reactor (LWR). The evolution deals with the reactor designs as well as the way to operate them in a more flexible manner. It is for instance related to the fuel technologies and management. On the one hand, the technologies are in continuous evolution, such as the fuel pellets (MOX, Gd fuel, or Cr doped fuels..) as well as advanced cladding materials (M5 TM , MDA or ZIRLO). On the other hand, the fuel management is also subject to continuous evolution in particular in terms of increasing the level of burn-up, the reactor (core) power, the enrichment, as well as the duration of reactor cycles. For instance, in a few years in France, the burn-up has raised beyond the value of 39 GWj/t, initially authorized up to 52 GWj/t for the UO 2 fuel. In the near future, utilities foreseen to reach fuel burn-up of 60 GWj/t for MOX fuel and 70 GWj/t for UO 2 fuel. Furthermore, the future reactor of fourth generation will use new fuels of advanced conception. Furthermore with the objective of improving the safety margins, methods and calculation tools used by the utilities in the elaboration of their safety demonstrations submitted to the Safety Authority, are in movement. The margin evaluation methodologies often consist of a calculation chain of best-estimate multi-field simulations (e.g. various codes being coupled to simulate in a realistic way the evolution of the thermohydraulic, neutronic and mechanic state of the reactor). The statistical methods are more and more sophisticated and the computer codes are integrating ever-complex physical models (e.g. three-dimensional at fine scale). Following this evolution, the Institute of Radioprotection and Nuclear Safety (IRSN), whose one of the roles is to examine the safety records and to rend a technical expertise, considers the necessity of reevaluating the safety issues for advanced

  18. Dynamics of nuclear reactor operational cycles

    International Nuclear Information System (INIS)

    Chapman, L.D.; Wayland, J.R.

    With this system dynamics computer model, one can explore the long term effects of a nuclear reactor program. Given an input demand for reactors, the consequences on each sector and the interactions among sectors can be simulated to provide a better understanding of the time development of a nuclear reactor program. The model permits the determination of various levels of activity as a function of time for plant enrichment, fuel fabrication, fuel reprocessing and storage of waste products. In addition, the rates of construction of reactors, spent fuel transit, disposal of waste, mining, shipping, recycling and enrichment can be investigated for optimal planning purposes. The model has been written in a very general manner so that it can be used to simulate any nuclear reactor program. It is an easy task to relate the amount of accidental or operational release of radioactive contaminants into our environment to the activity levels of each of the above sectors. (U.S.)

  19. 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.

  20. Nuclear reactor effluent monitoring

    International Nuclear Information System (INIS)

    Minns, J.L.; Essig, T.H.

    1993-01-01

    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

  1. Ultrasonic in-service testing of pressure vessel bodies of nuclear power reactors

    International Nuclear Information System (INIS)

    Obraz, J.

    1978-01-01

    In-service ultrasonic testing of reactor pressure vessels is described using a system of probes for simultaneous testing of material or weld joint thicknesses. The signal is transmitted from a common output via a 30 m long cable to electronic evaluation equipment. The methods are described of ultrasonic detection of fatigue cracks. The static calculation of the dependence of echo amplitudes on crack orientation and the dynamic calculation of the crack orientation effect are described for the indirect reflection technique. In testing, angular probes with gap-type acoustic coupling operating at a frequency of 2 MHz were preferably used. For detecting planar defects of more than 10 mm in size inclined by more than +-10deg probes operating at a frequency of 1 MHz were more advantageous. The direct reflection technique is suitable for detecting defects near the surface (10 to 20 mm) and for cases when the indirect reflection technique cannot be used. For this technique a focusing probe operating at a frequency of 2 MHz is suitable. The strong dependence of the echo amplitude on the crack depth is a disadvantage of the technique. Defects near the surface, i.e., immediately under cladding are best detected by means of a double probe transmitting transversal waves at an angle of 60deg. Experimental measurements were carried out on materials with artificial defects of the type of bores with flat bottom. (J.P.)

  2. Nuclear reactors for the future

    International Nuclear Information System (INIS)

    Vijayan, P.K.; Kamble, M.T.; Dulera, I.V.

    2013-01-01

    For the sustainable development of nuclear power plants with enhanced safety features, economic competitiveness, proliferation resistance and physical protection, several advanced reactor developments have been initiated world-wide. The major advanced reactor initiatives and the proposed advanced reactor concepts have been briefly reviewed along with their advantages and challenges. Various advanced reactor designs being pursued in India have also been briefly described in the paper. (author)

  3. Activity transport in nuclear reactors

    International Nuclear Information System (INIS)

    Narasimhan, S.V.

    2000-01-01

    The chemistry of the primary coolant is such that the general material loss is immeasurably low. However, the generation of radioactive corrosion products in the coolant, their transportation and distribution to different out of core surfaces occur irrevocably through the life cycle of the reactor. This phenomena leading to the build up of radiation field, which is unique to the nuclear reactor systems, is the only major problem of any significance. Minimization of this phenomenon can be done by many ways. The processes involved in the mechanism of activity transport are quite complex and are not at all thoroughly understood. The codes that have been developed so far use many empirical coefficients for some of the rate processes, which are either partially justified by simulated experimental studies or supported theoretically. In a multi-metal system like that of the reactor, the corrosion rates or release rates need not be similar especially in reactors like PHWRs. The mechanisms involved in the formation of protective oxide coating are quite complex to model in a simplified manner. The paper brings out some these features involved in the activity transport modeling and analyses the need for extensive field related experimental work to substantiate the model. (author)

  4. Nuclear reactor cooling device

    International Nuclear Information System (INIS)

    Hoshi, Masaya; Makihara, Yoshiaki.

    1985-01-01

    Purpose: To improve the heat transfer performance, as well as reducing and simplifying the structure while preventing the intrusion of primary coolants to utilization systems. Constitution: Heat transfer from the primary coolant circuit to the utilization circuits is conducted by means of heat pipe type heat exchangers. The heat exchanger comprises a tightly closed vessel divided by a partition wall, through which a plurality of heat pipes are passed. The primary coolants receiving the heat from the nuclear reactor enter the first chamber of the heat exchanger to heat the evaporating portion of the heat pipes. The heated flow of steams in the heat pipes transfer to the condensating portion in the second chamber to conduct heat exchange with the utilization system. In this way, since secondary coolant circuits are saved, the heat transfer performance can be improved significantly and the risk of failure can be reduced. (Kamimura, M,)

  5. Nuclear reactor control apparatus

    International Nuclear Information System (INIS)

    Sridhar, B.N.

    1983-01-01

    Nuclear reactor safety rod release apparatus comprises a ring which carries detents normally positioned in an annular recess in outer side of the rod, the ring being held against the lower end of a drive shaft by magnetic force exerted by a solenoid carried by the drive shaft. When the solenoid is de-energized, the detent-carrying ring drops until the detents contact a cam surface associated with the lower end of the drive shaft, at which point the detents are cammed out of the recess in the safety rod to release the rod from the drive shaft. In preferred embodiments of the invention, an additional latch is provided to release a lower portion of a safety rod under conditions that may interfere with movement of the entire rod

  6. Nuclear reactor spacer assembly

    International Nuclear Information System (INIS)

    Anthony, A.J.; Groves, M.D.

    1979-01-01

    A fuel assembly for a nuclear reactor is disclosed wherein the fuel element receiving and supporting grid is comprised of a first metal, the guide tubes which pass through the grid assembly are comprised of a second metal and the grid is supported on the guide tubes by means of expanded sleeves located intermediate the grid and guide tubes. The fuel assembly is fabricated by inserting the sleeves, of initial outer diameter commensurate with the guide tube outer diameters, through the holes in the grid assembly provided for the guide tubes and thereafter expanding the sleeves radially outwardly along their entire length such that the guide tubes can subsequently be passed through the sleeves. The step of radial expansion, as a result of windows provided in the sleeves having dimensions commensurate with the geometry of the grid, mechanically captures the grid and simultaneously preloads the sleeve against the grid whereby relative motion between the grid and guide tube will be precluded

  7. Nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    Vikhorev, Yu.V.; Biryukov, G.I.; Kirilyuk, N.A.; Lobanov, V.N.

    1977-01-01

    A fuel assembly is proposed for nuclear reactors allowing remote replacement of control rod bundles or their shifting from one assembly to another, i.e., their multipurpose use. This leads to a significant increase in fuel assembly usability. In the fuel assembly the control rod bundle is placed in guide tube channels to which baffles are attached for fuel element spacing. The remote handling of control rods is provided by a hollow cylinder with openings in its lower bottom through which the control rods pass. All control rods in a bundle are mounted to a cross beam which in turn is mounted in the cylinder and is designed for grasping the whole rod bundle by a remotely controlled telescopic mechanism in bundle replacement or shifting. (Z.M.)

  8. Nuclear reactor container

    International Nuclear Information System (INIS)

    Shioiri, Akio.

    1992-01-01

    In a nuclear reactor container, a vent tube communication port is disposed to a pressure suppression pool at a position higher than the pool water therein for communication with an upper dry well, and the upper end opening of a dry well communication pipe is disposed at a position higher than the communication port. When condensate return pipeline is ruptured in the upper dry well, water in a water source pool is injected to the pressure vessel and partially discharged out of the ruptured port and a depressurization valve connected to the pressure vessel to the inside of the upper dry well. The discharged water stays in the upper dry well and, when the water level reaches the height of the vent tube communication port, it flows into the pressure suppression pool. Even in a state that the entire amount of water in the water source pool is supplied, since water does not reach the upper opening port of the dry well communication pipe, water does not flow into a lower dry well. Accordingly, the motor of a control rod drives disposed in the lower dry well can be prevented from submerging. The reactor core can be cooled more reliably, to improve the reliability of the pressure suppression function. (N.H.)

  9. Research nuclear reactor operation management

    International Nuclear Information System (INIS)

    Preda, M.; Carabulea, A.

    2008-01-01

    Some aspects of reactor operation management are highlighted. The main mission of the operational staff at a testing reactor is to operate it safely and efficiently, to ensure proper conditions for different research programs implying the use of the reactor. For reaching this aim, there were settled down operating plans for every objective, and procedure and working instructions for staff training were established, both for the start-up and for the safe operation of the reactor. Damages during operation or special situations which can arise, at stop, start-up, maintenance procedures were thoroughly considered. While the technical skill is considered to be the most important quality of the staff, the organising capacity is a must in the operation of any nuclear facility. Staff training aims at gaining both theoretical and practical experience based on standards about staff quality at each work level. 'Plow' sheet has to be carefully done, setting clear the decision responsibility for each person so that everyone's own technical level to be coupled to the problems which implies his responsibility. Possible events which may arise in operation, e.g., criticality, irradiation, contamination, and which do not arise in other fields, have to be carefully studied. One stresses that the management based on technical and scientific arguments have to cover through technical, economical and nuclear safety requirements a series of interlinked subprograms. Every such subprograms is subject to some peculiar demands by the help of which the entire activity field is coordinated. Hence for any subprogram there are established the objectives to be achieved, the applicable regulations, well-defined responsibilities, training of the personnel involved, the material and documentation basis required and activity planning. The following up of positive or negative responses generated by experiments and the information synthesis close the management scope. Important management aspects

  10. Review of electricity supply failures and plant improvements over 25 years operation of the Harwell materials test reactors

    International Nuclear Information System (INIS)

    Taylor, D.J.

    1986-01-01

    The evolution of the on-site electrical power sources is described, operational experience is reported and shortcomings are identified. Disturbances in the external power supplies to the reactors are listed for the past 25 years and failure probabilities are derived from this historical data. The 132 kV overhead supply to the Harwell site is identified as the source of nearly 90% of the disturbances. (author)

  11. Review of electricity supply failures and plant improvements over 25 years operation of the Harwell materials test reactors

    Energy Technology Data Exchange (ETDEWEB)

    Taylor, D. J. [UKAEA Harwell (United Kingdom)

    1986-02-15

    The evolution of the on-site electrical power sources is described, operational experience is reported and shortcomings are identified. Disturbances in the external power supplies to the reactors are listed for the past 25 years and failure probabilities are derived from this historical data. The 132 kV overhead supply to the Harwell site is identified as the source of nearly 90% of the disturbances. (author)

  12. Off reactor testings. Technological engineering applicative research

    International Nuclear Information System (INIS)

    Doca, Cezar

    2001-01-01

    By the end of year 2000 over 400 nuclear electro-power units were operating world wide, summing up a 350,000 MW total capacity, with a total production of 2,300 TWh, representing 16% of the world's electricity production. Other 36 units, totalizing 28,000 MW, were in construction, while a manifest orientation towards nuclear power development was observed in principal Asian countries like China, India, Japan and Korea. In the same world's trend one find also Romania, the Cernavoda NPP Unit 1 generating electrical energy into the national system beginning with 2 December 1996. Recently, the commercial contract was completed for finishing the Cernavoda NPP Unit 2 and launching it into operation by the end of year 2004. An important role in developing the activity of research and technological engineering, as technical support for manufacturing the CANDU type nuclear fuel and supplying with equipment the Cernavoda units, was played by the Division 7 TAR of the INR Pitesti. Qualification testings were conducted for: - off-reactor CANDU type nuclear fuel; - FARE tools, pressure regulators, explosion proof panels; channel shutting, as well as functional testing for spare pushing facility as a first step in the frame of the qualification tests for the charging/discharging machine (MID) 4 and 5 endings. Testing facilities are described, as well as high pressure hot/cool loops, measuring chains, all of them fulfilling the requirements of quality assurance. The nuclear fuel off-reactor tests were carried out to determine: strength; endurance; impact, pressure fall and wear resistance. For Cernavoda NPP equipment testings were carried out for: the explosion proof panels, pressure regulators, behaviour to vibration and wear of the steam generation tubings, effects of vibration upon different electronic component, channel shutting (for Cernavoda Unit 2), MID operating at 300 and 500 cycles. A number of R and D programs were conducted in the frame of division 7 TAR of INR

  13. Concerning modification of plan for installation of reactors (addition of No.4 reactor and modification of No.1, No.2 and No.3 reactor facilities) at Hamaoka nuclear power plant of Chubu Electric Power Co., Ltd. (reply to inquiry)

    International Nuclear Information System (INIS)

    1988-01-01

    The Nuclear Safety Commission on Oct. 22, 1987 directed the Nuclear Reactor Safety Expert Group to carry out a study on it, made deliberations after receiving a report from the Group on Jun. 28, and submitted the findings to the Minister of International Trade and Industry on Jul. 14. The study and deliberations were intended to determine the conformity of the modifications to the applicable laws relating to control of nuclear material, nuclear fuel and nuclear reactor. The investigation on the location covered the site conditions, geology, seismic environment, meteorology, hydrology, and social environment. The investigations on the safety design of the reactor facilities addressed the design of the overall reactor facilities, anti-earthquake design, and design of other equipment including reactor core, fuel equipment, monitors, reactivity controllers, safety protection equipment, and emergency core cooler. Other investigations included exposure evaluation and accident analysis. It was concluded that the modifications would not have adverse effects on the safety of the reactor facilities. (Nogami, K.)

  14. High-temperature and breeder reactors - economic nuclear reactors of the future

    International Nuclear Information System (INIS)

    Djalilzadeh, A.M.

    1977-01-01

    The thesis begins with a review of the theory of nuclear fission and sections on the basic technology of nuclear reactors and the development of the first generation of gas-cooled reactors applied to electricity generation. It then deals in some detail with currently available and suggested types of high temperature reactor and with some related subsidiary issues such as the coupling of different reactor systems and various schemes for combining nuclear reactors with chemical processes (hydrogenation, hydrogen production, etc.), going on to discuss breeder reactors and their application. Further sections deal with questions of cost, comparison of nuclear with coal- and oil-fired stations, system analysis of reactor systems and the effect of nuclear generation on electricity supply. (C.J.O.G.)

  15. Analytical hierarchy process for the selection of nuclear reactors for Mexico

    International Nuclear Information System (INIS)

    Martin-Del-Campo, C.; Nelson, P. F.; Francois, J. L.

    2008-01-01

    The Analytical Hierarchy Process (AHP) was applied as a decision making technique to select the next new nuclear power reactor for Mexico. The main objective is to select the most suitable nuclear reactor technology for Mexico, to start the bidding process within one or two years, and the reactor to initiate commercial operation by 2016. Four alternatives which comply with these specific restrictions were compared: the Advanced Boiling Water Reactor (AB WR) from General Electric, Hitachi and Toshiba; the European Pressurized Water Reactor (ERR) from AREVA; the pressurized water reactor - advanced passive plant, AP1000, from Westinghouse; and the Economic Simplified Boiling Water Reactor (ESBWR) from General Electric. The evaluation criteria include economic and safety indicator, some of them quantitative and others qualitative. Additional complexity to the solution of this problem is that there are stakeholders that can be involved in the definition of the evaluation criteria and in the definition of relative importance between them, according to each stakeholder interests or benefits. Only the two main stakeholders were considered, these are the Federal Commission of Electricity (Comision Federal de Electricidad, CFE) and the National Commission of Nuclear Safety and Safeguards (Comision Nacional de Seguridad Nuclear y Salvaguardias, CNSNS). Considering the evaluation criteria used and the authors' assessment of their importance, the ABWR was ranked the highest. Sensitivity analyses were performed in order to define the most suitable reactor for the needs of Mexico. (authors)

  16. Graphite materials for nuclear reactors

    International Nuclear Information System (INIS)

    Oku, Tatsuo

    1991-01-01

    Graphite materials have been used in the nuclear fission reactors from the beginning of the reactor development for the speed reduction and reflection of neutron. Graphite materials are used both as a moderator and as a reflector in the core of high temperature gas-cooled reactors, and both as a radiation shielding material and as a reflector in the surrounding of the core for the fast breeder reactor. On the other hand, graphite materials are being positively used as a first wall of plasma as it is known that low Z materials are useful for holding high temperature plasma in the nuclear fusion devices. In this paper the present status of the application of graphite materials to the nuclear fission reactors and fusion devices (reactors) is presented. In addition, a part of results on the related properties to the structural design and safety evaluation and results examined on the subjects that should be done in the future are also described. (author)

  17. Nuclear Power Reactor simulator - based training program

    International Nuclear Information System (INIS)

    Abdelwahab, S.A.S.

    2009-01-01

    nuclear power stations will continue playing a major role as an energy source for electric generation and heat production in the world. in this paper, a nuclear power reactor simulator- based training program will be presented . this program is designed to aid in training of the reactor operators about the principles of operation of the plant. also it could help the researchers and the designers to analyze and to estimate the performance of the nuclear reactors and facilitate further studies for selection of the proper controller and its optimization process as it is difficult and time consuming to do all experiments in the real nuclear environment.this program is written in MATLAB code as MATLAB software provides sophisticated tools comparable to those in other software such as visual basic for the creation of graphical user interface (GUI). moreover MATLAB is available for all major operating systems. the used SIMULINK reactor model for the nuclear reactor can be used to model different types by adopting appropriate parameters. the model of each component of the reactor is based on physical laws rather than the use of look up tables or curve fitting.this simulation based training program will improve acquisition and retention knowledge also trainee will learn faster and will have better attitude

  18. Small reactors and the 'second nuclear era'

    International Nuclear Information System (INIS)

    Egan, J.R.

    1984-01-01

    Predictions of the nuclear industry's demise are premature and distort both history and politics. The industry is reemerging in a form commensurate with the priorities of those people and nations controlling the global forces of production. The current lull in plant orders is due primarily to the world recession and to factors related specifically to reactor size. Traditional economies of scale for nuclear plants have been greatly exaggerated. Reactor vendors and governments in Great Britain, France, West Germany, Japan, the United States, Sweden, Canada, and the Soviet Union are developing small reactors for both domestic applications and export to the Third World. The prefabricated, factory-assembled plants under 500 MWe may alleviate many of the existing socioeconomic constraints on nuclear manufacturing, construction, and operation. In the industrialized world, small reactors could furnish a qualitatively new energy option for utilities. But developing nations hold the largest potential market for small reactors due to the modest size of their electrical systems. These units could double or triple the market potential for nuclear power in this century. Small reactors will both qualitatively and quantitatively change the nature of nuclear technology transfers, offering unique advantages and problems vis-a-vis conventional arrangements. (author)

  19. State of exposure control for workers engaging in radiation works and state of radioactive waste management in nuclear reactor facilities for test and research and nuclear reactor facilities at research and development stage, fiscal year 1995

    International Nuclear Information System (INIS)

    1996-01-01

    This is the summary of the reports submitted in fiscal year 1995 by the installers of the nuclear reactor facilities for test and research or at research and development stage, conforming to the related law. The individual dose equivalent of the workers engaging in radiation works in fiscal year 1995 was sufficiently lower than the prescribed limit in all reactor facilities. As for the released quantities of gaseous and liquid wastes, the radioactive substances in the air and water outside the monitor zones never exceeded the prescribed concentration limit in all reactor facilities. In the reactor facilities, for which the target values of release control have been determined, the values were less than the targets in all cases. The increase of stored radioactive solid waste decreased as the dismantling works of the reactor auxiliary system of the nuclear powered ship 'Mutsu' were finished in fiscal year 1994. As the amount of stored radioactive solid waste approaches the installed capacity, the preservation capacity of the existing waste preservation building was increased. (K.I.)

  20. Nuclear reactor container

    International Nuclear Information System (INIS)

    Moriyama, Takeo; Ochiai, Kanehiro; Niino, Tsuyoshi; Kodama, Toyokazu; Hirako, Shizuka.

    1988-01-01

    Purpose: To obtain structures suitable to a container structures for nuclear power plants used in those districts where earthquakes occur frequently, in which no local stresses are caused to the fundamental base portions and the workability for the fundamental structures is improved. Constitution: Basic stabilizers are attached to a nuclear reactor container (PCV) and a basic concrete recess for receiving a basic stabilizer is disposed in basic concretes. A top stabilizer is joined and fixed to a top stabilizer receiving plate at the inside of an outer shielding wall. On the other hand, a PCV top recess for conducting the load of PCV to the top stabilizer is attached to the top of the PCV. By disposing stabilizer structures allowing miner displacements at the two points, that is, the top and the lowermost portion of the PCV, no local stress concentrations can be generated to the extension on the axial direction of components due to the inner pressure of the PCV and to the horizontal load applied to the upper portion of the PCV upon earthquakes. (Yoshino, Y.)

  1. Power supply with nuclear reactor

    International Nuclear Information System (INIS)

    Cook, B.M.

    1982-01-01

    Each parameter of the processes of a nuclear reactor and components operatively associated therewith is monitored by a set of four like sensors. A trip system normally operates on a 'two out of four' configuration; i.e., to trip the reactor it is necessary that at least two sensors of a set sense an off-normal parameter. This assumes that all sensors are in normal operating condition. However, when a sensor is in test or is subject to maintenance or is defective or disabled, the 'two out of four' configuration would be reduced to a 'one out of three' configuration because the affected sensor is taken out of service. This would expose the system to the possibility that a single sensor failure, which may be spurious, will cause a trip of the reactor. To prevent this, it is necessary that the affected sensor be bypassed. If only one sensor is bypassed, the system operates on a 'two out of three' configuration. With two sensors bypassed, the sensing of an off-normal parameter by a third sensor trips the reactor

  2. Safety of nuclear power reactors

    International Nuclear Information System (INIS)

    MacPherson, H.G.

    1982-01-01

    Safety is the major public issue to be resolved or accommodated if nuclear power is to have a future. Probabilistic Risk Analysis (PRA) of accidental releases of low-level radiation, the spread and activity of radiation in populated areas, and the impacts on public health from exposure evolved from the earlier Rasmussen Reactor Safety Study. Applications of the PRA technique have identified design peculiarities in specific reactors, thus increasing reactor safety and establishing a quide for evaluating reactor regulations. The Nuclear Regulatory Commission and reactor vendors must share with utilities the responsibility for reactor safety in the US and for providing reasonable assurance to the public. This entails persuasive public education and information that with safety a top priority, changes now being made in light water reactor hardware and operations will be adequate. 17 references, 2 figures, 2 tables

  3. Nuclear reactor power supply system

    International Nuclear Information System (INIS)

    Cook, B.M.

    1982-01-01

    The redundant signals from the sensor assemblies measuring the process parameters of a nuclear reactor power supply are transmitted each in its turn to a protection system which operates to actuate the protection apparatus for signals indicating off-process conditions. Each sensor assembly includes a number of like sensors measuring the same parameters. The sets of process signals derived from the sensor assemblies are each in its turn transmitted from the protection system to the control system which impresses control signals on the reactor or its components to counteract the tendency for conditions to drift off-normal status requiring operation of the protection system. A parameter signal selector prevents a parameter signal which differs from the other parameter signals of the set by more than twice the allowable variation from passing to the control system. Test signals are periodically impressed by a test unit on a selected pair of a selection unit and control channels. This arrangement eliminates the possibility that a single component failure which may be spurious will cause an inadvertent trip of the reactor during test. (author)

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

    International Nuclear Information System (INIS)

    2009-01-01

    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

  5. Maintenance quality surveillance for nuclear reactor instrument and control systems

    International Nuclear Information System (INIS)

    Clement, T.M.

    1976-01-01

    A description is given of a formal program of mandatory testing and inspection for assurance of reliability at the N-reactor. The techniques and procedures are called Equipment Maintenance Standards (EMS) and cover the nuclear plant systems which affect nuclear safety, environmental control, continuity of operation and reactor life. The scope of the program is such that all electrical, chemical, mechanical, and nuclear devices with interconnecting circuitry, linkages, and piping are functionally checked and/or inspected from sensors through actuating media necessary for carrying out the defined functions under startup, operating, and shutdown conditions and also under any credible accident condition. Records are kept of the AS FOUND results of the tests and inspections along with adjustments and repairs required to correct any out-of-limits conditions found. These records are accumulated and are used as a source for quantitative reliability information and for evaluating wear-out or design problems

  6. Complete automation of nuclear reactors control

    International Nuclear Information System (INIS)

    Weill, J.

    1955-01-01

    The use of nuclear reactor for energy production induces the installation of automatic control systems which need to be safe enough and can adapt to the industrial scale of energy production. These automatic control systems have to insure the constancy of power level and adjust the power produced to the energy demand. Two functioning modes are considered: nuclear plant connected up to other electric production systems as hydraulic or thermic plants or nuclear plants functioning on an independent network. For nuclear plants connected up with other production plants, xenon poisoning and operating cost lead to keep working at maximum power the nuclear reactors. Thus, the power modulation control system will not be considered and only start-up control, safety control, and control systems will be automated. For nuclear power plants working on an independent network, the power modulation control system is needed to economize fuel. It described the automated control system for reactors functioning with constant power: a power measurement system constituted of an ionization chamber and a direct-current amplifier will control the steadfastness of the power produced. For reactors functioning with variable power, the automated power control system will allow to change the power and maintain it steady with all the necessary safety and will control that working conditions under P max and R max (maximum power and maximum reactivity). The effects of temperature and xenon poisoning will also be discussed. Safety systems will be added to stop completely the functioning of the reactor if P max is reached. (M.P.)

  7. General Electric Company proposed test and evaluation plan, commercial buildings. National Solar Demonstration Program

    Energy Technology Data Exchange (ETDEWEB)

    None

    1976-04-01

    The general requirements and methods for instrumenting, testing, and evaluating solar HVAC systems forming a part of ERDA's ''Commercial Demonstration Program'' commensurate with ERDA 23A and the Proposed Management Plan 75SDS4270 are defined. Design requirements are specified for the performance of components and subsystems comprising the instrumentation and data gathering system, as well as the support functions required to perform the diagnostic measurements, collection and processing of data, and documentation of reports on solar HVAC system performance, including economic and societal evaluations.

  8. Definition of parameters for a test section for the analysis of natural convection and coolant loss in the AP1000 nuclear reactor by similarity laws and fractional scaling analysis

    Energy Technology Data Exchange (ETDEWEB)

    Cadiz, Luis Felipe S.; Bezerra, Mario Augusto [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil). Departamento de Energia Nuclear; Lima, Fernando Roberto A., E-mail: falima@cnen.gov.br [Centro Regional de Ciências Nucleares do Nordeste (CRCN-NE/CNEN-PB), Recife, PB (Brazil)

    2017-07-01

    The present work develops and analyzes the main parameters of a test section for natural convection in case of a failure of the pumping system as much as the loss of coolant in refrigeration accidents. For this realization, a combination of laws of basic similarity and an innovative scale methodology, known as Fractional Scaling Analysis (FSA), was developed. The depressurizing is analyzed when a rupture occurs in one of the primary system piping of the AP1000 nuclear reactor. This reactor is developed by Westinghouse Electric Co., which is a PWR (Pressurized Water Reactor) with an electric power equal to 1000MW. Such a reactor is provided with a passive safety system that promotes considerable improvements in the safety, reliability, protection and reduction of costs of a nuclear power plant. The FSA is based on two concepts: fractional scale and hierarchy. It is used to provide experimental data that generate quantitative evaluation criteria as well as operational parameters in thermal and hydraulic processes of nuclear power plants. The results were analyzed with the use of computational codes. (author)

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

    International Nuclear Information System (INIS)

    1992-09-01

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

  10. Small nuclear reactor safety design requirements for autonomous operation

    International Nuclear Information System (INIS)

    Kozier, K.S.; Kupca, S.

    1997-01-01

    Small nuclear power reactors offer compelling safety advantages in terms of the limited consequences that can arise from major accident events and the enhanced ability to use reliable, passive means to eliminate their occurrence by design. Accordingly, for some small reactor designs featuring a high degree of safety autonomy, it may be-possible to delineate a ''safety envelope'' for a given set of reactor circumstances within which safe reactor operation can be guaranteed without outside intervention for time periods of practical significance (i.e., days or weeks). The capability to operate a small reactor without the need for highly skilled technical staff permanently present, but with continuous remote monitoring, would aid the economic case for small reactors, simplify their use in remote regions and enhance safety by limiting the potential for accidents initiated by inappropriate operator action. This paper considers some of the technical design options and issues associated with the use of small power reactors in an autonomous mode for limited periods. The focus is on systems that are suitable for a variety of applications, producing steam for electricity generation, district heating, water desalination and/or marine propulsion. Near-term prospects at low power levels favour the use of pressurized, light-water-cooled reactor designs, among which those having an integral core arrangement appear to offer cost and passive-safety advantages. Small integral pressurized water reactors have been studied in many countries, including the test operation of prototype systems. (author)

  11. 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.

  12. Small specimen test technology of fracture toughness in structural material F82H steel for fusion nuclear reactors

    International Nuclear Information System (INIS)

    Wakai, Eiichi; Ohtsuka, Hideo; Jitsukawa, Shiro; Matsukawa, Shingo; Ando, Masami

    2006-03-01

    Small specimen test technology (SSTT) has been developed to investigate mechanical properties of nuclear materials. SSTT has been driven by limited availability of effective irradiation volumes in test reactors and accelerator-based neutron and charged particle sources, and it is very useful for the reduction of waste materials produced in nuclear engineering. In this study new bend test machines have been developed to obtain fracture behaviors of F82H steel for very small bend specimens of pre-cracked t/2-1/3CVN (Charpy V-notch) with 20 mm-length and DFMB (deformation and fracture mini bend specimen) with 9 mm-length and disk compact tension of 0.18DCT type, and fracture behaviors were examined to evaluate DBTT (ductile-brittle transition temperature) at temperature from -180 to 25degC. The effect of specimen size on DBTT of F82H steel was also examined by using Charpy type specimens such as 1/2t-CVN, 1/3CVN and t/2-1/3CVN. In this paper, it also provides the information of the specimens irradiated at 250degC and 350degC to about 2 dpa in the capsule of 04M-67A and 04M-68A of JMTR experiments. (author)

  13. Electricity utilities: Nuclear sector

    International Nuclear Information System (INIS)

    Brosche, D.

    1992-01-01

    The safe and economic operation of nuclear power plants requires an appropriate infrastructure on the part of the operator as well as a high level of technical quality of the plants and of qualification of the personnel. Added to this are a variety of services rendered by specialist firms. The Bayernwerk utility, with plants of its own, has played a major role in the development of nuclear power in the Federal Republic of Germany. The importance of nuclear power to this firm is reflected in the pattern of its electricity sources and in the composition of its power plants. (orig.) [de

  14. EPR (European Pressurized water Reactor) The advanced nuclear reactor

    International Nuclear Information System (INIS)

    2004-01-01

    Nuclear energy, which provides a steady supply of electricity at low cost, has its rightful place in the energy mix of the 21. century, which puts the emphasis on sustainable development. The EPR is the only 3. generation reactor under construction today. It is an evolutionary reactor that represents a new generation of pressurized water reactors with no break in the technology used for the most recent models. The EPR was developed by Framatome and Siemens, whose nuclear activities were combined in January 2001 to form Framatome ANP, a subsidiary of AREVA and Siemens. EDF and the major German electricity companies played an active part in the project. The safety authorities of the two countries joined forces to bring their respective safety standards into line and draw up joint design rules for the new reactor. The project had three objectives: meet the requirements of European utilities, comply with the safety standards laid down by the French safety authority for future pressurized water reactors, in concert with its German counterpart, and make nuclear energy even more competitive than energy generated using fossil fuels. The EPR can guarantee a safe, inexpensive electricity supply, without adding to the greenhouse effect. It meets the requirements of the safety authorities and lives up to the expectations of electricity utilities. This document presents the main characteristics of the EPR, and in particular the additional measures to prevent the occurrence of events likely to damage the core, the leak-tight containment, the measures to reduce the exposure of operating and maintenance personnel, the solutions for an even greater protection of the environment. The foreseen development of the EPR in France and abroad (Finland, China, the United States) is summarized

  15. Support vector machines for nuclear reactor state estimation

    Energy Technology Data Exchange (ETDEWEB)

    Zavaljevski, N.; Gross, K. C.

    2000-02-14

    Validation of nuclear power reactor signals is often performed by comparing signal prototypes with the actual reactor signals. The signal prototypes are often computed based on empirical data. The implementation of an estimation algorithm which can make predictions on limited data is an important issue. A new machine learning algorithm called support vector machines (SVMS) recently developed by Vladimir Vapnik and his coworkers enables a high level of generalization with finite high-dimensional data. The improved generalization in comparison with standard methods like neural networks is due mainly to the following characteristics of the method. The input data space is transformed into a high-dimensional feature space using a kernel function, and the learning problem is formulated as a convex quadratic programming problem with a unique solution. In this paper the authors have applied the SVM method for data-based state estimation in nuclear power reactors. In particular, they implemented and tested kernels developed at Argonne National Laboratory for the Multivariate State Estimation Technique (MSET), a nonlinear, nonparametric estimation technique with a wide range of applications in nuclear reactors. The methodology has been applied to three data sets from experimental and commercial nuclear power reactor applications. The results are promising. The combination of MSET kernels with the SVM method has better noise reduction and generalization properties than the standard MSET algorithm.

  16. Support vector machines for nuclear reactor state estimation

    International Nuclear Information System (INIS)

    Zavaljevski, N.; Gross, K. C.

    2000-01-01

    Validation of nuclear power reactor signals is often performed by comparing signal prototypes with the actual reactor signals. The signal prototypes are often computed based on empirical data. The implementation of an estimation algorithm which can make predictions on limited data is an important issue. A new machine learning algorithm called support vector machines (SVMS) recently developed by Vladimir Vapnik and his coworkers enables a high level of generalization with finite high-dimensional data. The improved generalization in comparison with standard methods like neural networks is due mainly to the following characteristics of the method. The input data space is transformed into a high-dimensional feature space using a kernel function, and the learning problem is formulated as a convex quadratic programming problem with a unique solution. In this paper the authors have applied the SVM method for data-based state estimation in nuclear power reactors. In particular, they implemented and tested kernels developed at Argonne National Laboratory for the Multivariate State Estimation Technique (MSET), a nonlinear, nonparametric estimation technique with a wide range of applications in nuclear reactors. The methodology has been applied to three data sets from experimental and commercial nuclear power reactor applications. The results are promising. The combination of MSET kernels with the SVM method has better noise reduction and generalization properties than the standard MSET algorithm

  17. Contribution of prototypic material tests on the Plinius platform to the study of nuclear reactor severe accident

    International Nuclear Information System (INIS)

    Journeau, Ch.

    2008-01-01

    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)

  18. the JHR Material Testing Reactor

    International Nuclear Information System (INIS)

    Roure, C.; Cornu, B.; Berthet, B.; Simon, E.; Estre, N.; Guimbal, P.; Kinnunen, P.; Kotiluoto, P.

    2013-06-01

    The Jules Horowitz Reactor (JHR) is a European experimental reactor under construction in CEA Cadarache. It will be dedicated to material and fuel irradiation tests, and to medical isotopes production. Non-Destructive nuclear Examinations systems (NDE) will be implemented in pools to analyse the irradiated fuel or tested material in their supporting experimental irradiation devices extracted from the core or its immediate periphery. The Nuclear Measurement Laboratory (NML) of CEA Cadarache is working in collaboration with VTT (Technical Research Centre in Finland) in designing and developing NDE systems implementing gamma-ray spectroscopy and high energy X-ray imaging of the sample and irradiation device. CEA is also designing a neutron radiography system for which NML is working on the detection system. Design studies are performed with Monte Carlo transport codes and specific simulation tools developed by the NML for Xray and neutron imaging. (authors)

  19. IEEE standard for type test of class 1E electric cables, field splices, and connections for nuclear power generating stations

    International Nuclear Information System (INIS)

    Anon.

    1974-01-01

    The Institute of Electrical and Electronics Engineers has generated this document to provide guidance for developing a program to type test cables, field splices, and connections and obtain specific type test data. It supplements IEEE Std 323-1974 Standard for Qualifying Class IE Equipment for Nuclear Power Generating Stations, which describes basic requirements for equipment qualification. It is the integrated performance of the structures, fluid systems, the electrical systems, the instrumentation systems of the station, and in particular, the plant protection system, that limits the consequences of accidents. Seismic effects on installed cable systems are not within the scope of this document. Section 2 of this guide is an example of type tests. It is the purpose of this guide to deal with cable and connections; however, at the time of issue, detailed examples of tests for connections were not available

  20. 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.

  1. The control of emissions from nuclear power reactors in Canada

    International Nuclear Information System (INIS)

    Gorman, D.J.; Neil, B.C.J.; Chatterjee, R.M.

    1988-01-01

    Nuclear power reactors in Canada are of the CANDU pressurised heavy water design. These are located in the provinces of Ontario, Quebec, and New Brunswick. Most of the nuclear generating capacity is in the province of Ontario which has 16 commissioned reactors with a total capacity of 11,500 MWe. There are four reactors under construction with an additional capacity of 3400 MWe. Nuclear power currently accounts for approximately 50% of the electrical power generation of Ontario. Regulation of the reactors is a Federal Government responsibility administered by the Atomic Energy Control Board (AECB) which licenses the reactors and sets occupational and public dose limits

  2. Nuclear reactor kinetics and control

    International Nuclear Information System (INIS)

    Lewins, J.

    1978-01-01

    A consistent, integrated account of modern developments in the study of nuclear reactor kinetics and the problem of their efficient and safe control. It aims to prepare the student for advanced study and research or practical work in the field. Special features include treatments of noise theory, reliability theory and safety related studies. It covers all aspects of the operation and control of nuclear reactors, power and research and is complete in providing physical data methods of calculation and solution including questions of equipment reliability. The work uses illustrations of the main types of reactors in use in the UK, USA and Europe. Each chapter contains problems and worked examples suitable for course work and study. The subject is covered in chapters, entitled: introductory review; neutron and precursor equations; elementary solutions at low power; linear reactor process dynamics with feedback; power reactor control systems; fluctuations and reactor noise; safety and reliability; nonlinear systems (safety and control); analogue computing. (author)

  3. Safety re-assessment of AECL test and research reactors

    International Nuclear Information System (INIS)

    Winfield, D.J.

    1990-01-01

    Atomic Energy of Canada Limited currently has four operating engineering test/research reactors of various sizes and ages; a new isotope-production reactor Maple-X10, under construction at Chalk River Nuclear Laboratories (CRNL), and a heating demonstration reactor, SDR, undergoing high-power commissioning at Whiteshell Nuclear Research Establishment (WNRE). The company is also performing design studies of small reactors for hot water and electricity production. The older reactors are ZED-2, PTR, NRX, and NRU; these range in age from 42 years (NRX) to 29 years (ZED-2). Since 1984, limited-scope safety re-assessments have been underway on three of these reactors (ZED-2, NRX AND NRU). ZED-2 and PTR are operated by the Reactor Physics Branch; all other reactors are operated by the respective site Reactor Operations Branches. For the older reactors the original safety reports produced were entirely deterministic in nature and based on the design-basis accident concept. The limited scope safety re-assessments for these older reactors, carried out over the past 5 years, have comprised both quantitative probabilistic safety-assessment techniques, such as event tree and fault analysis, and/or qualitative techniques, such as failure mode and effect analysis. The technique used for an individual assessment was dependent upon the specific scope required. This paper discusses the types of analyses carried out, specific insights/recommendations resulting from the analysis, and the plan for future analysis. In addition, during the last four years safety assessments have been carried out on the new isotope-, heat-, and electricity-producing reactors, as part of the safety design review, commissioning and licensing activities

  4. Calculation of particulate dispersion in a design-basis tornadic storm from the General Electric Vallecitos Nuclear Center, Vallecitos, California

    International Nuclear Information System (INIS)

    Pepper, D.W.

    1979-11-01

    A three-dimensional numerical model is used to calculate ground-level air concentration and deposition (due to precipitation scavenging) after a hypothetical tornado strike at the General Electric Vallecitos Nuclear Center at Vallecitos, California. Plutonium particles less than 20 μm in diameter are assumed to be lifted into the tornadic storm cell by the vortex. The rotational characteristics of the tornadic storm are embedded within the larger mesoscale flow of the storm system. The design-basis translational wind values are based on probabilities associated with existing records of tornado strikes in the vicinity of the plant site. Turbulence exchange coefficients are based on empirical values deduced from experimental data in severe storms and from theoretical assumptions obtained from the literature. The method of moments is used to incorporate subgrid-scale resolution of the concentration within a grid cell volume. This method is a quasi-Lagrangian scheme which minimizes numerical error associated with advection. In all case studies, the effects of updrafts and downdrafts, coupled with scavenging of the particulates by precipitation, account for most of the material being deposited within 20 to 45 km downwind of the plant site. Ground-level isopleths in the x-y plane show that most of the material is deposited behind and slightly to the left of the centerline trajectory of the storm. Approximately 5% of the material is dispersed into the stratosphere and anvil section of the storm

  5. Calculation of particulate dispersion in a design-basis tornadic storm from the General Electric Vallecitos Nuclear Center, Vallecitos, California

    Energy Technology Data Exchange (ETDEWEB)

    Pepper, D.W.

    1979-11-01

    A three-dimensional numerical model is used to calculate ground-level air concentration and deposition (due to precipitation scavenging) after a hypothetical tornado strike at the General Electric Vallecitos Nuclear Center at Vallecitos, California. Plutonium particles less than 20 ..mu..m in diameter are assumed to be lifted into the tornadic storm cell by the vortex. The rotational characteristics of the tornadic storm are embedded within the larger mesoscale flow of the storm system. The design-basis translational wind values are based on probabilities associated with existing records of tornado strikes in the vicinity of the plant site. Turbulence exchange coefficients are based on empirical values deduced from experimental data in severe storms and from theoretical assumptions obtained from the literature. The method of moments is used to incorporate subgrid-scale resolution of the concentration within a grid cell volume. This method is a quasi-Lagrangian scheme which minimizes numerical error associated with advection. In all case studies, the effects of updrafts and downdrafts, coupled with scavenging of the particulates by precipitation, account for most of the material being deposited within 20 to 45 km downwind of the plant site. Ground-level isopleths in the x-y plane show that most of the material is deposited behind and slightly to the left of the centerline trajectory of the storm. Approximately 5% of the material is dispersed into the stratosphere and anvil section of the storm.

  6. The physics of nuclear reactors

    CERN Document Server

    Marguet, Serge

    2017-01-01

    This comprehensive volume offers readers a progressive and highly detailed introduction to the complex behavior of neutrons in general, and in the context of nuclear power generation. A compendium and handbook for nuclear engineers, a source of teaching material for academic lecturers as well as a graduate text for advanced students and other non-experts wishing to enter this field, it is based on the author’s teaching and research experience and his recognized expertise in nuclear safety. After recapping a number of points in nuclear physics, placing the theoretical notions in their historical context, the book successively reveals the latest quantitative theories concerning: •   The slowing-down of neutrons in matter •   The charged particles and electromagnetic rays •   The calculation scheme, especially the simplification hypothesis •   The concept of criticality based on chain reactions •   The theory of homogeneous and heterogeneous reactors •   The problem of self-shielding �...

  7. Nuclear reactor control rod

    International Nuclear Information System (INIS)

    Cearley, J.E.; Izzo, K.R.

    1987-01-01

    This patent describes a vertically oriented bottom entry control rod from a nuclear reactor: a frame including an elongated central spine of cruciform cross section connected between an upper support member and a lower support member both of cruciform shape having four laterally extending arms. The arms are in alignment with the arms of the lower support member and each aligned upper and lower support members has a sheath extending between; absorber plates of neutron absorber material, different from the material of the frame, one of the absorber plates is positioned within a sheath beneath each of the arms; attachment means suspends the absorber plates from the arms of the upper support member within a sheath; elongated absorber members positioned within a sheath between each of the suspended absorber plates and an arm of the lower support member; and joint means between the upper ends of the absorber members and the lower ends of the suspended absorber plates for minimizing gaps; the sheath means encloses the suspended absorber plates and the absorber members extending between aligned arms of the upper and lower support members and secured

  8. Nuclear reactor container

    International Nuclear Information System (INIS)

    Fukui, Tooru; Murase, Michio; Kataoka, Yoshiyuki; Hidaka, Masataka; Sumita, Isao; Tominaga, Kenji.

    1992-01-01

    In a nuclear reactor container, a chamber in communication with a wet well of a pressure suppression chamber is disposed and situated to such a position that the temperature is lower than a chamber containing pool water upon occurrence of loss of coolant accident. In addition, the inner surface of the pressure suppression chamber is constituted with steel walls in contact with pool water, and an outer circumferential pool is disposed at the outer circumferential surface thereof. Further, a circulation channel is disposed, and a water intake port is disposed at a position higher than an exit to the pool water, and a water discharge port is opened in the pool water at a position lower than the exit to the pool water. With such a constitution, the allowable temperature of the pressure suppression pool water can be elevated to a saturated steam temperature corresponding to the resistant pressure of the container, so that the temperature difference between the pressure suppression pool and the outer side thereof is increased by so much, to improve thermal radiation performance. Accordingly, it can be utilized as a pressure suppression means for a plant of greater power. Further, thermal conduction efficiency from the pool water region of the pressure suppression chamber to the outer circumferential pool water is improved, or thermal radiation area is enlarged due to the circulation channel, to improve the heat radiation performance. (N.H.)

  9. New materials in nuclear fusion reactors

    International Nuclear Information System (INIS)

    Iwata, Shuichi

    1988-01-01

    In the autumn of 1987, the critical condition was attained in the JET in Europe and Japanese JT-60, thus the first subject in the physical verification of nuclear fusion reactors was resolved, and the challenge to the next attainment of self ignition condition started. As the development process of nuclear fusion reactors, there are the steps of engineering, economical and social verifications after this physical verification, and in respective steps, there are the critical problems related to materials, therefore the development of new materials must be advanced. The condition of using nuclear fusion reactors is characterized by high fluence, high thermal flux and strong magnetic field, and under such extreme condition, the microscopic structures of materials change, and they behave much differently from usual case. The subjects of material development for nuclear fusion reactors, the material data base being built up, the materials for facing plasma and high thermal flux, first walls, blanket structures, electric insulators and others are described. The serious effect of irradiation and the rate of defect inducement must be taken in consideration in the structural materials for nuclear fusion reactors. (Kako, I.)

  10. Nuclear reactor vessel inspection apparatus

    International Nuclear Information System (INIS)

    Blackstone, E.G.; Lofy, R.A.; Williams, L.P.

    1979-01-01

    Apparatus for the in situ inspection of a nuclear reactor vessel to detect the location and character of flaws in the walls of the vessel, in the welds joining the various sections of the vessel, in the welds joining attachments such as nozzles, elbows and the like to the reactor vessel and in such attachments wherein an inspection head carrying one or more ultrasonic transducers follows predetermined paths in scanning the various reactor sections, welds and attachments

  11. Control rod drive of nuclear reactor

    International Nuclear Information System (INIS)

    Zhuchkov, I.I.; Gorjunov, V.S.; Zaitsev, B.I.

    1980-01-01

    This invention relates to nuclear reactors and, more particularly, to a drive of a control rod of a nuclear reactor and allows power control, excess reactivity compensation, and emergency shut-down of a reactor. (author)

  12. 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)

  13. Nuclear reactor in deep water

    International Nuclear Information System (INIS)

    Anon.

    1980-01-01

    Events during October 1980, when the Indian Point 2 nuclear reactor was flooded by almost 500 000 litres of water from the Hudson river, are traced and the jumble of human errors and equipment failures chronicled. Possible damage which could result from the reactor getting wet and from thermal shock are considered. (U.K.)

  14. Reactor coolant pump for a nuclear reactor

    International Nuclear Information System (INIS)

    Burkhardt, W.; Richter, G.

    1976-01-01

    An improvement is proposed concerning the easier disengagement of the coupling at the reactor coolant pump for a nuclear reactor transporting a pressurized coolant. According to the invention the disengaging coupling consists of two parts separated by screws. At least one of the screws contains a propellent charge ananged within a bore and provided with a speed-dependent ignition device in such a way that by separation of the screws at overspeeds the coupling is disengaged. The sub-claims are concerned with the kind of ignition ot the propellent charge. (UWI) [de

  15. Transmutation of nuclear waste in nuclear reactors

    International Nuclear Information System (INIS)

    Abrahams, K.; Kloosterman, J.L.; Pilate, S.; Wehmann, U.K.

    1996-03-01

    The objective of this joint study of ECN, Belgonucleaire, and Siemens is to investigate possibilities for transmutation of nuclear waste in regular nuclear reactors or in special transmutation devices. Studies of possibilities included the limits and technological development steps which would be needed. Burning plutonium in fast reactors, gas-cooled high-temperature reactors and light water reactors (LWR) have been considered. For minor actinides the transmutation rate mainly depends on the content of the minor actinides in the reactor and to a much less degree on the fact whether one uses a homogeneous system (with the actinides mixed into the fuel) or a heterogeneous system. If one wishes to stabilise the amount of actinides from the present LWRs, about 20% of all nuclear power would have to be generated in special burner reactors. It turned out that reactor transmutation of fission products would require considerable recycling efforts and that the time needed for a substantial transmutation would be rather long for the presently available levels of the neutron flux. If one would like to design burner systems which can serve more light water reactors, a large effort would be needed and other burners (possibly driven by accelerators) should be considered. (orig.)

  16. Nuclear power within liberalised electricity markets

    International Nuclear Information System (INIS)

    Kidd, Stephen W.

    2002-01-01

    Competition between various methods of generating electricity in liberalised markets means that all power plants must be cost-effective. The price of electricity from nuclear power includes all waste disposal and decommissioning costs, unlike other electricity generating technologies. Most existing nuclear power plants are likely to prosper under electricity liberalization. Many will receive operating life extensions and be able to compete in the electricity market for many years to come. Investment costs are particularly heavy for nuclear plants. Capital expenditure appraisal methodologies mean that such plants suffer financial disadvantages in times of high interest rates. Low and stable fuel costs are the prime advantage of nuclear plants against other sources of generating electricity. There will be significant demand for new generating capacity, both incremental and replacement, in the next 20 years. Under present conditions, where there is access to a stable and cheap supply of piped gas, nuclear and coal plants find it difficult to compete against gas-fired plants. The nuclear industry is addressing the need for new reactor designs, offering significant capital and operating cost reductions from the previous generation of reactors. This development and the need for carbon abatement on a worldwide basis offers nuclear plants a further economic advantage against alternative technologies. (author)

  17. Nuclear power reactors: reactor safety and military and civil defence

    International Nuclear Information System (INIS)

    Hvinden, T.

    1976-01-01

    The formation of fission products and plutonium in reactors is briefly described, followed by a short general discussion of reactor safety. The interaction of reactor safety and radioactive release considerations with military and civil defence is thereafter discussed. Reactors and other nuclear plants are factors which must be taken into account in the defence of the district around the site, and as potential targets of both conventional and guerilla attacks and sabotage, requiring special defence. The radiological hazards arising from serious damage to a power reactor by conventional weapons are briefly discussed, and the benefits of underground siting evaluated. Finally the author discusses the significance of the IAEA safeguards work as a preventive factor. (JIW)

  18. Evaluation of neutronic characteristics of in-pile test reactor for fast reactor safety research

    Energy Technology Data Exchange (ETDEWEB)

    Uto, N.; Ohno, S.; Kawata, N. [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1996-09-01

    An extensive research program has been carried out at the Power Reactor and Nuclear Fuel Development Corporation for the safety of future liquid-metal fast breeder reactors to be commercialized. A major part of this program is investigation and planning of advanced safety experiments conducted with a new in-pile safety test facility, which is larger and more advanced than any of the currently existing test reactors. Such a transient safety test reactor generally has unique neutronic characteristics that require various studies from the reactor physics point of view. In this paper, the outcome of the neutronics study is highlighted with presenting a reference core design concept and its performance in regard to the safety test objectives. (author)

  19. Nuclear reactors: physics and materials

    Energy Technology Data Exchange (ETDEWEB)

    Yadigaroglu, G

    2005-07-01

    In the form of a tutorial addressed to non-specialists, the article provides an introduction to nuclear reactor technology and more specifically to Light Water Reactors (LWR); it also shows where materials and chemistry problems are encountered in reactor technology. The basics of reactor physics are reviewed, as well as the various strategies in reactor design and the corresponding choices of materials (fuel, coolant, structural materials, etc.). A brief description of the various types of commercial power reactors follows. The design of LWRs is discussed in greater detail; the properties of light water as coolant and moderator are put in perspective. The physicochemical and metallurgical properties of the materials impose thermal limits that determine the performance and the maximum power a reactor can deliver. (author)

  20. Design and testing of integrated circuits for reactor protection channels

    International Nuclear Information System (INIS)

    Battle, R.E.; Vandermolen, R.I.; Jagadish, U.; Swail, B.K.; Naser, J.; Rana, I.

    1995-01-01

    Custom and semicustom application-specific integrated circuit design and testing methods are investigated for use in research and commercial nuclear reactor safety systems. The Electric Power Research Institute and Oak Ridge National Laboratory are working together through a cooperative research and development agreement to apply modern technology to a nuclear reactor protection system. Purpose of this project is to demonstrate to the nuclear industry an alternative approach for new or upgrade reactor protection and safety system signal processing and voting logic. Motivation for this project stems from (1) the difficulty of proving that software-based protection systems are adequately reliable, (2) the obsolescence of the original equipment, and (3) the improved performance of digital processing