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Sample records for japan power demonstration reactor

  1. Decommissioning experience of the Japan power demonstration reactor

    International Nuclear Information System (INIS)

    Hoshi, T.; Yanagihara, S.; Tachibana, M.; Momma, T.

    1992-01-01

    Actual dismantling of the Japan Power Demonstration Reactor (JPDR) has been progressing since 1986 aiming to make stage 3 condition as the final goal. Such highly activated components as the reactor pressure vessel (RPV) and the inner portion of biological shield concrete close to the RPV have removed using the remotely operated cutting machines. Useful data on the dismantling techniques and their safety as well as the manpower expenditure and radiation exposure of workers have been obtained. Experiences gained through the dismantling works are described in this paper. (author)

  2. The Japan Power Demonstration Reactor dismantling project. Radiation control

    International Nuclear Information System (INIS)

    Tomii, Hiroyuki; Seiki, Yoshihiro

    1996-01-01

    In the Japan Power Demonstration Reactor (JPDR) dismantling project, radiation control was performed properly with routine and special monitoring to keep the occupational safety and to collect data necessary for future dismantling of nuclear facilities. This report describes a summary of radiation control in the dismantling activities and some results of parametric analysis on dose equivalent evaluation, and introduces the following knowledge on radiological protection effectiveness of the dismantling systems applied in the project. a) Use of remote dismantling systems was effective in reducing equivalent workplace exposure. b) Utilization of existing facilities as radiation shield or radioactivity containment was effective in reducing workplace exposure, and also in increasing work efficiency. c) Use of underwater cutting systems was useful to minimize air contamination, and to reduce the dose equivalent rate in the working area. d) In the planning of dismantling, it is necessary to optimize the radiation protection by analyzing dismantling work procedures and evaluating radiological features of the dismantling systems applied, including additional work which the systems require brought from such activities. (author)

  3. Data analysis on work activities in dismantling of Japan Power Demonstration Reactor (JPDR). Contract research

    International Nuclear Information System (INIS)

    Shiraishi, Kunio; Sukegawa, Takenori; Yanagihara, Satoshi

    1998-03-01

    The safe dismantling of a retired nuclear power plant was demonstrated by completion of dismantling activities for the Japan Power Demonstration Reactor (JPDR), March, 1996, which had been conducted since 1986. This project was a flag ship project for dismantling of nuclear power plants in Japan, aiming at demonstrating an applicability of developed dismantling techniques in actual dismantling work, developing database on work activities as well as dismantling of components and structures. Various data on dismantling activities were therefore systematically collected and these were accumulated on computer files to build the decommissioning database; dismantling activities were characterized by analyzing the data. The data analysis resulted in producing general forms such as unit activity factors, for example, manpower need per unit weight of component to be dismantled, and simple arithmetic forms for forecasting of project management data to be applied to planning another dismantling project through the evaluation for general use of the analyzed data. The results of data analysis could be usefully applied to planning of future decommissioning of commercial nuclear power plants in Japan. This report describes the data collection and analysis on the JPDR dismantling activities. (author)

  4. The Japan Power Demonstration Reactor (JPDR) dismantling activities. Management of JPDR dismantling waste

    International Nuclear Information System (INIS)

    Abe, Masayoshi; Nakata, Susumu; Ito, Shinichi

    1996-01-01

    The management of wastes, both radioactive and non-radioactive, is one of the most important issues for a safe and reasonable dismantling operation of nuclear power plants. A large amount of radioactive wastes is arising from a reactor dismantling operation in a relatively short period time, ranging in a wide variety from very low level to relatively high level. Moreover non-radioactive waste is also in a huge amount. The dismantling operation of Japan Power Demonstration Reactor (JPDR) resulted in 24,440 tons of dismantling wastes, of which about 15% was radioactive and 85% non-radioactive. These wastes were managed successfully implementing a well developed management plan for JPDR dismantling waste. Research and development works for handling of JPDR dismantling wastes were performed, including fixation of loose contamination on surface, volume reduction and waste containers for on-site transportation and interim storage. The JPDR dismantling wastes generated were classified and categorized depending on their materials, characteristics and activity level. Approximately 2,100 tons of radioactive wastes were stored in the interim storage facilities on site using developed containers, and 1,670 tons of radioactive concrete waste were used for a safe demonstration test of a simple near-surface disposal for very low level waste. Other dismantling wastes such as steel and concrete which were categorized as non-radioactive were recycled and reused as useful resources. This paper describes the management of the JPDR dismantling wastes. (author)

  5. Decontamination and radioactivity measurement on building surfaces related to dismantling of Japan power demonstration reactor (JPDR)

    International Nuclear Information System (INIS)

    Hatakeyama, Mutsuo; Tachibana, Mitsuo; Yanagihara, Satoshi

    1997-12-01

    In the final stage of dismantling activities for decommissioning a nuclear power plant, building structures have to be demolished to release the site for unrestricted use. Since building structures are generally made from massive reinforced concrete materials, it is not a rational way to treat all concrete materials arising from its demolition as radioactive waste. Segregation of radioactive parts from building structures is therefore indispensable. The rational procedures were studied for demolition of building structures by treating arising waste as non-radioactive materials, based on the concept established by Nuclear Safety Commission, then these were implemented in the following way by the JPDR dismantling demonstration project. Areas of the JPDR facilities are categorized into two groups : possibly contaminated areas, and possibly non-contaminated areas, based on the document of the reactor operation. Radioactivity on the building surfaces was then measured to confirm that the qualitative categorization is reasonable. After that, building surfaces were decontaminated in such a way that the contaminated layers were removed with enough margin to separate radioactive parts from non-radioactive building structures. Thought it might be possible to demolish the building structures by treating arising waste as non-radioactive materials, confirmation survey for radioactivity was conducted to show that there is no artificial radioactive nuclides produced by operation in the facility. This report describes the procedures studied on measurement of radioactivity and decontamination, and the results of its implementation in the JPDR dismantling demonstration project. (author)

  6. Role of Fugen-HWR in Japan and design of a 600 MWe demonstration reactor

    International Nuclear Information System (INIS)

    Sawai, S.

    1982-01-01

    Fugen, a 165 MWe prototype of a heavy water moderated boiling light water cooled reactor; has been in commercial operation since March 20, 1979. In parallel with the Fugen project, the design work of the 600 MWe demonstration plant has been carried out since 1973. Important system and components, such as pressure tube assemblies, control rod drive mechanism, etc., are essentially the same as those of Fugen. Some modifications, however, are made especially from the stand point of experiences In the Fugen-HWR, plutonium and uranium would be effectively used; and plutonium could make the coolant void reactivity more negative which would give good results in increasing the reactor stability and safety. On the other hand, nuclear power plants are mainly consisted of LWRs in Japan. Considering the above situations, the Fugen-HWR, coupled with LWRs, is now considered in Japan to contribute to our energy security by using plutonium and depleted uranium extracted from spent fuels of LWRs: thereby reducing the demands On August 4, 1981, the ad hoc committee on the 600 MWe demonstration Fugen-HWR submitted the final report to the Japan AEC, after having had discussions and evaluations. In the report, the ad hoc committee recommended to build the 600 MWE demonstration plant with appropriate supports of the Government. The Japan AEC will be expected to make her decision on the program in the near future. As for the reactor safety R and C, development has been stressed on coolant leak detectors and ECCS performances or Since 1965, many development works have been done for mixed oxide fuel assemblies, both for establishment of the fabrication technology and for clarification of irradiation performances. 196 mixed oxide fuel assemblies have been manufactured for Fugen. 168 of them were loaded and 92 were withdrawn. No fuel has been failured yet. (author)

  7. Current status of nuclear power generation in Japan and directions in water cooled reactor technology development

    International Nuclear Information System (INIS)

    Miwa, T.

    1991-01-01

    Electric power demand aspects and current status of nuclear power generation in Japan are outlined. Although the future plan for nuclear power generation has not been determined yet the Japanese nuclear research centers and institutes are investigating and developing some projects on the next generation of light water reactors and other types of reactors. The paper describes these main activities

  8. Evolution of the Fusion Power Demonstration tandem mirror reactor configuration

    International Nuclear Information System (INIS)

    O'Toole, J.A.; Lousteau, D.C.

    1985-01-01

    This paper gives a presentation of the evolution of configurations proposed for tandem mirror Fusion Power Demonstration (FPD) machines. The FPD study was undertaken to scope the mission as well as the technical and design requirements of the next tandem mirror device. Three configurations, entitled FPD I, II, and III were studied. During this process new systems were conceived and integrated into the design, resulting in a significantly changed overall machine configuration. The machine can be divided into two areas. A new center cell configuration, minimizing magnetic field ripple and thus maximizing center cell fusion power, features a semicontinuous solenoid. A new end cell has evolved which maintains the required thermal barrier in a significantly reduced axial length. The reduced end cell effective length leads to a shorter central cell length being required to obtain minimum ignition conditions. Introduced is the concept of an electron mantle stabilized octopole arrangement. The engineering features of the new end cell and maintenance concepts developed are influenced to a great extent by the octopole-based design. The new ideas introduced during the FPD study have brought forth a new perspective of the size, design, and maintenance of tandem mirror reactors, making them more attractive as commercial power sources

  9. Demonstration of leak-before-break in Japan Sodium cooled Fast Reactor (JSFR) pipes

    International Nuclear Information System (INIS)

    Wakai, Takashi; Machida, Hideo; Yoshida, Shinji; Xu, Yang; Tsukimori, Kazuyuki

    2014-01-01

    This paper describes the leak-before-break (LBB) assessment procedure applicable to Japan Sodium cooled Fast Reactor (JSFR) pipes made of modified 9Cr–1Mo steel. For the sodium pipes of JSFR, the continuous leak monitoring will be adopted as an alternative to a volumetric test of the weld joints under conditions that satisfy LBB. Firstly, a LBB assessment flowchart eliminating uncertainty resulted from small scale leakage, such as self plugging phenomenon and influence of crack surface roughness on leak rate, was proposed. Secondly, a rational unstable fracture assessment technique, taking the compliance changing with crack extension into account, was also proposed. Thirdly, a crack opening displacement (COD) assessment technique was developed, because COD assessment method applicable to JSFR pipes – thin wall and small work hardening material – had not been proposed yet. In addition, fracture toughness tests were performed using compact tension (CT) specimens to obtain the fracture toughness, J IC , and the crack growth resistance (J–R) curve at elevated temperature. Finally, by using the flowchart, proposed techniques and collected data, LBB assessment for the primary sodium pipes of JSFR was conducted. As a result, LBB aspect was successfully demonstrated with sufficient margins

  10. Conceptual design of a demonstration reactor for electric power generation

    International Nuclear Information System (INIS)

    Asaoka, Y.; Hiwatari, R.; Okano, K.; Ogawa, Y.; Ise, H.; Nomoto, Y.; Kuroda, T.; Mori, S.; Shinya, K.

    2005-01-01

    Conceptual study on a demonstration plant for electric power generation, named Demo-CREST, was conducted based on the consideration that a demo-plant should have capacities both (1) to demonstrate electric power generation in a plant scale with moderate plasma performance, which will be achieved in the early stage of the ITER operation, and foreseeable technologies and materials and (2) to have a possibility to show an economical competitiveness with advanced plasma performance and high performance blanket systems. The plasma core was optimized to be a minimum size for both net electric power generation with the ITER basic plasma parameters and commercial-scale generation with advance plasma parameters, which would be attained by the end of ITER operation. The engineering concept, especially the breeding blanket structure and its maintenance scheme, is also optimized to demonstrate the tritium self-sustainability and maintainability of in-vessel components. Within the plasma performance as planned in the present ITER program, the net electric power from 0 MW to 500 MW is possible with the basic blanket system under the engineering conditions of maximum magnetic field 16 T, NBI system efficiency 50%, and NBI current drive power restricted to 200 MW. Capacities of stabilization of reversed shear plasma and the high thermal efficiency are additional factors for optimization of the advanced blanket. By replacing the blanket system with the advanced one of higher thermal efficiency, the net electric power of about 1000 MW is also possible so that the economic performance toward the commercial plant can be also examined with Demo-CREST. (author)

  11. Power-up of Fugen reactor and development of demonstration plant

    International Nuclear Information System (INIS)

    Sawai, Sadamu; Akebi, Michio; Yazaki, Akira.

    1979-06-01

    The Fugen Nuclear Power Station is the 165 MWe prototype plant characterized by heavy water-moderated, boiling light water-cooled, pressure tube type, and was developed by the Power Reactor and Nuclear Fuel Development Corporation, Japan. The plant went into commercial operation on March 20, 1979, in Tsuruga, Fukui Prefecture. Some delay in the overall schedule occurred due to the shortage of cement caused by the oil crisis, more stringent regulations as the result of stress corrosion cracking experienced in BWRs, and design modifications. All functional tests, the final check-up of the whole plant, and remaining modifying works had been completed by March 10, 1978. The minimum criticality was achieved with 22 mixed oxide fuel assemblies on March 20, 1978. Thereafter, the tests on reactor physics, plant dynamics, the performances of components and systems, and radiation and water chemistry have been carried out. 5 MWe was sent to grid system for the first time on July 29, 1978. The commercial operation of the plant was licenced by the Government on March 30, 1979. The conceptual design of the 600 MWe demonstration plant was finished in early 1979, and the detailed design is to be carried out in 1979 and 1980. The main design principle was incorporated in the conceptual design, but some modifications are to be made to reduce the power cost and to facilitate the easy maintenance. (Kako, I.)

  12. Role of Fugen HWR in Japan and design of a 600 MWe demonstration reactor

    International Nuclear Information System (INIS)

    Sawai, Sadamu.

    1982-03-01

    Fugen, a 165 MWe prototype of a heavy water-moderated, boiling light water-cooled reactor, has been in commercial operation since March 20, 1979. In parallel with the Fugen project, the design work for a 600 MWe demonstration plant has been carried out since 1973. The important systems and components, such as pressure tube assemblies and control rod drive mechanism, are essentially the same as those of Fugen. However, some modification is made owing to the experience obtained in Fugen and LWrs. In the HWR Fugen, plutonium and uranium are effectively used, and plutonium makes the coolant void reactivity more negative, which results in the increase of the stability and safety of the reactor. On August 4, 1981, the ad hoc committee submitted the final report to the Japanese Atomic Energy Commission, in which the construction of a 600 MWe demonstration plant was recommended. As for the research and development on reactor safety, coolant leak detectors, the performance of ECCS, and safety design codes are enumerated. Since 1965, mixed oxide fuel has been developed, and 168 fuel assemblies were loaded in Fugen, but failure did not occur. (Kako, I.)

  13. Problems and legislative remedies of the parallel law systems in Japan for nuclear power reactors

    International Nuclear Information System (INIS)

    Irie, Kazutomo

    2011-01-01

    There are two established laws governing nuclear power reactors in Japan. One is the Electricity Utilities Industry Law, which regulates the nuclear power reactors, and the other is the so-called 'Reactor Regulation Law', which dually regulates the reactors in some phases. When a graded approach on the regulation of nuclear reactors was adopted, it extended over these two laws and was legislatively imperfect. Such imperfection created problems from the beginning. Also, the original regulatory structures presented by these laws had become obscure during the operation process of the graded regulation. The situation becomes further complicated by the revision of these laws in recent years. It appears that the trait of the regulatory procedural structure of the Electricity Utilities Industry Law has been weakened. As there is a pressing need to review the entire regulatory structure and to propose a unified regulatory system by combining these laws, this paper examines the merits and demerits of combining these laws under a unified regulation. (author)

  14. Data list of nuclear power plants of pressurized-water reactor type in Japan

    International Nuclear Information System (INIS)

    Izumi, Fumio; Harayama, Yasuo

    1981-08-01

    This report has collected and compiled the data concerning performances, equipments and installations for nuclear power plants of the pressurized-water reactor type in Japan. The data used in the report are based on informations that were collected before December in 1980. The report is edited by modifing changes of the data appeared after publication of 1979 edition (JAERI-M 8947), and extending the data-package to cover new plants proposed thereafter. All data have been processed and tabulated with a computer program FREP, which has been developed as an exclusive use of data processing. (author)

  15. The current uranium exploration activities of the Power Reactor and Nuclear Fuel Development Corporation (PNC), Japan

    International Nuclear Information System (INIS)

    Miyada, H.

    2001-01-01

    As of November 1996, Japan's total installed commercial nuclear power generation capacity was 42 GW(e), accounting for 34% of total electric energy generation. By 2010, Japan intends to have an installed electricity generation capacity of 70.5 GW(e). This will increase the country's demand for nat Ural uranium from 7,700 t U in 1994 (13% of the world consumption) to 13,800 t U in 2010 (17%-19% of the world projected consumption). However, Japan's known uranium resources at Ningyo-Toge and Tono deposits, are estimated at roughly only 6,600 t U. The Long-term Programme for Research, Development and Utilization of Nuclear Energy (adopted in 1994) calls for diversification through long-term purchasing contracts, independent exploration and involvement in mining vent Ures, with the objective of ensuring independence and stability in Japan's development and utilization of nuclear energy. The Power Reactor and Nuclear Fuel Development Corporation (PNC) has been commissioned to carry out the task of independent exploration. PNC is carrying out exploration projects in Canada, Australia, USA and China targeting unconformity related type deposits with an eye to privatizing them. Currently about 40,000 t U of uranium resources are held by PNC. PNC has been carrying out the following related activities: (1) Reference surveys on uranium resources to delineate the promising areas; (2) Development of uranium exploration technology; (3) Information surveys on the nuclear industries to project long-term supply and demand; (4) International Cooperation programme on uranium exploration with Asian countries. (author)

  16. Decommissioning and decontamination of licensed reactor facilities and demonstration nuclear power plants

    International Nuclear Information System (INIS)

    Lear, G.; Erickson, P.B.

    1975-01-01

    Decommissioning of licensed reactors and demonstration nuclear power plants has been accomplished by mothballing (protective storage), entombment, and dismantling or a combination of these three. The alternative selected by a licensee seems to be primarily based on cost. A licensee must, however, show that the decommissioning process provides adequate protection of the health and safety of the public and no adverse impact on the environment. To date the NRC has approved each of the alternatives in the decommissioning of different facilities. The decommissioning of small research reactors has been accomplished primarily by dismantling. Licensed nuclear power plants, however, have been decommissioned primarily by being placed in a mothballed state in which they continue to retain a reactor license and the associated licensee responsibilities

  17. Fusion power demonstration - a baseline for the mirror engineering test reactor

    International Nuclear Information System (INIS)

    Henning, C.D.; Logan, B.G.; Neef, W.S.

    1983-01-01

    Developing a definition of an engineering test reactor (ETR) is a current goal of the Office of Fusion Energy (OFE). As a baseline for the mirror ETR, the Fusion Power Demonstration (FPD) concept has been pursued at Lawrence Livermore National Laboratory (LLNL) in cooperation with Grumman Aerospace, TRW, and the Idaho National Engineering Laboratory. Envisioned as an intermediate step to fusion power applications, the FPD would achieve DT ignition in the central cell, after which blankets and power conversion would be added to produce net power. To achieve ignition, a minimum central cell length of 67.5 m is needed to supply the ion and alpha particles radial drift pumping losses in the transition region. The resulting fusion power is 360 MW. Low electron-cyclotron heating power of 12 MW, ion-cyclotron heating of 2.5 MW, and a sloshing ion beam power of 1.0 MW result in a net plasma Q of 22. A primary technological challenge is the 24-T, 45-cm bore choke coil, comprising a copper hybrid insert within a 15 to 18 T superconducting coil

  18. Datafile: [nuclear power in] Japan

    International Nuclear Information System (INIS)

    Anon.

    1989-01-01

    Japan is third after the USA and France in terms of the Western World's installed nuclear capacity, but it has by far the largest forward programme. Great effort is also being put into the fuel cycle and advanced reactors. There is close co-operation between the government, utilities and manufacturers, but Japan has not sought to export reactors. The government has responded to the growing public opposition to nuclear power with a massive increase in its budget for public relations. Details of the nuclear power programme are given. (author)

  19. Database structure and file layout of Nuclear Power Plant Database. Database for design information on Light Water Reactors in Japan

    International Nuclear Information System (INIS)

    Yamamoto, Nobuo; Izumi, Fumio.

    1995-12-01

    The Nuclear Power Plant Database (PPD) has been developed at the Japan Atomic Energy Research Institute (JAERI) to provide plant design information on domestic Light Water Reactors (LWRs) to be used for nuclear safety research and so forth. This database can run on the main frame computer in the JAERI Tokai Establishment. The PPD contains the information on the plant design concepts, the numbers, capacities, materials, structures and types of equipment and components, etc, based on the safety analysis reports of the domestic LWRs. This report describes the details of the PPD focusing on the database structure and layout of data files so that the users can utilize it efficiently. (author)

  20. Fusion-power demonstration

    International Nuclear Information System (INIS)

    Henning, C.D.; Logan, B.G.; Carlson, G.A.; Neef, W.S.; Moir, R.W.; Campbell, R.B.; Botwin, R.; Clarkson, I.R.; Carpenter, T.J.

    1983-01-01

    As a satellite to the MARS (Mirror Advanced Reactor Study) a smaller, near-term device has been scoped, called the FPD (Fusion Power Demonstration). Envisioned as the next logical step toward a power reactor, it would advance the mirror fusion program beyond MFTF-B and provide an intermediate step toward commercial fusion power. Breakeven net electric power capability would be the goal such that no net utility power would be required to sustain the operation. A phased implementation is envisioned, with a deuterium checkout first to verify the plasma systems before significant neutron activation has occurred. Major tritium-related facilities would be installed with the second phase to produce sufficient fusion power to supply the recirculating power to maintain the neutral beams, ECRH, magnets and other auxiliary equipment

  1. Fusion power demonstration

    International Nuclear Information System (INIS)

    Henning, C.D.; Logan, B.G.

    1983-01-01

    As a satellite to the MARS (Mirror Advanced Reactor Study) a smaller, near-term device has been scoped, called the FPD (Fusion Power Demonstration). Envisioned as the next logical step toward a power reactor, it would advance the mirror fusion program beyond MFTF-B and provide an intermediate step toward commercial fusion power. Breakeven net electric power capability would be the goal such that no net utility power would be required to sustain the operation. A phased implementation is envisioned, with a deuterium checkout first to verify the plasma systems before significant neutron activation has occurred. Major tritium-related facilities would be installed with the second phase to produce sufficient fusion power to supply the recirculating power to maintain the neutral beams, ECRH, magnets and other auxiliary equipment

  2. Reduction of releases of radioactive effluents from light-water-power-reactors in Japan

    International Nuclear Information System (INIS)

    Yoshida, Y.; Itakura, T.; Kanai, T.

    1977-01-01

    Japan Atomic Energy Commission established the dose objectives to the population around the light-water-reactors in May, 1975, based on the ''ALAP'' concept. These values are respectively, 5 mrems per year for total body and 15 mrems per year for thyroid of an individual in the critical group in the environs, due to both gaseous and liquid effluents from LWRs in one site. The present paper describes the implications of the dose objective values, control measures which have been adopted to reduce releases of radioactive materials and related technical developments in Japan. The main control measures for reduction of radioactive gaseous effluents are an installation of a charcoal gas holdup system for decay of noble gases and a supply of clean steam for the gland seal of a turbine in BWR, and a storage tank system allowing decay of noble gases in PWR. For liquid effluents are taken measures to re-use them as the primary coolant. Consequently, the amounts of radioactivity released to the environment from any LWR during normal operation have been maintained under the level to meet the above dose objective values. For research reactors, reduction of release of effluents has also been carried out in a similar way to LWRs. In order to establish the techniques applicable for further reduction, studies are being made on the control measures to reduce leakage of radioiodine, an apparatus for removal of krypton, the treatment of laundry waste and measures to remove the crud in the primary coolant. Presentation is also made on the energy-integrated gas monitor for gaseous effluent and systems of measuring γ dose from radioactive cloud descriminating from natural background, which have been developed for effective monitoring thus reduced environmental dose

  3. Nuclear power development in Japan

    International Nuclear Information System (INIS)

    Sugawara, A.

    1994-01-01

    The energy situation in Japan is briefly outlined. Vulnerability in energy structure of the country is shown by a comparison of primary energy supply patterns of Japan and Western countries. Japan's energy policy consists in reducing dependence on oil, promoting efficient use of energy and increasing use of non-fossil fuels. Nuclear power is a core of alternative energy for petroleum because of stable supply of nuclear fuel, low detrimental emissions and less dependence on the fuel. A short historical review of nuclear power development in Japan is presented. Some future issues as development of entire nuclear fuel cycle, social acceptance, reactor safety and nuclear power economics are also discussed. 6 figs. (R.T.)

  4. A review of fast reactor programme in Japan

    International Nuclear Information System (INIS)

    1998-01-01

    This report describes the development and activities on fast reactor in Japan for the period of April 1996 - March 1997. During this period, the 30th duty cycle operation has been started in the Experimental Fast Reactor ''''Joyo''''. The cause investigation on the sodium leak incident has completed and the safety examination are being performed in the Prototype Fast Breeder Reactor ''''Monju''''. The three years design study since FY1994 on the plant optimization of the Demonstration FBR has been completed by the Japan Atomic Power Company (JAPC). Related research and development works are underway at several organizations under the discussion and coordination of the Japanese FBR R and D Steering Committee, which is composed of Power Reactor and Nuclear Fuel Development Corporation (PNC), JAPC, Japan Atomic Energy Research Institute (JAERI) and Central Research Institute of Electric Power Industry (CRIEPI). In November 1996, the Japan Atomic Energy Commission (JAEC) established a Social Gathering Meeting to discuss generally the significance of FBR development in Japan for the future. (author)

  5. 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

  6. The development of breeder reactors in Japan

    International Nuclear Information System (INIS)

    Segawa, M.

    1984-01-01

    In the framework of a global analysis of the various available sources of energy, Japan has reserved a prominent place to the nuclear energy and, in the long-term view, to the breeder reactor which will be due for commercial deployment in 20)10. To achieve these objectives, three stages are envisaged, one of the experimental reactor Joyo (in service), one of the demonstration reactor Monju (its construction has been decided), and one of the pre-commercial reactor (due to be taken in hand at the beginning of the Nineties). Efforts will be made in parallel concerning the fuel cycle [fr

  7. Demonstration tests of tritium removal device under the conditions of nuclear fusion reactor. Cooperation test between Japan and USA

    International Nuclear Information System (INIS)

    Hayashi, Takumi; Kobayashi, Kazuhiro; Nishi, Masataka

    2001-01-01

    Performance of oxidation catalysis in emergency tritium removal device was tested in Los Alamos National Laboratory by cooperation between Japan and USA on November 8, 2000. To reduce the effects of tritium on the environment, a plan of the closed space for trapping tritium was made. A tritium removal device using oxidation catalysis and water vapor adsorption removes the tritium in the closed space. The treatment flow rate of the device is about 2,500 m 3 /h, the same as ITER(3,000 to 4,500 m 3 /h). Catalysis is Pt/ alumina. The closed space is 3,000m 2 . The initial concentration of tritium was about 7 Bq/cm 2 , ten times as large as the concentration limit in atmosphere. The concentration of tritium in the test laboratory decreased linearly with time and attained to the limit value after about 200 min. Residue of tritium on the wall had been removed and the significant quantity was not detected after three days. The results proved to satisfy safety of ITER. (S.Y.)

  8. Demonstration tokamak power plant

    International Nuclear Information System (INIS)

    Abdou, M.; Baker, C.; Brooks, J.; Ehst, D.; Mattas, R.; Smith, D.L.; DeFreece, D.; Morgan, G.D.; Trachsel, C.

    1983-01-01

    A conceptual design for a tokamak demonstration power plant (DEMO) was developed. A large part of the study focused on examining the key issues and identifying the R and D needs for: (1) current drive for steady-state operation, (2) impurity control and exhaust, (3) tritium breeding blanket, and (4) reactor configuration and maintenance. Impurity control and exhaust will not be covered in this paper but is discussed in another paper in these proceedings, entitled Key Issues of FED/INTOR Impurity Control System

  9. Project WAGR: the UK demonstration project for power reactor decommissioning - a review of the tools used to dismantle the reactor core

    International Nuclear Information System (INIS)

    Benest, T.G.

    2008-01-01

    The United Kingdom Atomic Energy Authority (UKAEA) has built and operated a wide range of nuclear facilities since the late 1940. UKAEA mission is to restore the environment of its sites in a safe and secure manner. This restoration includes the decommissioning of a number of redundant research and power reactors. The Windscale Advanced Gas-cooled Reactor (WAGR) was the UK prototype Advanced gas cooled reactor and became the forerunner of a family of 14 reactors built to generate cheaper and more efficient electricity in the UK. WAGR was constructed between 1957 and 1961 and was a carbon dioxide cooled, graphite moderated reactor using uranium oxide fuel in stainless steel cans. The reactor consisted of a graphite moderator housed in a cylindrical reactor vessel with hemispherical ends. The reactor and associated heat exchangers were enclosed in the iconic spherical containment building regularly used by the media in the UK as an illustration of the nuclear industry. The reactor first produced power in August 1962 and achieved full design output in 1963. It operated at an electrical output of 33 MW (E) for 18 years (average load factor of 75%). In 1981 the reactor was shut down after satisfactory completion of all the research and development objectives. In anticipation of the UK likely nuclear decommissioning needs the UKAEA decided to decommission WAGR to the International Atomic Energy Agency (IAEA) stage 3 (restoration of the area occupied by the facility to a condition of unrestricted re-usability) as the national demonstration exercise for power reactor decommissioning. Since 1998 the UKAEA and its contractors have been undertaking the dismantling of the reactor core components and pressure vessel in a series of 10 campaigns. These contain neutron activated components expected to produce dose rates well in excess of 1 Sv/hr. To carry out the work UKAEA installed an 8M remote dismantling machine (RDM) a waste recovery and transport system and a shielded waste

  10. Performance demonstration of a high-power space-reactor heat-pipe design

    International Nuclear Information System (INIS)

    Merrigan, M.A.; Martinez, E.H.; Keddy, E.S.; Runyan, J.; Kemme, J.E.

    1983-01-01

    Performance of a 15.9-mm diam, 2-m long, artery heat pipe has been demonstrated at power levels to 22.6 kW and temperatures to 1500 0 K. The heat pipe employed lithium as a working fluid with distribution wicks and arteries fabricated from 400 mesh Mo-41 wt % Re screen. Molybdenum alloy (TZM) was used for the container. Peak axial power density attained in the testing was 19 kW/cm 2 at 1465 0 K. The corresponding radial flux density in the evaporator region of the heat pipe was 150 W/cm 2 . The extrapolated limit for the heat pipe at its 1500 0 K design point is 30 kW, corresponding to an axial flux density of 25 kW/cm 2 . Sonic and capillary limits for the design were investigated in the 1100 to 1500 0 K temperature range. Excellent agreement of measured and predicted temperature and power levels was observed

  11. 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.)

  12. Supercritical CO2 Brayton power cycles for DEMO (demonstration power plant) fusion reactor based on dual coolant lithium lead blanket

    International Nuclear Information System (INIS)

    Linares, José Ignacio; Cantizano, Alexis; Moratilla, Beatriz Yolanda; Martín-Palacios, Víctor; Batet, Lluis

    2016-01-01

    This paper presents an exploratory analysis of the suitability of supercritical CO 2 Brayton power cycles as alternative energy conversion systems for a future fusion reactor based on a DCLL (dual coolant lithium-lead) blanket, as prescribed by EUROfusion. The main issue dealt is the optimization of the integration of the different thermal sources with the power cycle in order to achieve the highest electricity production. The analysis includes the assessment of the pumping consumption in the heating and cooling loops, taking into account additional considerations as control issues and integration of thermal energy storage systems. An exergy analysis has been performed in order to understand the behavior of each layout. Up to ten scenarios have been analyzed assessing different locations for thermal sources heat exchangers. Neglecting the worst four scenarios, it is observed less than 2% of variation among the other six ones. One of the best six scenarios clearly stands out over the others due to the location of the thermal sources in a unique island, being this scenario compatible with the control criteria. In this proposal 34.6% of electric efficiency (before the self-consumptions of the reactor but including pumping consumptions and generator efficiency) is achieved. - Highlights: • Supercritical CO 2 Brayton cycles have been proposed for BoP of DCLL fusion reactor. • Integration of different available thermal sources has been analyzed considering ten scenarios. • Neglecting the four worst scenarios the electricity production varies less than 2%. • Control and energy storage integration issues have been considered in the analysis. • Discarding the vacuum vessel and joining the other sources in an island is proposed.

  13. Reactor construction programme in Japan

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1959-04-15

    In September last year, the Japanese Government requested the International Atomic Energy Agency to supply three tons of natural uranium for a research reactor, and the Agency has now arranged for its sale to Japan. The metal will be supplied in ingot form and after fabrication it will be used as fuel in a reactor of the natural uranium, heavy water type. The uranium will be obtained from Canada and sold to Japan by IAEA. The Agency had invited tenders for its supply, and after considering the tenders received, the Agency's Board of Governors decided that the Canadian offer to the Agency of three tons of natural uranium free of charge should be accepted and that the selling price to Japan should be US $35. 50 per kilogramme. The price takes into account Article XIV/E of the Agency's Statute which says that the Agency shall establish a scale of charges (including those for storage and handling) for materials furnished to Member States, and that the scale shall be designed to produce revenues to meet expenses in connexion with materials acquired by the Agency and costs of materials and services provided by it under agreements with one or more members. This is the first operation of its kind to be undertaken by the Agency, and the reactor for which the supply is being made will be the first in Japan to be constructed by Japanese scientists and technicians. IAEA's Board of Governors has given the necessary approval to the reactor project for which the Agency is providing assistance

  14. Nuclear power in Japan in 1987

    International Nuclear Information System (INIS)

    Molodtsov, S.D.

    1989-01-01

    Data on the development level of nuclear power in Japan as of 1988 beginning are presented. Total registed electric power of 36 nuclear power units under operation constituted 28046 MW. 13 power units with 12268 MW total power are under construction. In 1987 188.4 TWH electric power was generated at the Japanese NPPs, it constituted 31.7% of total electric power generation. About 360 bil. yens were assigned from the state budget to further development of nuclear power engineering. Efforts to create the improved BWR type reactor, as well as, scientific and research efforts on the development of fast breeder reactors, improvement of uranium enrichment and radioactive waste storage are carried out. It is expected that share of nuclear power in electric power generation in Japan will reach 40% to the beginning of the 21-th century

  15. A review of fast reactor program in Japan

    International Nuclear Information System (INIS)

    1996-01-01

    The main R and D results of Japanese activities are summarized as follows: (1) the experimental 140 MW(th) sodium cooled fast reactor 'Joyo' provided abundant experimental data and excellent operational records, attaining more than 50,000 hours of operation since its first criticality in 1977; (2) the prototype 280 MW(e) fast reactor 'Monju' reached initial criticality on 5 April 1994; presently Monju is under the cold shutdown state because of secondary sodium leak on 8 December 1995, and multiple cause investigations of the sodium leak are being performed; (3) the Japan Atomic Power Company is promoting design studies for demonstration fast reactor (DFBR) with a power output of 600 MW(e) and R and D for DFBR are being conducted under the cooperation of governmental and private sectors. (author)

  16. Project WAGR: The UK demonstration project for power reactor decommissioning - removing the core and looking to completion

    International Nuclear Information System (INIS)

    Benest, T. G.

    2003-01-01

    The United Kingdom Atomic Energy Authority (UKAEA) has built and operated a wide range of nuclear facilities since the late 1940's. UKAEA's present mission is to restore the environment of these facilities in a safe and environmentally responsible manner. This restoration includes the decommissioning of a number of redundant research and power reactors, one of which is the Windscale Advanced Gas-cooled Reactor (WAGR). Following shut down, UKAEA decided to continue the prototype function of the reactor into the decommissioning phase to develop dismantling techniques and establish waste routes. The reactor core and pressure vessel are now being dismantled in a programme of 10 campaigns, seven of which have been completed since 1998. It is anticipated that the current programme will be completed by summer 2005. This paper outlines the history of the reactor, the operation of the waste-processing route, the installed dismantling equipment and the successful completion of the first seven campaigns. This earlier work has been described in a number of publications and conferences, so this paper concentrates on recent work to select and develop cutting equipment to dismantle the core support structures and the pressure vessel. The decommissioning of the Windscale Advance Gas-cooled reactor is being undertaken to demonstrate that a power reactor can be decommissioned shortly after shutdown. The removal of the core and pressure vessel has been broken down into a series of 10 campaigns associated with particular core components. The first 7 campaigns have been successfully completed and the 8., is expected to commence in September 2003 17 months earlier than planned. Dismantling methodologies and tools have been developed specifically for each of these campaigns. Full-scale mock-ups have been used to test the tools, train the operators and assess the duration of operations. However, despite successful trials, operational experience has shown that some of these tools have not

  17. Nuclear power development in Japan

    International Nuclear Information System (INIS)

    Sugawara, A.

    1994-01-01

    Energy situation in Japan and Japan's strategy for stable supply of energy are discussed. Benefits of nuclear power in comparison with other energy sources is considered. History of nuclear power development in Japan, modern status and future trends are described. 6 figs

  18. Development of radiation resistant structural materials utilizing fission research reactors in Japan (Role of research reactors)

    International Nuclear Information System (INIS)

    Shikama, T.; Tanigawa, H.; Nozawa, T.; Muroga, T.; Aoyama, T.; Kawamura, H.; Ishihara, M.; Ito, C.; Kaneda, S.; Mimura, S.

    2009-01-01

    Structural materials for next-generation nuclear power systems should have a good radiation resistance, where the expected accumulation dose will largely exceed 10 dpa. Among several candidate materials, materials of five categories, 1. Austenitic steels, including high nickel alloys, 2. Low activation ferritic martensitic steels, 3. ODS steels (austenitic and ferritic), 4. Vanadium based alloys, 5. Silicon carbide composites (SiC/SiCf). All have been most extensively studied in Japan, in collaboration among industries, national institutes such as Japan Atomic Energy Agency (JAEA), National Institute for Fusion Science (NIFS) and National Institute for Materials Science (NIMS), and universities. The high nickel base alloys were studied for their low swelling behaviors mainly by the NIMS and the austenitic steels are studied for their reliable engineering data base and their reliable performance in irradiation environments mainly by the JAEA, mainly for their application in the near-term projects such as the ITER and the Sodium Cooled Fast Reactors. The most extensive studies are now concentrated on the Low Activation Ferritic Marsensitic steels and ODS steels, for their application in a demonstration fusion reactor and prototype sodium cooled fast reactors. Fundamental studies on radiation effects are carried out, mainly utilizing Japan Materials Testing Rector (JMTR) with its flexible irradiation ability, up to a few dpa. For higher dpa irradiation, a fast test reactor, JOYO is utilized up to several 10s dpa. Some international collaborations such as Japan/USA and Japan/France are effective to utilize reactors abroad, such as High Flux Isotope Reactor (HFIR) of Oak Ridge National Laboratory, and sodium cooled high flux fast reactors in France. Silicon carbide based composites are extensively studied by university groups led by Kyoto University and the JAEA. For their performance in heavy irradiation environments, the Japan/USA collaboration plays an important role

  19. A review of fast reactor programme in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Masuno, Y [Experimental Fast Reactor Division, O-arai Engineering Center, PNC (Japan); Bando, S [Project Planning and Management Division, PNC, Minato-ku, Tokyo (Japan)

    1981-05-01

    The fast breeder reactor development project in Japan has been in progress in the past twelve months and will be continued in the next fiscal year, from April 1981 through March 1982, at a similar scale of effort both in budget and personnel to those of the fiscal year of 1980. The 1981 year budget for P and D work and for construction of a prototype fast breeder reactor, Monju, will be approximately 20 and 27 billion Yen respectively, excluding wages of the personnel of the Power Reactor and Nuclear Fuel Development Corporation, PNC. The number of the technical people currently engaging in the fast breeder reactor development in the PNC is approximately 530, excluding those working for plutonium fuel fabrication. Concerning the experimental fast reactor, Joyo, power increase from 50 MWt to 75 MWt was made in July 1979 and three operational cycles at 75 MWt have been completed in August 1980 and the forth cycle has started in the middle of March 1981. With respect to the prototype reactor Monju, progress toward construction has been made and an environmental impact statement of the reactor was approved by the concerned authorities. Preliminary design studies of large LMFBR are being made by PNC and also by utilities. A design study being conducted by PNC is on a 1000 MW e plant of loop type by extrapolating the technology to be developed by the time of commissioning of Monju. A group of utilities is conducting a similar study, but covering somewhat wider range of parameters and options of design. Close contact between the group and PNC has been kept. In the future, those design efforts will be combined as a single design effort, when a major effort for developing a large demonstration reactor will be initiated at around the commencement of construction of the prototype reactor Monju. Highlights and topics of the fast breeder reactor development activities in the past twelve months are summarized in this report.

  20. A review of fast reactor programme in Japan

    International Nuclear Information System (INIS)

    Masuno, Y.; Bando, S.

    1981-01-01

    The fast breeder reactor development project in Japan has been in progress in the past twelve months and will be continued in the next fiscal year, from April 1981 through March 1982, at a similar scale of effort both in budget and personnel to those of the fiscal year of 1980. The 1981 year budget for P and D work and for construction of a prototype fast breeder reactor, Monju, will be approximately 20 and 27 billion Yen respectively, excluding wages of the personnel of the Power Reactor and Nuclear Fuel Development Corporation, PNC. The number of the technical people currently engaging in the fast breeder reactor development in the PNC is approximately 530, excluding those working for plutonium fuel fabrication. Concerning the experimental fast reactor, Joyo, power increase from 50 MWt to 75 MWt was made in July 1979 and three operational cycles at 75 MWt have been completed in August 1980 and the forth cycle has started in the middle of March 1981. With respect to the prototype reactor Monju, progress toward construction has been made and an environmental impact statement of the reactor was approved by the concerned authorities. Preliminary design studies of large LMFBR are being made by PNC and also by utilities. A design study being conducted by PNC is on a 1000 MW e plant of loop type by extrapolating the technology to be developed by the time of commissioning of Monju. A group of utilities is conducting a similar study, but covering somewhat wider range of parameters and options of design. Close contact between the group and PNC has been kept. In the future, those design efforts will be combined as a single design effort, when a major effort for developing a large demonstration reactor will be initiated at around the commencement of construction of the prototype reactor Monju. Highlights and topics of the fast breeder reactor development activities in the past twelve months are summarized in this report

  1. Status of fusion reactor concept development in Japan

    International Nuclear Information System (INIS)

    Tsuji-Iio, Shunji

    1996-01-01

    Fusion power reactor studies in Japan based on magnetic confinement schemes are reviewed. As D-T fusion reactors, a steady-state tokamak reactor (SSTR) was proposed and extensively studied at the Japan Atomic Energy Research Institute (JAERI) and an inductively operated day-long tokamak reactor (IDLT) was proposed by a group at the University of Tokyo. The concept of a drastically easy maintenance (DREAM) tokamak reactor is being developed at JAERI. A high-field tokamak reactor with force-balanced coils as a volumetric neutron source is being studied by our group at Tokyo Institute of Technology. The conceptual design of a force-free helical reactor (FFHR) is under way at the National Institute for Fusion Science. A design study of a D- 3 He field-reversed configuration (FRC) fusion reactor called ARTEMIS was conducted by the FRC fusion working group of research committee of lunar base an lunar resources. (author)

  2. Conceptual design study for the demonstration reactor of JSFR. (1) Current status of JSFR development

    International Nuclear Information System (INIS)

    Hayafune, Hiroki; Sakamoto, Yoshihiko; Kotake, Shoji; Aoto, Kazumi; Ohshima, Jun; Ito, Takaya

    2011-01-01

    JAEA is now conducting 'Fast Reactor Cycle Technology Development (FaCT)' project for the commercialization before 2050s. A demonstration reactor of Japan Sodium-cooled Fast Reactor (JSFR) is planned to start operation around 2025. In the FaCT project, conceptual design study on the demonstration reactor has been performed since 2007 to determine the referential reactor specifications for the next stage design work from 2011 for the licensing and construction. Plant performance as a demonstration reactor for the 1.5 GWe commercial reactor JSFR is being compared between 750 MWe and 500 MWe plant designs. By using the results of conceptual design study, output power will be determined during year of 2010. This paper describes development status of key technologies and comparison between 750 MWe and 500 MWe plants with the view points of demonstration ability for commercial JSFR plant. (author)

  3. A review of fast reactor program in Japan

    International Nuclear Information System (INIS)

    Matsuno, Y.

    1982-01-01

    The fast breeder reactor development project in Japan has been in progress for the past twelve months and will be continued this fiscal year, from April 1982 through March 1983, at a similar scale of effort both in budget and personnel to those of the fiscal year of 1981. The 1982 year budget for R and D work and for construction of a prototype fast breeder reactor MONJU is approximately 20 and 27 billion yen respectively, excluding wages for the personnel of the Power Reactor and Nuclear Fuel Development Corporation, PNC. The number of the technical people currently engaged in the fast breeder reactor development in the PNC is approximately 530, excluding those working for plutonium fuel fabrication. Concerning the experimental fast reactor JOYO, power increase from 50 MWt to 75 MWt was made in July 1979 and six operational cycles at 75 MWt were completed in December 1981. With respect to the prototype reactor MONJU, progress toward construction has been made and an environmental impact statement of the reactor was approved by the authorities concerned, and the licensing of the first step was completed at the end of 1981. Preliminary design studies of a large LMFBR are being made by PNC and also by utilities. A design study being conducted by PNC is on a 1000 MWe plant of loop type by extrapolating the technology to be developed by the time of commissioning of MONJU. A group of utilities is conducting a similar study, but covering somewhat wider range of parameters and options of design. Close contact between the group and PNC has been kept. In the future, those design efforts will be combined as a single design effort, when a major effort for developing a large demonstration reactor will be initiated at around the commencement of construction of the prototype reactor MONJU

  4. Recent development of nuclear power in Japan and instrumentation and control system and control room equipment for advanced light water reactors

    International Nuclear Information System (INIS)

    Wakayama, N.

    1992-01-01

    This paper was provided for the 13th IAEA/IWG-NPPCI Meeting and aims to introduce an outline of recent development of nuclear power in Japan and some topics in the field of nuclear power plant control and instrumentation. Forty units of nuclear power plants are in operation in Japan and five units of BWRs and six PWRs are under construction. Construction of prototype FBR Monju have almost completed an construction of High-Temperature Engineering Test Reactor, HTTR, started in March 1991. In parallel of those, extensive effort has been carried out to develop the third generation LWRs which are called Advanced BWR (ABWR) and Advanced PWR (APWR). Two Advanced BWRs are under safety review for construction. Instrumentation and control system of these Advanced LWRs adopts integrated digital I and C system, optical multiplexing signal transmission, fault tolerant control systems and software logic for reactor protection and safety systems and enhances plant control performance and provides human-friendly operation and maintenance environments. Main control room of these Advanced LWRs, comprised with large display panels and advanced console, has special futures such as one-man sit-down operation, human friendly man-machine interface, high level automation in operation and maintenance. (author). 7 refs, 9 figs, 1 tab

  5. Nuclear power in Japan

    International Nuclear Information System (INIS)

    Kishida, J.

    1990-01-01

    The Japanese movement against nuclear energy reached a climax in its upsurge in 1988 two years after the Chernobyl accident. At the outset of that year, this trend was triggered by the government acknowledgement that the Tokyo market was open to foods contaminated by the fallout from Chernobyl. Anti-nuclear activists played an agitating role and many housewives were persuaded to join them. Among many public opinion surveys conducted at that time by newspapers and broadcasting networks, I would like to give you some figures of results from the poll carried out by NHK: Sixty percent of respondents said that nuclear power 'should be promoted', either 'vigorously' 7 or 'carefully' 53%). Sixty-six percent doubted the 'safety of nuclear power', describing it as either 'very dangerous' 20%) or 'rather dangerous' (46%). Only 27% said it was 'safe'. In other words, those who acknowledged the need for nuclear power were almost equal in number with those who found it dangerous. What should these figures be taken to mean? I would take note of the fact that nearly two-thirds of valid responses were in favor of nuclear power even at the time when public opinion reacted most strongly to the impact of the Chernobyl accident. This apparently indicates that the majority of the Japanese people are of the opinion that they would 'promote nuclear power though it is dangerous' or that they would 'promote it, but with the understanding that it is dangerous'. But the anti-nuclear movement is continuing. It remains a headache for both the government and the electric utilities. But we can regard the anti-nuclear movement in Japan as not so serious as that faced by other industrial nations

  6. Nuclear power in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Kishida, J [Japan Research Institute, Ltd., Tokyo (Japan)

    1990-07-01

    The Japanese movement against nuclear energy reached a climax in its upsurge in 1988 two years after the Chernobyl accident. At the outset of that year, this trend was triggered by the government acknowledgement that the Tokyo market was open to foods contaminated by the fallout from Chernobyl. Anti-nuclear activists played an agitating role and many housewives were persuaded to join them. Among many public opinion surveys conducted at that time by newspapers and broadcasting networks, I would like to give you some figures of results from the poll carried out by NHK: Sixty percent of respondents said that nuclear power 'should be promoted', either 'vigorously' 7 or 'carefully' 53%). Sixty-six percent doubted the 'safety of nuclear power', describing it as either 'very dangerous' 20%) or 'rather dangerous' (46%). Only 27% said it was 'safe'. In other words, those who acknowledged the need for nuclear power were almost equal in number with those who found it dangerous. What should these figures be taken to mean? I would take note of the fact that nearly two-thirds of valid responses were in favor of nuclear power even at the time when public opinion reacted most strongly to the impact of the Chernobyl accident. This apparently indicates that the majority of the Japanese people are of the opinion that they would 'promote nuclear power though it is dangerous' or that they would 'promote it, but with the understanding that it is dangerous'. But the anti-nuclear movement is continuing. It remains a headache for both the government and the electric utilities. But we can regard the anti-nuclear movement in Japan as not so serious as that faced by other industrial nations.

  7. A review of fast reactor program in Japan

    International Nuclear Information System (INIS)

    1992-01-01

    In accordance with the Long-term Program for Development and Utilization of Nuclear Energy defined by the Japan Atomic Energy Commission (JAEC), Power Reactor and Nuclear Fuel Development Corporation (PNC) is playing the key role in the development of a plutonium utilization system by fast breeder reactor (FBR), which is superior to the uranium utilization system by light water reactor, aiming to achieve future stable long-term energy supply and energy security of Japan. The experimental reactor Joyo, located in the O-arai Engineering Center (OEC) of PNC, has provided abundant experimental data and excellent operational records attaining 43,500 hours operation in total by the end of 1991, since its first criticality in 1977. On the prototype reactor Monju, 97.6% of construction works has already been completed and the function tests are in progress aiming at the initial criticality by the end of FY 1992. As for the demonstration fast breeder reactor (DFBR) of Japan, the Japan Atomic Power Company (JAPC) is promoting design study under the contracts with several leading Japanese fabricators, including Toshiba, Hitachi and Mitsubishi Heavy Industries, for selection of the basic specifications of DFBR. The related research and development (R and D) works are underway at several organizations under the discussion and coordination of the Japanese FBR R and D Steering Committee, which was established by the JAPAC, PNC, Japan Atomic Energy Research Institute (JAERI) and Central Research Institute of Electric Power Industry (CRIEPI). Progress of the design study and the related R and D are reported to the Subcommittee on FBR Development Program of JAEC. Recent major emphases on the PNC R and D are placed on the integrated feedback of all existing R and D results and experiences to the development of demonstration reactor. Furthermore, the overall functional and performance tests of Monju, is another important key role to attain further excellency of FBR technology, with

  8. A review of fast reactor program in Japan

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1992-07-01

    In accordance with the Long-term Program for Development and Utilization of Nuclear Energy defined by the Japan Atomic Energy Commission (JAEC), Power Reactor and Nuclear Fuel Development Corporation (PNC) is playing the key role in the development of a plutonium utilization system by fast breeder reactor (FBR), which is superior to the uranium utilization system by light water reactor, aiming to achieve future stable long-term energy supply and energy security of Japan. The experimental reactor Joyo, located in the O-arai Engineering Center (OEC) of PNC, has provided abundant experimental data and excellent operational records attaining 43,500 hours operation in total by the end of 1991, since its first criticality in 1977. On the prototype reactor Monju, 97.6% of construction works has already been completed and the function tests are in progress aiming at the initial criticality by the end of FY 1992. As for the demonstration fast breeder reactor (DFBR) of Japan, the Japan Atomic Power Company (JAPC) is promoting design study under the contracts with several leading Japanese fabricators, including Toshiba, Hitachi and Mitsubishi Heavy Industries, for selection of the basic specifications of DFBR. The related research and development (R and D) works are underway at several organizations under the discussion and coordination of the Japanese FBR R and D Steering Committee, which was established by the JAPAC, PNC, Japan Atomic Energy Research Institute (JAERI) and Central Research Institute of Electric Power Industry (CRIEPI). Progress of the design study and the related R and D are reported to the Subcommittee on FBR Development Program of JAEC. Recent major emphases on the PNC R and D are placed on the integrated feedback of all existing R and D results and experiences to the development of demonstration reactor. Furthermore, the overall functional and performance tests of Monju, is another important key role to attain further excellency of FBR technology, with

  9. Reactor power control device

    International Nuclear Information System (INIS)

    Ishii, Yoshihiko; Arita, Setsuo; Miyamoto, Yoshiyuki; Fukazawa, Yukihisa; Ishii, Kazuhiko

    1998-01-01

    The present invention provides a reactor power control device capable of enhancing an operation efficiency while keeping high reliability and safety in a BWR type nuclear power plant. Namely, the device of the present invention comprises (1) a means for inputting a set value of a generator power and a set value of a reactor power, (2) a means for controlling the reactor power to either smaller one of the reactor power corresponding to the set value of the generator power and the set value of the reactor power. With such procedures, even if the nuclear power plant is set so as to operate it to make the reactor power 100%, when the generator power reaches the upper limit, the reactor power is controlled with a preference given to the upper limit value of the generator power. Accordingly, safety and reliability are not deteriorated. The operation efficiency of the plant can be improved. (I.S.)

  10. 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

  11. Increased SRP reactor power

    International Nuclear Information System (INIS)

    MacAfee, I.M.

    1983-01-01

    Major changes in the current reactor hydraulic systems could be made to achieve a total of about 1500 MW increase of reactor power for P, K, and C reactors. The changes would be to install new, larger heat exchangers in the reactor buildings to increase heat transfer area about 24%, to increase H 2 O flow about 30% per reactor, to increase D 2 O flow 15 to 18% per reactor, and increase reactor blanket gas pressure from 5 psig to 10 psig. The increased reactor power is possible because of reduced inlet temperature of reactor coolant, increased heat removal capacity, and increased operating pressure (larger margin from boiling). The 23% reactor power increase, after adjustment for increased off-line time for reactor reloading, will provide a 15% increase of production from P, K, and C reactors. Restart of L Reactor would increase SRP production 33%

  12. 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

  13. Reactor-vessel-sectioning demonstration

    International Nuclear Information System (INIS)

    Lundgren, R.A.

    1981-07-01

    A successful technical demonstration of simulated reactor vessel sectioning was completed using the combined techniques of air arc gouging and flame cutting. A 4-ft x 3-ft x 9-in. thick sample was fabricated of A36 carbon steel to simulate a reactor vessel wall. A 1/4-in layer of stainless steel (SS) was tungsten inert gas (TIG)-welded to the carbon steel. Several techniques were considered to section the simulated reactor vessel: an air arc gouger was chosen to penetrate the stainless steel, and flame cutting was selected to sever the carbon steel. After the simulated vessel was successfully cut from the SS side, another cut was made, starting from the carbon steel side. This cut was also successful. Cutting from the carbon steel side has the advantages of cost reduction since the air arc gouging step is eliminated and contamination controlled because the molten metal is blown inward

  14. Guide to power reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1959-07-15

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

  15. 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.

  16. Reactor-vessel-sectioning demonstration

    International Nuclear Information System (INIS)

    Lundgren, R.A.

    1981-09-01

    A technical demonstration was successfully completed of simulated reactor vessel sectioning using the combined techniques of air arc gouging and flame cutting. A 4-ft x 3-ft x 9-in. thick sample was fabricated of A36 carbon steel to simulate a reactor vessel wall. A 1/4-in. layer of stainless steel (SS) was tungsten inert gas (TIG)-welded to the carbon steel. Several techniques were considered to section the simulated reactor vessel; air arc gouging was selected to penetrate the stainless steel, and flame cutting was selected to sever the carbon steel. Three sectioning operations were demonstrated. For all three, the operating parameters were the same; but the position of the sample was varied. For the first cut, the sample was placed in a horizontal position, and it was successfully severed from the SS side. For the second cut, the sample was turned over and cut from the carbon steel side. Cutting from the carbon steel side has the advantages of cost reduction

  17. A review of fast reactor program in Japan - April 1983

    International Nuclear Information System (INIS)

    Matsuno, Y.

    1983-01-01

    The fast breeder reactor development project in Japan has been in progress during the past twelve months and will be continued in the next fiscal year, from April 1983 through March 1984, at a similar scale of effort both in budget and personnel to those of the fiscal year of 1982. Concerning the experimental fast reactor, JOYO, all the scheduled testings and operations were completed by the end of 1981 and from the beginning of 1982 the change-out work from Mark-I core to Mark-II core has been continued for 11 months. The initial criticality on the Mark-II core was achieved on 22 Nov. 1982 and after 3 months low power physics tests the reactor power was raised up to 100% (100 MWt) in the middle of March 1983. With respect to the prototype reactor MONJU, progress toward construction has been made and the licensing of the second step will be completed in the first half of 1983. Preliminary design studies of large LMFBR are being made by PNC and also by utilities. A design study being conducted by PNC is on a 1000 MWe plant of loop type by extrapolating the technology to be developed by the time of commissioning of MONJU. A group of utilities is conducting a similar study, but covering somewhat wider range of parameters and options of design. Close contact between the group and PNC has been kept. In the future, those design efforts will be combined as a single design effort, when a major effort for developing a large demonstration reactor will be initiated at around the commencement of construction of the prototype reactor MONJU

  18. State of opening the cover and carrying out the checkup of the reactor vessel of the nuclear-powered ship 'Mutsu' by Japan Atomic Energy Research Institute

    International Nuclear Information System (INIS)

    1989-01-01

    In the checkup by opening the cover of the reactor vessel of the nuclear-powered ship 'Mutsu', Japan Atomic Energy Research Institute carried out the checkup and maintenance for the reactor proper, control system and primary coolant facilities including the secondary side of steam generators and the pressure balancing valve of the containment vessel. The works were classified into the opening of the reactor, checkup, maintenance and restoration. The opening was begun on August 4, 1988, and finished on December 5. The checkup and maintenance were begun on September 22, and are still continued now. The maximum radiation dose rate on the surfaces of fuel assemblies and control rods and at the positions 1 m distant from them was measured. The results of the checkup of various components are reported. In 290 absorbent rods of control rods, spot corrosion and discoloration were observed, of which the spot corrosion penetrated the walls of 4 rods. Also in 12 fuel rods, spot corrosion was observed near the welded end plugs, but leak was not observed. (K.I.)

  19. Reactor power measuring device

    International Nuclear Information System (INIS)

    Izumi, Mikio; Sano, Yuji; Seki, Eiji; Yoshida, Toshifumi; Ito, Toshiaki.

    1993-01-01

    The present invention provides a self-powered long detector having a sensitivity over the entire length of a reactor core as an entire control rod withdrawal range of a BWR type reactor, and a reactor power measuring device using a gamma ray thermometer which scarcely causes sensitivity degradation. That is, a hollow protection pipe is disposed passing through the reactor core from the outside of a reactor pressure vessel. The self-powered long detectors and the gamma ray thermometers are inserted and installed in the protection pipe. An average reactor power in an axial direction of the reactor relative to a certain position in the horizontal cross section of the reactor core is determined based on the power of the self-powered long detector over the entire length of the reactor core. Since the response of the self-powered detector relative to a local power change is rapid, the output is used as an input signal to a safety protection device of the reactor core. Further, a gamma ray thermometer secured in the reactor and having scarce sensitivity degradation is used instead of an incore travelling neutron monitor used for relative calibration of an existent neutron monitor secured in the reactor. (I.S.)

  20. Development of demonstration advanced thermal reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nishimura, Seiji; Oguchi, Isao; Touhei, Kazushige

    1982-08-01

    The design of the advanced thermal demonstration reactor with 600 MWe output was started in 1975. In order to make the compact core, 648 fuel assemblies, each comprising 36 fuel rods, were used, and the mean channel output was increased by 20% as compared with the prototype reactor. The heavy water dumping mechanism for the calandria was abolished. Advanced thermal reactors are suitable to burn plutonium, since the control rod worth does not change, the void reactivity coefficient of coolant shifts to the negative side, and the harmful influence of high order plutonium is small. The void reactivity coefficient is nearly zero, the fluctuation of output in relation to pressure disturbance is small, and the local output change of fuel by the operation of control rods is small, therefore, the operation following load change is relatively easy. The coolant recirculation system is of independent loop construction dividing the core into two, and steam and water are separated in respective steam drums. At present, the rationalizing design is in progress by the leadership of the Power Reactor and Nuclear Fuel Development Corp. The outline of the demonstration reactor, the reactor construction, the nuclear-thermal-hydraulic characteristics and the output control characteristics are reported.

  1. Development of demonstration advanced thermal reactor

    International Nuclear Information System (INIS)

    Nishimura, Seiji; Oguchi, Isao; Touhei, Kazushige.

    1982-01-01

    The design of the advanced thermal demonstration reactor with 600 MWe output was started in 1975. In order to make the compact core, 648 fuel assemblies, each comprising 36 fuel rods, were used, and the mean channel output was increased by 20% as compared with the prototype reactor. The heavy water dumping mechanism for the calandria was abolished. Advanced thermal reactors are suitable to burn plutonium, since the control rod worth does not change, the void reactivity coefficient of coolant shifts to the negative side, and the harmful influence of high order plutonium is small. The void reactivity coefficient is nearly zero, the fluctuation of output in relation to pressure disturbance is small, and the local output change of fuel by the operation of control rods is small, therefore, the operation following load change is relatively easy. The coolant recirculation system is of independent loop construction dividing the core into two, and steam and water are separated in respective steam drums. At present, the rationalizing design is in progress by the leadership of the Power Reactor and Nuclear Fuel Development Corp. The outline of the demonstration reactor, the reactor construction, the nuclear-thermal-hydraulic characteristics and the output control characteristics are reported. (Kako, I.)

  2. Representativeness elements of an hybrid reactor demonstrator

    International Nuclear Information System (INIS)

    Kerdraon, D.; Billebaud, A.; Brissot, R.; David, S.; Giorni, A.; Heuer, D.; Loiseaux, J.M.; Meplan, O.

    2000-11-01

    This document deals with the quantification of the minimum thermal power level for a demonstrator and the definition of the physical criteria which define the representative character of a demonstrator towards a power reactor. Solutions allowing to keep an acceptable flow in an industrial core, have also been studied. The document is divided in three parts: the representativeness elements, the considered solutions and the characterization of the neutrons flows at the interfaces and the dose rates at the outer surface of the vessel. (A.L.B.)

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

    International Nuclear Information System (INIS)

    1987-01-01

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

  4. Reactor power monitoring device

    International Nuclear Information System (INIS)

    Dogen, Ayumi; Ozawa, Michihiro.

    1983-01-01

    Purpose: To significantly improve the working efficiency of a nuclear reactor by reflecting the control rod history effect on thermal variants required for the monitoring of the reactor operation. Constitution: An incore power distribution calculation section reads the incore neutron fluxes detected by neutron detectors disposed in the reactor to calculate the incore power distribution. A burnup degree distribution calculation section calculates the burnup degree distribution in the reactor based on the thus calculated incore power distribution. A control rod history date store device supplied with the burnup degree distribution renews the stored control rod history data based on the present control rod pattern and the burnup degree distribution. Then, thermal variants of the nuclear reactor are calculated based on the thus renewed control rod history data. Since the control rod history effect is reflected on the thermal variants required for the monitoring of the reactor operation, the working efficiency of the nuclear reactor can be improved significantly. (Seki, T.)

  5. Japan

    International Nuclear Information System (INIS)

    Huttner, Kevin; Suzuki, Tatsujiro

    1987-01-01

    The Japanese nuclear power programme began with reactors imported from the United States. A natural uranium heavy water reactor using domestic materials was started in 1958. Subsequent progress was with light water reactors imported from the United States. Domestic reactor development was of a fast breeder reactor and an advanced thermal reactor. By March 1986 there were 32 commercial power plants in operation which produced approximately 23% of the electricity consumed in Japan. Ten more are under construction and six more are planned. Their location and comparative generating cost are tabulated. Energy demand and targets for nuclear power generation are discussed. The FBR advanced thermal reactor and high temperature reactor programmes are summarized. The Japanese nuclear fuel cycle - uranium prospecting, enrichment, reprocessing, the development of mixed oxide fuels, thermal recycling and radioactive waste management is also discussed. (U.K.)

  6. Programme and current status of fast breeder reactor development in Japan

    International Nuclear Information System (INIS)

    Suita, T.; Oyama, A.

    1977-01-01

    In 1967 the Japan Atomic Energy Commission revised her long term programme after a two year study for giving principles to her nuclear energy development programme, which indicated the dominant role of nuclear energy mid 1980's in the electric power generation and stressed the necessity of developing fast breeder reactors. It also recommended to organize a nucleus to undertake this nation-wide project, bringing together the total capability available throughout the country. Accordingly, the Power Reactor and Nuclear Fuel Development Corporation (PNC) was established in 1967 to develop two sodium-cooled fast reactors, an experimental fast reactor of about 100 MW thermal and a prototype fast breeder reactor of about 300 MW electrical, both using mixed oxide fuels. Construction of the experimental fast reactor started in 1970 and was essentially completed at the end of in 1974. The precommissioning test was followed in parallel with re-evaluating quality assurance of all systems. Physics test will be initiated around the end of 1976. The conceptual design of the prototype fast breeder reactor is now toward its final stage. Surveys on its proposed site have just started. Construction will start in 1978. Beside R and D works conducted by many organizations in Japan as well as under the international cooperation, several key test facilities were installed by PNC itself to conduct in-sodium test of full-size prototype components including 50 MW steam generators and post-irradiation-examination of fuels and materials. Recently an interim report was issued to an ad-hoc committee organized by JAEC to evaluate future prospect of the fuel cycle and power reactors. This recommended start of construction of the prototype reactor as scheduled and the large demonstration reactor to be followed to the prototype. Thus the fast breeder reactor is indicated as the most indispensable in 1990's

  7. Fusion Power Demonstration III

    International Nuclear Information System (INIS)

    Lee, J.D.

    1985-07-01

    This is the third in the series of reports covering the Fusion Power Demonstration (FPD) design study. This volume considers the FPD-III configuration that incorporates an octopole end plug. As compared with the quadrupole end-plugged designs of FPD-I and FPD-II, this octopole configuration reduces the number of end cell magnets and shortens the minimum ignition length of the central cell. The end-cell plasma length is also reduced, which in turn reduces the size and cost of the end cell magnets and shielding. As a contiuation in the series of documents covering the FPD, this report does not stand alone as a design description of FPD-III. Design details of FPD-III subsystems that do not differ significantly from those of the FPD-II configuration are not duplicated in this report

  8. Data list of nuclear power plants in Japan

    International Nuclear Information System (INIS)

    Izumi, Fumio; Ito, Noboru; Higuchi, Suminori; Kobayashi, Kensuke; Tobioka, Toshiaki

    1987-03-01

    The PPD (Nuclear Power Plant Data Base) has been under development in JAERI since 1983 as a six-year program to provide useful information for reactor safety regulation and reactor safety research. Information source of the PPD is mainly based on SAR's (Safety Analysis Reports) of 35 nuclear power plants which are operating, under construction or under licensing review in Japan. The report BWR edition consists of several data lists stored in the PPD, concerning safety design such as performances, equipments and installations of 18 BWR plants in Japan. The informations are based on SAR Attachment Chapter 8 ''Nuclear Reactor Facility Safety Design''. (author)

  9. Damage of the Unit 1 reactor building overhead bridge crane at Onagawa Nuclear Power Station caused by the Great East Japan Earthquake and its repair works

    International Nuclear Information System (INIS)

    Sugamata, Norihiko

    2014-01-01

    The driving shaft bearings of the Unit 1 overhead bridge crane were damaged by the Great East Japan Earthquake at Onagawa Nuclear Power Station. The situation, investigation and repair works of the bearing failure are introduced in this paper. (author)

  10. What is nuclear power in Japan?

    Science.gov (United States)

    Suzuki, Toshikazu

    2011-03-01

    The aggressive use of such non-fossil energy as the atomic energy with high power density and energy production efficiency is an indispensable choice aiming at the low-carbon society. There is a trial calculation that the carbon dioxide emission of 40000 ton can be suppressed by nuclear power generation by one ton of uranium. The basis of nuclear research after the Second World War in Japan was established by the researchers learnt in Argonne National Laboratory. In 2010, NPPs under operation are 54 units and the total electric generating power is 48.85GW. The amount of nuclear power generation per person of the people is 0.38kW in Japan, and it is near 0.34kW of the United States. However, the TMI accident and the Chernobyl disaster should have greatly stagnated the nuclear industry of Japan although it is not more serious than the United States. A lot of Japanese unconsciously associate a nuclear accident with the atomic bomb. According to the investigation which Science and Technology Agency carried out to the specialist in 1999, ``What will be the field where talent should be emphatically sent in the future?'' the rank of nuclear technology was the lowest in 32 fields. The influence of the nuclear industry stagnation was remarkable in the education. The subject related to the atomic energy of a university existed 19 in 1985 that was the previous year of the Chernobyl disaster decreased to 7 in 2003. In such a situation, we have to rely on the atomic energy because Japan depends for 96% of energy resources on import. The development of the fuel reprocessing and the fast breeder reactor has been continued in spite of a heavy failure. That is the only means left behind for Japan to be released from both fossil fuel and carbon dioxide.

  11. Reactor power control device

    International Nuclear Information System (INIS)

    Doi, Kazuyori.

    1981-01-01

    Purpose: To automatically control the BWR type reactor power by simple and short-time searching the load pattern nearest to the required pattern at a nuclear power plant side. Constitution: The reactor power is automatically regulated by periodical modifying of coefficients fitting to a reactor core model, according as a required load pattern. When a load requirement pattern is given, a simulator estimates the total power change and the axial power distribution change from a xenon density change output calculated by a xenon dynamic characteristic estimating device, and a load pattern capable of being realized is searched. The amount to be recirculated is controlled on the basis of the load patteren thus searched, and the operation of the BWR type reactor is automatically controlled at the side of the nuclear power plant. (Kamimura, M.)

  12. Reactor power distribution monitor

    International Nuclear Information System (INIS)

    Hoizumi, Atsushi.

    1986-01-01

    Purpose: To grasp the margin for the limit value of the power distribution peaking factor inside the reactor under operation by using the reactor power distribution monitor. Constitution: The monitor is composed of the 'constant' file, (to store in-reactor power distributions obtained from analysis), TIP and thermocouple, lateral output distribution calibrating apparatus, axial output distribution synthesizer and peaking factor synthesizer. The lateral output distribution calibrating apparatus is used to make calibration by comparing the power distribution obtained from the thermocouples to the power distribution obtained from the TIP, and then to provide the power distribution lateral peaking factors. The axial output distribution synthesizer provides the power distribution axial peaking factors in accordance with the signals from the out-pile neutron flux detector. These axial and lateral power peaking factors are synthesized with high precision in the three-dimensional format and can be monitored at any time. (Kamimura, M.)

  13. The japan a nuclear power?

    International Nuclear Information System (INIS)

    Cumin, D.; Joubert, J.P.

    2003-01-01

    This work analyzes the Japan nuclear policy, in the frame of its foreign and safety policy in Pacific Asia, since the end of the cold war, especially the relations with the Usa and China. The Japan is a civil power because it has submitted the military institution to juridical restrictions and because it does not rely on the armed force to promote its national interests. The anti nuclear speech is joined with the acknowledgement of the dissuasion necessity, of the control of industrial processes and energy channels susceptible of military applications. Cultivating the ambiguity, the Japanese government can send a dissuasive message, perfectly legible, kind of communication of latent intimidation constituted by the virtual nuclear power of a state that takes part to the non proliferation treaty. (N.C.)

  14. Reactor power control system

    International Nuclear Information System (INIS)

    Tomisawa, Teruaki.

    1981-01-01

    Purpose: To restore reactor-power condition in a minimum time after a termination of turbine bypass by reducing the throttling of the reactor power at the time of load-failure as low as possible. Constitution: The transient change of the internal pressure of condenser is continuously monitored. When a turbine is bypassed, a speed-control-command signal for a coolant recirculating pump is generated according as the internal pressure of the condenser. When the signal relating to the internal pressure of the condenser indicates insufficient power, a reactor-control-rod-drive signal is generated. (J.P.N.)

  15. Present status of Japan materials testing reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hori, Naohiko; Kaminaga, Masanori; Kusunoki, Tsuyoshi; Ishihara, Masahiro; Niimi, Motoji; Komori, Yoshihiro; Suzuki, Masahide; Kawamura, Hiroshi [Japan Atomic Energy Agency, Oarai Research and Development Center, Oarai, Ibaraki (Japan)

    2012-03-15

    The Japan Materials Testing Reactor (JMTR) in Japan Atomic Energy Agency (JAEA) is a light water cooled tank type reactor with first criticality in March 1968. Owing to the connection between the JMTR and hot laboratory by a canal, easy re-irradiation tests can be conducted with safe and quick transportation of irradiated samples. The JMTR has been applied to fuel/material irradiation examinations for LWRs, HTGR, fusion reactor and RI production. However, the JMTR operation was once stopped in August 2006, and check and review on the reoperation had been conducted by internal as well as external committees. As a result of the discussion, the JMTR reoperation was determined, and refurbishment works started from the beginning of JFY 2007. The refurbishment works have finished in March 2011 taking four years from JFY 2007. Unfortunately, at the end of the JFY 2010 on March 11, the Great-Eastern-Japan-Earthquake occurred, and functional tests before the JMTR restart, such as cooling system, reactor control system and so on, were delayed by the earthquake. Moreover, a detail inspection found some damages such as slight deformation of the truss structure at the roof of the JMTR reactor building. Consequently, the restart of the JMTR will be delayed from June to next October, 2012. Now, the safety evaluation after the earthquake disaster is being carried out aiming at the restart of the JMTR. The renewed JMTR will be started from JFY 2012 and operated for a period of about 20 years until around JFY 2030. The usability improvement of the JMTR, e.g. higher reactor availability, shortening turnaround time to get irradiation results, attractive irradiation cost, business confidence, is also discussed with users as the preparations for re-operation. (author)

  16. Present status of Japan materials testing reactor

    International Nuclear Information System (INIS)

    Hori, Naohiko; Kaminaga, Masanori; Kusunoki, Tsuyoshi; Ishihara, Masahiro; Niimi, Motoji; Komori, Yoshihiro; Suzuki, Masahide; Kawamura, Hiroshi

    2012-01-01

    The Japan Materials Testing Reactor (JMTR) in Japan Atomic Energy Agency (JAEA) is a light water cooled tank type reactor with first criticality in March 1968. Owing to the connection between the JMTR and hot laboratory by a canal, easy re-irradiation tests can be conducted with safe and quick transportation of irradiated samples. The JMTR has been applied to fuel/material irradiation examinations for LWRs, HTGR, fusion reactor and RI production. However, the JMTR operation was once stopped in August 2006, and check and review on the reoperation had been conducted by internal as well as external committees. As a result of the discussion, the JMTR reoperation was determined, and refurbishment works started from the beginning of JFY 2007. The refurbishment works have finished in March 2011 taking four years from JFY 2007. Unfortunately, at the end of the JFY 2010 on March 11, the Great-Eastern-Japan-Earthquake occurred, and functional tests before the JMTR restart, such as cooling system, reactor control system and so on, were delayed by the earthquake. Moreover, a detail inspection found some damages such as slight deformation of the truss structure at the roof of the JMTR reactor building. Consequently, the restart of the JMTR will be delayed from June to next October, 2012. Now, the safety evaluation after the earthquake disaster is being carried out aiming at the restart of the JMTR. The renewed JMTR will be started from JFY 2012 and operated for a period of about 20 years until around JFY 2030. The usability improvement of the JMTR, e.g. higher reactor availability, shortening turnaround time to get irradiation results, attractive irradiation cost, business confidence, is also discussed with users as the preparations for re-operation. (author)

  17. Power reactors operational diagnosis

    International Nuclear Information System (INIS)

    Dach, K.; Pecinka, L.

    1976-01-01

    The definition of reactor operational diagnostics is presented and the fundamental trends of research are determined. The possible sources of power reactor malfunctions, the methods of defect detection, the data evaluation and the analysis of the results are discussed in detail. In view of scarcity of a theoretical basis and of insufficient in-core instrumentation, operational diagnostics cannot be as yet incorporated in a computer-aided reactor control system. (author)

  18. Nuclear power development in Japan

    International Nuclear Information System (INIS)

    Mishiro, M.

    2000-01-01

    This article describes the advantages of nuclear energy for Japan. In 1997 the composition of the total primary energy supply (TPES) was oil 52.7%, coal 16.5%, nuclear 16.1% and natural gas 10.7%. Nuclear power has a significant role to play in contributing to 3 national interests: i) energy security, ii) economic growth and iii) environmental protection. Energy security is assured because a stable supply of uranium fuel can be reasonably expected in spite of dependence on import from abroad. Economic growth implies the reduction of energy costs. As nuclear power is capital intensive, the power generation cost is less affected by the fuel cost, therefore nuclear power can realize low cost by favoring high capacity utilization factor. Fossil fuels have substantial impacts on environment such as global warming and acid rain by releasing massive quantities of CO 2 , so nuclear power is a major option for meeting the Kyoto limitations. In Japan, in 2010 nuclear power is expected to reach 17% of TPES and 45% of electricity generated. (A.C.)

  19. The Japan Power Demonstration Reactor (JPDR) dismantling activities. Dismantling of the reactor enclosure and the auxiliary buildings

    International Nuclear Information System (INIS)

    Seiki, Yoshihiro; Kubo, Takashi.

    1996-01-01

    As the final stage of the JPDR decommissioning program, after the major components were removed from each building of JPDR, the dismantling activities proceeded to the decontamination of contaminated concrete surface and the final radiation survey of buildings. These activities were conducted to verify the developed techniques and the detailed procedures for decontamination, and to allow unrestricted use of the JPDR buildings. Following the decontamination of buildings, the dismantling of each building was started. Before dismantling the buildings, the radiation control designations were changed. The buildings that contaminated embedded pipes were changed from first-class radiation controlled areas to second-class radiation controlled areas. On the other hand, the buildings that had no contaminated pipes were changed to uncontrolled areas. A first-class radiation controlled area allows the use of unsealed sources ; thus, radioactive contamination may exist. A second-class radiation controlled area is one where only sealed sources are allowed. Significant quantities of data and experience were obtained during these activities. The practical procedures for decontamination, the final survey of radioactivity, and the dismantling work of buildings were described in this report. (author)

  20. A review of fast reactor progress in Japan, March 1979

    Energy Technology Data Exchange (ETDEWEB)

    Tomabechi, K

    1979-07-01

    The fast reactor development project in Japan will be continued in the next fiscal year, from April 1979 through March 1980, at a similar scale of effort both in budget and personnel, to those of the fiscal year of 1978. The total budget for LMFBR development for the next fiscal year is approximately 24 billion Yen, excluding wages of the personnel of the Power Reactor and Nuclear Fuel Development Corporation, PNC. The number of the technical people currently engaging in the fast reactor development in the PNC is approximately 500, excluding those working for plutonium fuel fabrication. Concerning the experimental fast reactor JOYO, approval for power increase from presently approved 50 MWt to 75 MWt with the present core and also to 100 MWt with a modified core in the future was granted by the regulatory authority in September 1978. Two operational cycles at 50 MWt have been completed very recently and preparation for power increase to 75 MWt is being made. With respect to the prototype fast breeder reactor MONJU, progress toward construction is being made and an environmental impact statement of MONJU filed last autumn is being reviewed by the concerned authorities. By the new atomic energy law recently made effective in Japan, the tasks of the former Japan Atomic Energy Commission were split into two and the Atomic Energy Safety Commission was newly established on 4th October 1978 in order to deal with nuclear safety problems in the country. All other problems are treated by the Atomic Energy Commission, as before. Highlights and topics of the fast reactor development activities in the past twelve months are summarized in this paper.

  1. A review of fast reactor programme in Japan

    International Nuclear Information System (INIS)

    Matsuno, Y.; Bando, S.

    1981-03-01

    The fast breeder reactor development project in Japan made progress in the past year, and will be continued in the next fiscal 1981. The scale of efforts both in budget and personnel will be similar to those in fiscal 1980. The budget for R and D works and for the construction of the fast breeder prototype reactor ''Monju'' will be approximately 20 billion yen and 27 billion yen, respectively, excluding the wage of the personnel concerned. The number of the technical personnel currently engaging in fast breeder reactor development in the Power Reactor and Nuclear Fuel Development Corp. is about 530. As for the experimental fast reactor ''Joyo'', three operational cycles at 75 MWt have been completed in August, 1980, and the fourth cycle has started in March, 1981. As for the prototype reactor ''Monju'', progress was made toward the construction, and the environmental impact statement on the reactor was approved by the authorities concerned. The studies on the preliminary design of large LMFBRs have been made by the PNC and also by power companies. The design study carried out by the PNC is concerned with a 1000 MWe plant of loop type by extrapolating the technology to be developed by the time of the commissioning of ''Monju''. The highlights and topics in the development activities for fast breeder reactors in the past twelve months are summarized in this report. (Kako, I.)

  2. 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.

  3. Safety aspect of digital reactor protection system in Japan

    International Nuclear Information System (INIS)

    Ogiso, Zen-Ichi

    1998-01-01

    It was early in 1980's that the digital controllers were first applied to nuclear power plant in japan. After that, their application area had been expanding gradually, reaching to the overall integrated digital system including the safety system in Kashiwazaki-Kariwa units 6 and 7. The software for computer-based systems has been produced using the graphical language ''POL'' in Japanese nuclear power plants. It is the fundamental principle that the reliability of the software should be assured through the properly managed quality assurance. The POL-based system is fitted to this principle. In applying POL-based systems to safety system, the MITI, Ministry of International Trade and Industry, identified the licensing issues as the regulatory body, while the utilities had developed the digital technology feasible to the safety application. Through the activities, a specific industrial design guide for the software important to safety was established and the adequacy of the technology was certified through the demonstration tests of the integrated system. In the safety examination of the digital reactor protection system of K-6/7, the application of POL were approved. The POL-based systems in nuclear power plants were successful design and production process of the POL-based systems. This paper describes the activities in licensing and maintaining the computer-based systems by the utilities and manufacturers as well as the MITI. (author)

  4. Reactor power control device

    International Nuclear Information System (INIS)

    Kobayashi, Akira.

    1980-01-01

    Purpose: To prevent misoperation in a control system for the adjustment of core coolant flow rate, and the increase in the neutron flux density caused from the misoperation in BWR type reactors. Constitution: In a reactor power control system adapted to control the reactor power by the adjustment of core flow rate, average neutron flux signals of a reactor core, entire core flow rate signals and operation state signals for coolant recycling system are inputted to a microcomputer. The outputs from the computer are sent to a recycling MG set speed controller to control the reactor core flow rate. The computer calculates the change ratio with time in the average neutron flux signals, correlation between the average neutron flux signals and the entire core flow rate signals, change ratio with time in the operation state signals for the coolant recycling system and the like and judges the abnormality in the coolant recycling system based on the calculated results. (Ikeda, J.)

  5. Development of nuclear powered ship in Japan

    International Nuclear Information System (INIS)

    Sato, Hiroshi

    1976-01-01

    The development of nuclear merchant ship in Japan was started in 1955 by the establishment of Nuclear Ship Study Group, and since then, the investigation, test and research on nuclear ships have been continued. As a result, a nuclear ocean observation and supply ship was designed for trial. Researches were carried out also in JAERI and Institute for Technical Research of Ships. Meanwhile, the nuclear icebreaker Lenin was completed in Soviet Union in 1959, the nuclear ship Savannah set out for maiden voyage in U.S. in 1962, and the construction of the nuclear ore carrier Otto Hahn was prepared in FRG. Japan Nuclear Ship Development Corp. was established in 1963, and started the design and construction of the first nuclear ship in Japan, Mutsu. The basic policy in the construction is the improvement of nuclear ship technology, the securing of safety, and the use of domestic technologies as far as possible. The progress of the design, construction and test of the Mutsu is described. Owing to the problem of radiation leak, the development of nuclear ships stagnated for a while, but the nuclear plant of the Mutsu demonstrated the expected performance in the functional test, land criticality test and zero output test, and it is expected that the bud of the independent development brought up so far can bear valuable fruit. The independent development of marine nuclear reactors should be continued by selecting the way most suitable to Japan. (Kako, I.)

  6. Data list of nuclear power plants in Japan

    International Nuclear Information System (INIS)

    Izumi, Fumio; Horikami, Kunihiko; Kobayashi, Kensuke; Namatame, Ken.

    1993-01-01

    The development of the database called PPD (Nuclear Power Plant Database) has started in 1983 at JAERI as a six-year program to provide useful information for reactor safety regulation and reactor safety research. In 1988 the program has been accomplished, and since then the data in the database has been updating and adding. Information source of the PPD is based on SAR's (Safety Analysis Report) of 47 nuclear power plants which are operating, under construction or under licensing review in Japan. The report, BWR edition, consists of lists of major data stored in the PPD, relating to safety design of 25 BWR plants in Japan. (author)

  7. JTEC panel on nuclear power in Japan. Final report

    International Nuclear Information System (INIS)

    Hansen, K.F.; Behnke, W.B.; Cousin, S.B.; Evans, E.A.; Olander, D.R.

    1990-10-01

    The report examines the status and direction of nuclear power-related research and development in Japan in six areas: the nuclear fuel cycle, nuclear materials, instrumentation and control technology, CAD/CAM, nuclear safety research, and nuclear plant construction. Overall findings suggest that the nuclear power industry in Japan is at an advanced state of development; in particular, Japan is now technologically self-sufficient. Long-term goals of the Japanese program include closure of the complete fuel cycle and pursuit of the liquid metal fast breeder reactor as the future base system

  8. Nuclear power supply (Japan Nuclear Safety Institute)

    International Nuclear Information System (INIS)

    Kameyama, Masashi

    2013-01-01

    After experienced nuclear disaster occurred on March 11, 2011, role of nuclear power in future energy share in Japan became uncertain because most public seemed to prefer nuclear power phase out to energy security or costs. Whether nuclear power plants were safe shutdown or operational, technologies were requisite for maintaining their equipment by refurbishment, partly replacement or pressure proof function recovery works, all of which were basically performed by welding. Nuclear power plants consisted of tanks, piping and pumps, and considered as giant welded structures welding was mostly used. Reactor pressure vessel subject to high temperature and high pressure was around 200mm thick and made of low-alloy steels (A533B), stainless steels (308, 316) and nickel base alloys (Alloy 600, 690). Kinds of welding at site were mostly shielded-metal arc welding and TIG welding, and sometimes laser welding. Radiation effects on welding of materials were limited although radiation protection was needed for welding works under radiation environment. New welding technologies had been applied after their technical validation by experiments applicable to required regulation standards. Latest developed welding technologies were seal welding to prevent SCC propagation and temper-bead welding for cladding after removal of cracks. Detailed procedures of repair welding of Alloy 600 at the reactor outlet pipe at Oi Nuclear Power Plants unit 3 due to PWSCC were described as an example of crack removal and water jet peening, and then overlay by temper-bead welding using Alloy 600 and clad welding using Alloy 690. (T. Tanaka)

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

    International Nuclear Information System (INIS)

    Kaneko, Yoshihiko; Kaieda, Keisuke

    2000-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-10-01

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

  11. Operating US power reactors

    International Nuclear Information System (INIS)

    Silver, E.G.

    1988-01-01

    This update, which appears regularly in each issue of Nuclear Safety, surveys the operations of those power reactors in the US which have been issued operating licenses. Table 1 shows the number of such reactors and their net capacities as of September 30, 1987, the end of the three-month period covered in this report. Table 2 lists the unit capacity and forced outage rate for each licensed reactor for each of the three months (July, August, and September 1987) covered in this report and the cumulative values of these parameters since the beginning of commercial operation. In addition to the tabular data, this article discusses other significant occurrences and developments that affected licensed US power reactors during this reporting period. Status changes at Braidwood Unit 1, Nine Mile Point 2, and Beaver Valley 2 are discussed. Other occurrences discussed are: retraining of control-room operators at Peach Bottom; a request for 25% power for Shoreham, problems at Fermi 2 which delayed the request to go to 75% power; the results of a safety study of the N Reactor at Hanford; a proposed merger of Pacific Gas and Electric with Sacramento Municipal Utility District which would result in the decommissioning of Rancho Seco; the ordered shutdown of Oyster Creek; a minor radioactivity release caused by a steam generator tube rupture at North Anna 1; and 13 fines levied by the NRC on reactor licensees

  12. ORNL fusion power demonstration study: arguments for a vacuum building in which to enclose a fusion reactor

    International Nuclear Information System (INIS)

    Werner, R.W.

    1976-12-01

    Fusion reactors as presently contemplated are excessively complicated, are virtually inaccessible for some repairs, and are subject to frequent loss of function. This dilemma arises in large part because the closed surface that separates the ''hard'' vacuum of the plasma zone from atmospheric pressure is located at the first wall or between blanket and shield. This closed surface is one containing hundreds to thousands of linear meters of welds or mechanical seals which are subject to radiation damage and cyclic fatigue. In situ repair is extremely difficult. This paper examines the arguments favoring the enclosing of the entire reactor in a vacuum building and thus changing the character of this closed surface from one requiring absolute vacuum integrity to one of high pumping impedance. Two differentially pumped vacuum zones are imagined, one clean zone for the plasma and one for the balance of the volume. Both would be at substantially the same pressure. Other advantages for the vacuum enclosure are also cited and discussed

  13. Progress on development of nuclear power in Japan

    International Nuclear Information System (INIS)

    Anon.

    2000-01-01

    Since three Laws on the nuclear power were published 45 years has passed. Now, development on nuclear power in Japan is at an emergent state. In Japan, 51 units of commercial nuclear reactors with 44.917 GW are in operation, occupy about 37% of total electric power generation, and is positioned at an essential basic energy source supporting economical society in Japan. However, an accident occurred at Tokai Works of the JCO Co., Ltd., one of the uranium reconversion company, on September 30, 1999, was the first critical accident in Japan, and became the worst case in history on development of nuclear power in Japan, because of forming three heavy radiation disabled persons (One of them was dead) in its operators. This was a big crisis with relation to existence on development of nuclear power in Japan, by which anxiety and distrust of the Japanese against the nuclear power were amplified rapidly. On the other side, for Japan short in energy sources and of a big energy consumption, in order to intend for a long term to carry out energy security, global environmental conservation, and sustainable maintenance of essential growth, it remains to be one of important optional methods to further promote nuclear power generation and to establish nuclear fuel cycle. Here were described on progress on peaceful applications of nuclear power in Japan, progress on the field of nuclear power in Japan (from 1955 to 1999), progress on Tokai nuclear power station, introduction of nuclear power generation and effort on its domestic production. (G.K.)

  14. Compact power reactor

    International Nuclear Information System (INIS)

    Wetch, J.R.; Dieckamp, H.M.; Wilson, L.A.

    1978-01-01

    There is disclosed a small compact nuclear reactor operating in the epithermal neutron energy range for supplying power at remote locations, as for a satellite. The core contains fuel moderator elements of Zr hydride with 7 w/o of 93% enriched uranium alloy. The core has a radial beryllium reflector and is cooled by liquid metal coolant such as NaK. The reactor is controlled and shut down by moving portions of the reflector

  15. Reactor power control device

    International Nuclear Information System (INIS)

    Imaruoka, Hiromitsu.

    1994-01-01

    A high pressure water injection recycling system comprising injection pipelines of a high pressure water injection system and a flow rate control means in communication with a pool of a pressure control chamber is disposed to a feedwater system of a BWR type reactor. In addition, the flow rate control means is controlled by a power control device comprising a scram impossible transient event judging section, a required injection flow rate calculation section for high pressure water injection system and a control signal calculation section. Feed water flow rate to be supplied to the reactor is controlled upon occurrence of a scram impossible transient event of the reactor. The scram impossible transient event is judged based on reactor output signals and scram operation demand signals and injection flow rate is calculated based on a predetermined reactor water level, and condensate storage tank water or pressure control chamber pool water is injected to the reactor. With such procedures, water level can be ensured and power can be suppressed. Further, condensate storage tank water of low enthalpy is introduced to the pressure suppression chamber pool to directly control elevation of water temperature and ensure integrity of the pressure vessel and the reactor container. (N.H.)

  16. Power reactor design trends

    International Nuclear Information System (INIS)

    Hogan, W.J.

    1985-01-01

    Cascade and Pulse Star represent new trends in ICF power reactor design that have emerged in the last few years. The most recent embodiments of these two concepts, and that of the HYLIFE design with which they will compare them, are shown. All three reactors depend upon protecting structural elements from neutrons, x rays and debris by injecting massive amounts of shielding material inside the reaction chamber. However, Cascade and Pulse Star introduce new ideas to improve the economics, safety, and environmental impact of ICF reactors. They also pose different development issues and thus represent technological alternatives to HYLIFE

  17. The plutonium utilization in thermal and fast reactor in Japan

    International Nuclear Information System (INIS)

    Amanuma, T.; Uematsu, K.

    1977-01-01

    The nuclear power development in Japan is rather extensive one, and the installed nuclear power capacity is expected to reach 49,000 MWe by 1985 and possibly to reach 170,000 MWe by 2000 according to a prediction. Currently istalled nuclear power is mainly based on Light Water type Reactor, and this trend is expected to persist for the time-being. The plutonium produced by LWR will be accumulated to 20 tons by 1985 and to more than 200 tons by 2000. If the produced plutonium will simply be stored, it will raise the economic pressure to utilities and the management and physical protection problems associated with plutonium storing. Therefore, it is not too wise simply to store plutonium in a locked vault. In Japan, there are three ways of solving these problems which are currently worked out. There is no doubt that the best solution is to use plutonium in fast reactors. To reach this goal, an Experimental Fast Reactor ''JOYO''has been constructed and it is waiting for criticality in very near future. A prototype fast breeder reactor ''MONJU'', which is designed for about 300 MWe, is nearing to the last stage of the design work. The start of its construction will take place in a few yesars. The domonstration fast breeder reactor will come next to ''MONJU'' and the large scale commercial use of fast breeder reactor is expected to start around 1995. To anwer the near-term need for plutonium utilization, two technologies, which are equally important to Japan, are currently developed. One is the recycle use of plutonium into LWR. This technology has long been jointly developed by research organizations and utilities. Some of fuel irradiation data are already obtained and the physics study has also been extensive. The application of this technology is expected to start about 1987. The other is to burn plutonium in an Advanced Thermal Reactor (D 2 O moderated, Boiling Water Cooled) which shows better characteristics of using plutonium. The 160 MWe ''Fugen'' is a prototype

  18. The ORNL fusion power demonstration study

    International Nuclear Information System (INIS)

    Shannon, T.E.; Steiner, D.

    1978-01-01

    In this paper, we review the design approach developed in the ORNL Fusion Power Demonstration Study [1]. The major emphasis of this study is in the application of current and near-term technology as the most logical path to near-term demonstration of tokamak fusion power. In addition we are pursuing a number of concepts to simplify the tokamak reactor to be more acceptable to the utility industry as a future source of energy. The discussion will focus on the areas having the greatest overall impact on reactor feasibility: 1) overall size and power output, 2) remote maintenance considerations, 3) electrical power supplies, 4) blanket design; and 5) economics. The tokamak device, by nature of its configuration and pulsed operation, is an exceptionally complex engineering design problem. We have concluded that innovative design concepts are essential to cope with this basic complexity. We feel that the feasibility of tokamak fusion power has been significantly improved by these design approaches. (author)

  19. The UK commercial demonstration fast reactor design

    International Nuclear Information System (INIS)

    Holmes, J.A.G.

    1987-01-01

    The paper on the UK Commercial Demonstration Fast Reactor design was presented to the seminar on 'European Commercial Fast Reactor Programme, London 1987. The design is discussed under the topic headings:- primary circuit, intermediate heat exchangers and pumps, fuel and core, refuelling, steam generators, and nuclear island layout. (U.K.)

  20. Operating parameters of a reactor for early demonstration of electric power generation and the expansion by realization of advanced tokamak plasma

    International Nuclear Information System (INIS)

    Okano, Kunihiko; Asaoka, Yoshiyuki; Hiwatari, Ryoji

    2004-01-01

    Beam driven stable equilibria for a conceptual reactor, Demo-CREST, which was designed for early demonstration of electric power generation, has been investigated. Considering current profiles driven by neutral beams, the attainable normalized beta β N with a stabilization wall is about 3.4 with a normal shear (NS). With reversed shear (RS), a higher β N is attainable. The stable equilibria up to 4.0 can be sustained by a couple of On- and Off-axis beams. In the range of 1.9 N N = 1.9 which is the base design point of Demo-CREST. In the case of RS operation with β N 4.0, the density ratio to the Greenwald limit can be maintain at about unity if high temperature operation with T e > 20 kV is allowable. (author)

  1. Reactor power distribution monitor

    International Nuclear Information System (INIS)

    Sekimizu, Koichi

    1980-01-01

    Purpose: To improve the performance and secure the safety of a nuclear reactor by rapidly computing and display the power density in the nuclear reactor by using a plurality of processors. Constitution: Plant data for a nuclear reactor containing the measured values from a local power monitor LPRM are sent and recorded in a magnetic disc. They are also sent to a core performance computer in which burn-up degree distribution and the like are computed, and the results are sent and recorded in the magnetic disc. A central processors loads programs to each of the processors and applies data recorded in the magnetic disc to each of the processors. Each of the processors computes the corresponding power distribution in four fuel assemblies surrounding the LPRM string by the above information. The central processor compiles the computation results and displays them on a display. In this way, power distribution in the fuel assemblies can rapidly be computed to thereby secure the improvement of the performance and safety of the reactor. (Seki, T.)

  2. 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)

  3. Power reactor noise

    International Nuclear Information System (INIS)

    Thie, J.A.

    1981-01-01

    This book concentrates on the different types of noise present in power reactors and how the analysis of this noise can be used as a tool for reactor monitoring and diagnostics. Noise analysis is a growing field that offers advantages such as simplicity, low cost, and natural multivariable interactions. A major advantage, continuous and undisturbed monitoring, supplies a means of obtaining early warnings of possible reactor malfunctions thus preventing further complications by alerting operators to a problem - and aiding in the diagnosis of that problem - before it demands major repairs. Following an introductory chapter, the theoretical basis for the various methods of noise analysis is explained, and full chapters are devoted to the fundamentals of statistics for time-domain analysis and Fourier series and related topics for frequency-domain analysis. General experimental techniques and associated theoretical considerations are reviewed, leading to discussion of practical applications in the latter half of the book. Besides chapters giving examples of neutron noise and acoustical noise, chapters are also devoted to extensive examples from pressurized water reactor and boiling water reactor power plants

  4. Reactor power monitoring device

    International Nuclear Information System (INIS)

    Kono, Shigehiro.

    1990-01-01

    Among a plurality of power monitoring programs in a reactor power monitoring device, rapid response is required for a scram judging program for the power judging processing of scram signals. Therefore, the scram judging program is stored independently from other power monitoring programs, applied with a priority order, and executed in parallel with other programs, to output scram signals when the detected data exceeds a predetermined value. As a result, the capacity required for the scram judging program is reduced and the processing can be conducted in a short period of time. In addition, since high priority is applied to the scram judging program which is divided into a small capacity, it is executed at higher frequency than other programs when they are executed in parallel. That is, since the entire processings for the power monitoring program are repeated in a short cycle, the response speed of the scram signals required for high responsivity can be increased. (N.H.)

  5. Japan: The Experimental Fast Reactor JOYO. Profile 12

    International Nuclear Information System (INIS)

    2017-01-01

    The experimental fast reactor JOYO of the Japan Atomic Energy Agency (JAEA) is the first sodium-cooled fast reactor (SFR) in Japan. JOYO attained its initial criticality as a breeder core (MK-I core) in 1977. During the MK-I operation, which consisted of two 50 MWt and six 75 MWt duty cycles, the basic characteristics of plutonium (Pu) and uranium (U) mixed oxide (MOX) fuel core and sodium cooling system were investigated and the breeding performance was verified. In 1983, the reactor increased its thermal output up to 100 MWt in order to start the irradiation tests of fuels and materials to be used mainly for other SFRs. Thirty-five duty cycle operations and many irradiation tests were successfully carried out using the MK-II core by 2000. The core was then modified to the MK-III core in 2003. In order to obtain higher fast neutron flux, the core was modified from one region core to two region core with different Pu fissile contents. Accordingly, the reactor power increased up to 140 MWt together with a renewal of intermediate heat exchangers (IHXs) and dump heat exchangers (DHXs). The rated power operation of the MK-III core started in 2004. The MK-III core has been used for the irradiation tests of fuels and materials for future SFRs and other R&D fields like innovative nuclear energy systems and technologies as well. This powerful neutron irradiation flux has an advantage especially for high burn-up fuel irradiation and material irradiation with high neutron dose. This paper shows the outline of the irradiation irradiation irradiation irradiation irradiation capabilities and capacities to develop capabilities and capacities to develop capabilities and capacities to develop capabilities and capacities to develop innovative nuclear energy systems and technologies.

  6. Summary view on demonstration reactor safety

    International Nuclear Information System (INIS)

    Satoh, Kazuziro; Kotake, Shoji; Tsukui, Yutaka; Inagaki, Tatsutoshi; Miura, Masanori

    1991-01-01

    This work presents a summary view on safety design approaches for the demonstration fast breeder reactor (DFBR). The safety objective of DFBR is to be at lea as safe as a LWR. Major safety issues discussed in this paper are; reduction of sodium void reactivity worth, adoption of self-actuated mechanism in the backup shutdown system, use of the direct reactor auxiliary cooling system (DRACS), provision of the containment system. (author)

  7. 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

  8. Reactor power control device

    International Nuclear Information System (INIS)

    Watanabe, Mitsutaka

    1997-01-01

    Hardware of an analog nuclear instrumentation system is reformed, a function generator is added to a setting calculation circuit of the nuclear instrumentation system, and each of setting lines of the nuclear instrumentation system is set in parallel with an upper limit curve in an operation region defined by a second order or third order equation. Upon transient change of abnormal power elevation during operation, scram signals are generated by power change in the same state as 100% rated operation due to elevation of reactor thermal power. Since the operation limit value relative to transient change due to power elevation can be made substantially equal with the same as that upon rated operation, the operation limit value for partial power operation state can be kept substantially the same level as that upon rated operation. When transition change caused by abnormal control rod withdrawal occurs during operation, a control rod withdrawal inhibition signal can ensure the power elevation width equal with that upon rated power operation, and since the withdrawal inhibition signal is generated in substantially the same withdrawing state, the operation limit value relative to a partial power operation state can be kept at the same level as that during rated operation. (N.H.)

  9. Surveillance of nuclear power reactors

    International Nuclear Information System (INIS)

    Marini, J.

    1983-01-01

    Surveillance of nuclear power reactors is now a necessity imposed by such regulatory documents as USNRC Regulatory Guide 1.133. In addition to regulatory requirements, however, nuclear reactor surveillance offers plant operators significant economic advantages insofar as a single day's outage is very costly. The economic worth of a reactor surveillance system can be stated in terms of the improved plant availability provided through its capability to detect incidents before they occur and cause serious damage. Furthermore, the TMI accident has demonstrated the need for monitoring certain components to provide operators with clear information on their functional status. In response to the above considerations, Framatome has developed a line of products which includes: pressure vessel leakage detection systems, loose part detection systems, component vibration monitoring systems, and, crack detection and monitoring systems. Some of the surveillance systems developed by Framatome are described in this paper

  10. Power reactor noise

    International Nuclear Information System (INIS)

    Thie, J.A.

    1981-01-01

    Noise analysis is a growing field that offers advantages such as simplicity, low cost, and natural multivariable interactions. A major advantage, continuous and undisturbed monitoring, supplies a means of obtaining early warnings of possible reactor malfunctions, thus preventing further complications by alerting opeators to a problem - and aiding in the diagnosis of that problem - before it demands major repairs. Dr. Thie hopes to further, through detailed explanations and over 70 illustrations, the acceptance of the use of noise analysis by the nuclear utility industry. Following an introductory chapter, the theoretical basis for the various methods of noise analysis is explained, and full chapters are devoted to the fundamentals of statistics for time-domain analysis and Fourier series and related topics for frequency-domain analysis. General experimental techniques and associated theoretical considerations are reviewed, leading to discussions of practical applications in the latter half of the book. Besides chapters giving examples of neutron noise and acoustical noise, chapters are also devoted to extensive examples from pressurized water reactor and boiling water reactor power plants

  11. 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

  12. A comparative analysis of the domestic and foreign licensing processes for power and non-power reactors

    International Nuclear Information System (INIS)

    Joe, J. C.; Youn, Y. K.; Kim, W. S.; Kim, H. J.

    2003-01-01

    The System-integrated Modular Advanced Reactor (SMART), a small to medium sized integral type Pressurized Water Reactor (PWR) has been developed in Korea. Now, SMART-P, a 1/5 scaled-down of the SMART, is being developed for the purpose of demonstrating the safety and performance of SMART design. The SMART-P is a first-of-a-kind reactor which is utilized for the research and development of a power reactor. Since the licensing process of such a reactor is not clearly specified in the current Atomic Energy Act, a comparative survey and analysis of domestic and foreign licensing processes for power and non-power reactors has been carried out to develop the rationale and technical basis for establishing the licensing process of such a reactor. The domestic and foreign licensing processes of power and non-power reactors have been surveyed and compared, including those of the U.S.A., Japan, France, U.K., Canada, and IAEA. The general trends in nuclear reactor classification, licensing procedures, regulatory technical requirements, and other licensing requirements and regulations have been investigated. The results of this study will be used as the rationale and technical basis for establishing the licensing process of reactors at development stage such as SMART-P

  13. Tokamak experimental power reactor

    International Nuclear Information System (INIS)

    Stacey, W.M. Jr.; Abdou, M.A.; Brooks, J.N.

    1978-01-01

    A tokamak experimental power reactor has been designed that is capable of producing net electric power over a wide range of possible operating conditions. A net production of 81 MW of electricity is expected from the design reference conditions that assume a value of 0.07 for beta-toroidal, a maximum toroidal magnetic field of 9 T and a thermal conversion efficiency of 30%. Impurity control is achieved through the use of a low-Z first wall coating. This approach allows a burn time of 60 seconds without the incorporation of a divertor. The system is cooled by a dual pressurized water/steam system that could potentially provide thermal efficiencies as high as 39%. The first surface facing the plasma is a low-Z coated water cooled panel that is attached to a 20 cm thick blanket module. The vacuum boundary is removed a total of 22 cm from the plasma, thereby minimizing the amount of radiation damage in this vital component. Consideration is given in the design to the possible use of the EPR as a materials test reactor. It is estimated that the total system could be built for less than 550 million dollars

  14. Nuclear power in Japan and the USA

    International Nuclear Information System (INIS)

    Titterton, E.

    1979-06-01

    The development of the nuclear power industry in Japan and the USA is discussed. The author lists the number of nuclear power plants operating, under construction and planned and considers the contribution made by nuclear power stations to the total electricity generated. The advantages of nuclear power to both countries are outlined and forecasts are made of the role to be played by nuclear power in future years

  15. [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.

  16. Handling of spent fuel from research reactors in Japan

    International Nuclear Information System (INIS)

    Kanda, K.

    1997-01-01

    In Japan eleven research reactors are in operation. After the 19th International Meeting on Reduced Enrichment for Research Reactors and Test Reactors (RERTR) on October 6-10, 1996, Seoul, Korea, the Five Agency Committee on Highly Enriched Uranium, which consists of Science and Technology Agency, the Ministry of Education, Science and Culture, the Ministry of Foreign Affairs, Japan Atomic Energy Research Institute (JAERI) and Kyoto University Research Reactor Institute (KURRI) met on November 7,1996, to discuss the handling of spent fuel from research reactors in Japan. Advantages and disadvantages to return spent fuel to the USA in comparison to Europe were discussed. So far, a number of spent fuel elements in JAERI and KURRI are to be returned to the US. The first shipment to the US is planned for 60 HEU elements from JMTR in 1997. The shipment from KURRI is planned to start in 1999. (author)

  17. Conceptual design studies of experimental and demonstration fusion reactors

    International Nuclear Information System (INIS)

    1978-01-01

    Since 1973 the FINTOR Group has been involved in conceptual design studies of TOKAMAK-type fusion reactors to precede the construction of a prototype power reactor plant. FINTOR-1 was the first conceptual design aimed at investigating the main physics and engineering constraints on a minimum-size (both dimensions and thermal power) tokamak experimental reactor. The required plasma energy confinement time as evaluated by various power balance models was compared with the values resulting from different transport models. For the reference design, an energy confinement time ten times smaller than neoclassical was assumed. This also implied a rather high (thermally stable) working temperature (above 20 keV) for the reactor. Other relevant points of the design were: circular plasma cross section, single-null axisymmetric divertor; lithium breeder, stainless steel structures, helium coolant; modular blanket and shield structure; copper-stabilized, superconducting Nb-Ti toroidal field and divertor coils; vertical field and transformer coils inside the toroidal coils; vacuum-tight containment vessel. Solutions involving air and iron transformer cores were compared. These assumptions led to a minimum size reactor with a thermal power of about 100MW and rather large dimensions (major radius of about 9m) similar to those of full-scale power reactors considered in other conceptual studies. The FINTOR-1 analysis was completed by the end of 1976. In 1977 a conceptual design of a Demonstration Power Reactor Plant (FINTOR-D) was started. In this study the main working assumptions differing from those of FINTOR-1 are: non-circular plasma cross section; plasma confinement compatible with trapped ion instabilities; cold (gas) blanket sufficient for wall protection (no divertor); wall loading between 1-3MW/m 2 and thermal power of a few GW. (author)

  18. Power reactors in member states

    International Nuclear Information System (INIS)

    1975-01-01

    This is the first issue of a periodical computer-based listing of civilian nuclear power reactors in the Member States of the IAEA, presenting the situation as of 1 April 1975. It is intended as a replacement for the Agency's previous annual publication of ''Power and Research Reactors in Member States''. In the new format, the listing contains more information about power reactors in operation, under construction, planned and shut down. As far as possible all the basic design data relating to reactors in operation have been included. In future these data will be included also for other power reactors, so that the publication will serve to give a clear picture of the technical progress achieved. Test and research reactors and critical facilities are no longer listed. Of interest to nuclear power planners, nuclear system designers, nuclear plant operators and interested professional engineers and scientists

  19. LWR-plants. Their evolutionary progress in the last half-century. (4) The start of the nuclear power generation in Japan

    International Nuclear Information System (INIS)

    Ishikawa, Hiroshi

    2008-01-01

    Evolutionary progress of the LWR plants in the last half-century was reviewed in series. The start of the nuclear power generation in Japan was reviewed in this article. The Japan Atomic Energy Research Institute (JAERI) promoted nuclear power research and development and introduced the Japan Power Demonstration Reactor (JPDR)-a 12.5 MWe natural circulation BWR, which began operation in 1965. In 1969 its power uprate modifications to a forced circulation BWR (JPDR-II) began and attained operation in 1972. During 50% power test, primary coolant leakage was observed at reactor core spray pipes in 1972. In 1975 the operation resumed and faults observed at condenser tubes in 1976. Primary coolant leakage from in-core flux monitor guide tubes at the bottom of reactor pressure vessel in 1979 led to its permanent shutdown. The nuclear ship Mutsu was put into service in 1970 and during rising power test radiation leakage due to fast neutron streaming was observed in 1974. After modifications of shielding experimental voyage was made in 1991. The first commercial nuclear power reactor, Tokai-1-a 166 MWe gas-cooled (Magnox) reactor, began operation in 1966 and continued until 1998. The LWR plants became the mainstay in Japan. (T. Tanaka)

  20. Promotion of research and development for fast breeder reactors in Japan

    International Nuclear Information System (INIS)

    1989-01-01

    The ''Long-term Program for Development and Utilization of Nuclear Energy'' by Japan Atomic Energy Commission (JAEC) plans to develop fast breeder reactors (FBRs) as the mainstay of Japan's future nuclear power generation. For the development of the FBR, the main principle is to pursue the establishment of FBR plutonium utilization systems, which are superior to the LWR uranium utilization systems. The above program also states as follows: ''Incessant efforts, with proper co-operation of the public and private sectors, will be required for a considerably long time for FBR development, and the concrete measures to be implemented in this connection will be discussed, hereafter, in continuation at the Advisory Committee on FBR Development Program of the Atomic Energy Commission''. It was clearly stated in the ''Long-term Program'' that utilities would play a leading role in the design, construction, and operation of DFBRs to follow ''Monju''. They have entrusted Japan Atomic Power Company (JAPC) with the above tasks. Technical verification and demonstration tests for seismic design and feasibility studies of main components are being conducted with government support. Regarding the nuclear fuel cycle, PNC has been developing FBR fuel reprocessing technologies. Currently, the Recycle Equipment Test Facility (RETF) is being designed to conduct engineering-scale tests. A pilot plant is planned to start operation after 2000 assuming positive results of the tests. (author)

  1. Tokamak reactor startup power

    International Nuclear Information System (INIS)

    Weldon, D.M.; Murray, J.G.

    1983-01-01

    Tokamak startup with ohmic heating (OH)-induced voltages requires rather large voltages and power supplies. On present machines, with no radiofrequency (rf)-assist provisions, hundreds of volts have been specified for their designs. With the addition of electron cyclotron resonant heating (ECRH) assist, the design requirements have been lowered. To obtain information on the cost and complexity associated with this ECRH-assisted, OH-pulsed startup voltage for ignition-type machines, a trade-off study was completed. The Fusion Engineering Device (FED) configuration was selected as a model because information was available on the structure. The data obtained are applicable to all tokamaks of this general size and complexity, such as the Engineering Test Reactor

  2. Completion of reconstruction for Japan Research Reactor No.3

    International Nuclear Information System (INIS)

    Kakefuda, K.; Tani, M.; Isshiki, M.

    1992-01-01

    The works of the reconstruction for the Japan Research Reactor No.3 (JRR-3) started in 1985 and initial criticality of the new reactor achieved in March, 1990. After commissioning test, the new JRR-3 has been operated some operational cycles since November, 1990. This paper presents outline of the removal work on the old JRR-3 and the new JRR-3. (author)

  3. Progress report on fast breeder reactor development in Japan

    International Nuclear Information System (INIS)

    1980-01-01

    The performance test on the reactor power increase to 75 MW was started on July 3, and the target of 75 MW was reached on July 16, in the experimental fast reactor Joyo. The tests on the heat transport characteristics, power coefficient, the response to the change of outlet temperature, the loss of external power supply and so on were carried out, and the performance test was finished on August 23, except the test of 75 MW continuous operation. The annual inspection of the systems is being carried out in parallel with the regular inspection. The design to prepare for the manufacture of the prototype fast reactor Monju is being prepared. The analysis of decay heat removal is being carried out, and various calculation codes were developed. The technological survey on overseas LMFBRs is being made. The conceptual design of the demonstration reactor is being prepared. The research and development of reactor physics, structural components for Joyo and Monju, instrumentation and control, sodium technology, fuel materials, structural materials, safety problems and steam generators are reported. The tests on the transient boiling of sodium, fuel failure propagation, heat transfer between molten materials, post-accident decay heat removal and so on have been carried out. (Kako, I.)

  4. Calibration of RB reactor power

    International Nuclear Information System (INIS)

    Sotic, O.; Markovic, H.; Ninkovic, M.; Strugar, P.; Dimitrijevic, Z.; Takac, S.; Stefanovic, D.; Kocic, A.; Vranic, S.

    1976-09-01

    The first and only calibration of RB reactor power was done in 1962, and the obtained calibration ratio was used irrespective of the lattice pitch and core configuration. Since the RB reactor is being prepared for operation at higher power levels it was indispensable to reexamine the calibration ratio, estimate its dependence on the lattice pitch, critical level of heavy water and thickness of the side reflector. It was necessary to verify the reliability of control and dosimetry instruments, and establish neutron and gamma dose dependence on reactor power. Two series of experiments were done in June 1976. First series was devoted to tests of control and dosimetry instrumentation and measurements of radiation in the RB reactor building dependent on reactor power. Second series covered measurement of thermal and epithermal neuron fluxes in the reactor core and calculation of reactor power. Four different reactor cores were chosen for these experiments. Reactor pitches were 8, 8√2, and 16 cm with 40, 52 and 82 fuel channels containing 2% enriched fuel. Obtained results and analysis of these results are presented in this document with conclusions related to reactor safe operation

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

    International Nuclear Information System (INIS)

    1981-09-01

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

  6. Tendencies in operating power reactors

    International Nuclear Information System (INIS)

    Brinckmann, H.F.

    1987-01-01

    A survey is given about new tendencies in operating power reactors. In order to meet the high demands for control and monitoring of power reactors modern procedures are applicated such as the incore-neutron flux detection by means of electron emission detectors and multi-component activation probes, the noise diagnostics as well as high-efficient automation systems

  7. Technical feasibility and costs of the retention of radionuclides during accidents in nuclear power plants demonstrated by the example of a pressurized water reactor

    International Nuclear Information System (INIS)

    Braun, H.; Grigull, R.; Lahner, K.; Gutowski, H.; Weber, J.

    1985-01-01

    The maximum allowable radiation doses during accidents in nuclear power plants, i.e., 5 rem whole-body dose and 15 rem thyroid dose, have been laid down in the German Radiation Protection Act. In order to ensure that these limits are not exceeded for all exposure paths including the ingestion path or, if possible, to remain far below them, the Federal Ministry of the Interior has initiated a study on the effectiveness and cost of additional safety features for reducing the release of activity and the dose exposure during accidents in nuclear power plants. Detailed investigations were carried out for the following three radiologically representative types of accidents: break of a reactor coolant line, break of an instrument line in one of the outer ring rooms, and break of a main stream line outside the containment. The technical basis of the study was a BBR-type nuclear power plant with pressurized water reactor and once-through steam generator. I-131 was chosen for determining the activity release as this is the critical nuclide for the ingestion path. Altogether 33 feasible technical measures were investigated and their potential improvement was assessed

  8. Tokamak experimental power reactor

    International Nuclear Information System (INIS)

    Stacey, W.M. Jr.; Abdou, M.A.; Bertoncini, P.J.

    1976-01-01

    A conceptual design has been developed for a tokamak Experimental Power Reactor to operate at net electrical power conditions with a plant capacity factor of 50 percent for 10 yr. The EPR operates in a pulsed mode at a frequency of approximately 1/min, with approximately 75 percent duty cycle, is capable of producing approximately 72 MWe and requires 42 MWe. The annual tritium consumption is 16 kg. The EPR vacuum chamber is 6.25 m in major radius and 2.4 m in minor radius, is constructed of 2 cm thick stainless steel, and has 2 cm thick detachable, beryllium-coated coolant panels mounted on the interior. A 0.28 m stainless steel blanket and a shield ranging from 0.6 to 1.0 m surround the vacuum vessel. The coolant is H 2 O. Sixteen niobium-titanium superconducting toroidal field coils provide a field of 10 T at the coil and 4.47 T at the plasma. Superconducting ohmic heating and equilibrium field coils provide 135 V-s to drive the plasma current. Plasma heating is accomplished by 12 neutral beam injectors which provide 60 MW. The energy transfer and storage system consists of a central superconducting storage ring, a homopolar energy storage unit, and a variety of inductor-convertors

  9. Construction of the advanced boiling water reactor in Japan

    International Nuclear Information System (INIS)

    Natsume, Nobuo; Noda, Hiroshi

    1996-01-01

    The Advanced Boiling Reactor (ABWR) has been developed with international cooperation between Japan and the US as the generation of plants for the 1990s and beyond. It incorporates the best BWR technologies from the world in challengeable pursuit of improved safety and reliability, reduced construction and operating cost, reduced radiation exposure and radioactive waste. Tokyo Electric Power Company (MPCO) decided to apply the first ABWRs to unit No. 6 and 7 of Kashiwazaki-Kariwa nuclear power station (K-6 and 7). These units are scheduled to commence commercial operation in December 1996 and July 1997 respectively. Particular attention is given in this discussion to the construction period from rock inspection for the reactor building to commercial operation, which is to be achieved in only 52 months through innovative and challenging construction methods. To date, construction work is advancing ahead of the original schedule. This paper describes not only how to shorten the construction period by adoption of a variety of new technologies, such as all-weather construction method and large block module construction method, but also how to check and test the state of the art technologies during manufacturing and installation of new equipment for K-6 and 7

  10. Seismic design of equipment and piping systems for nuclear power plants in Japan

    International Nuclear Information System (INIS)

    Minematsu, Akiyoshi

    1997-01-01

    The philosophy of seismic design for nuclear power plant facilities in Japan is based on 'Examination Guide for Seismic Design of Nuclear Power Reactor Facilities: Nuclear Power Safety Committee, July 20, 1981' (referred to as 'Examination Guide' hereinafter) and the present design criteria have been established based on the survey of governmental improvement and standardization program. The detailed design implementation procedure is further described in 'Technical Guidelines for Aseismic Design of Nuclear Power Plants, JEAG4601-1987: Japan Electric Association'. This report describes the principles and design procedure of the seismic design of equipment/piping systems for nuclear power plant in Japan. (J.P.N.)

  11. Seismic design of equipment and piping systems for nuclear power plants in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Minematsu, Akiyoshi [Tokyo Electric Power Co., Inc. (Japan)

    1997-03-01

    The philosophy of seismic design for nuclear power plant facilities in Japan is based on `Examination Guide for Seismic Design of Nuclear Power Reactor Facilities: Nuclear Power Safety Committee, July 20, 1981` (referred to as `Examination Guide` hereinafter) and the present design criteria have been established based on the survey of governmental improvement and standardization program. The detailed design implementation procedure is further described in `Technical Guidelines for Aseismic Design of Nuclear Power Plants, JEAG4601-1987: Japan Electric Association`. This report describes the principles and design procedure of the seismic design of equipment/piping systems for nuclear power plant in Japan. (J.P.N.)

  12. Status of core material development for fast reactor in Japan

    International Nuclear Information System (INIS)

    Ukai, S.; Shibahara, I.; Nagai, S.

    1994-01-01

    In the last two decades, extensive efforts have been devoted to the development of mixed-oxide fuel for LMFBR in Japan. For the fuel of the prototype reactor MONJU, drastic improvement in creep rupture strength and swelling resistance were attained by modification within the compositional specification of the standard Type 316 stainless steel (PNC316). For the fuel of future large-scale reactors, extensive research and development program are under way to realize the long life fuel. The candidate material for demonstration reactor is advanced austenitic stainless steel (PNC1520) which intended to modify the composition beyond the Type 316 stainless steel specification. In order to further improve the swelling resistance, the austenitic stainless steel with higher nickel content (High Ni alloy) and ferritic/martensitic steel (PNC-FMS) are developed. In a prospective cladding material for the long life fuel, the development of oxide dispersion strengthened (ODS) ferritic steel is focused to establish the alloying design and fabrication process toward as high as 250dpa. (author)

  13. Outline of geothermal power generation in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Ezaki, Y

    1960-01-01

    The utilization of geothermal energy in electrical power generation throughout the world is described. Details of generating capacity and cost are given for Larderello, Italy; Wairakei, New Zealand: and the Geysers, USA. In Japan three types of conversion systems are used. These include the direct use of steam, direct use of hot water and binary fluid type systems. The history of Japanese investigation and exploitation of geothermal energy is reviewed and the status of the Matsukawa, Hakone, Otake and Takenoyu geothermal power plants is discussed. It is recommended that laws be enacted in Japan to encourage the development of this form of energy conversion.

  14. Gaseous fuel reactors for power systems

    International Nuclear Information System (INIS)

    Helmick, H.H.; Schwenk, F.C.

    1978-01-01

    The Los Alamos Scientific Laboratory is participating in a NASA-sponsored program to demonstrate the feasibility of a gaseous uranium fueled reactor. The work is aimed at acquiring experimental and theoretical information for the design of a prototype plasma core reactor which will test heat removal by optical radiation. The basic goal of this work is for space applications, however, other NASA-sponsored work suggests several attractive applications to help meet earth-bound energy needs. Such potential benefits are small critical mass, on-site fuel processing, high fuel burnup, low fission fragment inventory in reactor core, high temperature for process heat, optical radiation for photochemistry and space power transmission, and high temperature for advanced propulsion systems. Low power reactor experiments using uranium hexafluoride gas as fuel demonstrated performance in accordance with reactor physics predictions. The final phase of experimental activity now in progress is the fabrication and testing of a buffer gas vortex confinement system

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

    International Nuclear Information System (INIS)

    1979-09-01

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

  16. The summary of the nuclear power development in Japan

    International Nuclear Information System (INIS)

    Mizoguchi, Kenzo; Hirose, Yasuo; Fukai, Yuzo; Hada, Mikio; Ogawa, Nagao.

    1980-01-01

    A quarter of century has elapsed since the development of atomic energy was started in Japan. At present, the scale of nuclear power generation reached the operation of 22 plants with about 15.12 million kW capacity, and 12% of the total installation capacity for power generation. Efforts have been exerted to bring up the domestic technologies gradually, while importing and digesting quickly the foreign technologies. Now in LWRs, the equipments of nearly 100% can be produced by the domestic technologies, moreover, the technologies have reached such level that they can be exported to foreign countries. In the last five years, the improvement and standardization of LWR technologies have been promoted. The development of the reactors of new types has been continued by the domestic technologies. According to the long term plan, the nuclear power generation of 53 million kW is expected by 1990, but various problems such as the location of nuclear power stations and nuclear fuel cycle remain, and considerable difficulty is expected in its materialization. The history of nuclear power generation in Japan, the features and progress of LWRs, the photographs and the main specifications of notable nuclear power plants, and the future perspectives of LWRs, the reactors of new types, nuclear fusion and nuclear fuel cycle are described. (Kako, I.)

  17. Spontaneous stabilization of HTGRs without reactor scram and core cooling—Safety demonstration tests using the HTTR: Loss of reactivity control and core cooling

    Energy Technology Data Exchange (ETDEWEB)

    Takamatsu, Kuniyoshi, E-mail: takamatsu.kuniyoshi@jaea.go.jp; Yan, Xing L.; Nakagawa, Shigeaki; Sakaba, Nariaki; Kunitomi, Kazuhiko

    2014-05-01

    It is well known that a High-Temperature Gas-cooled Reactor (HTGR) has superior safety characteristics; for example, an HTGR has a self-control system that uses only physical phenomena against various accidents. Moreover, the large heat capacity and low power density of the core result in very slow temperature transients. Therefore, an HTGR serves inherently safety features against loss of core cooling accidents such as the Tokyo Electric Power Co., Inc. (TEPCO)’s Fukushima Daiichi Nuclear Power Station (NPS) disaster. Herein we would like to demonstrate the inherent safety features using the High-Temperature Engineering Test Reactor (HTTR). The HTTR is the first HTGR in Japan with a thermal power of 30 MW and a maximum reactor outlet coolant temperature of 950 °C; it was built at the Oarai Research and Development Center of Japan Atomic Energy Agency (JAEA). In this study, an all-gas-circulator trip test was analyzed as a loss of forced cooling (LOFC) test with an initial reactor power of 9 MW to demonstrate LOFC accidents. The analytical results indicate that reactor power decreases from 9 MW to 0 MW owing to the negative reactivity feedback effect of the core, even if the reactor shutdown system is not activated. The total reactivity decreases for 2–3 h and then gradually increases in proportion to xenon reactivity; therefore, the HTTR achieves recritical after an elapsed time of 6–7 h, which is different from the elapsed time at reactor power peak occurrence. After the reactor power peak occurs, the total reactivity oscillates several times because of the negative reactivity feedback effect and gradually decreases to zero. Moreover, the new conclusions are as follows: the greater the amount of residual heat removed from the reactor core, the larger the stable reactor power after recriticality owing to the heat balance of the reactor system. The minimum reactor power and the reactor power peak occurrence are affected by the neutron source. The greater the

  18. Conceptual design of the JAERI demonstration fusion reactor

    International Nuclear Information System (INIS)

    Sako, K.; Tone, T.; Seki, Y.

    1976-01-01

    Conceptual design of a tokamak demonstration fusion reactor is carried out. This design is an extended and improved version of the previous design which was presented at the 5th IAEA Conference. The main design parameters are as follows: the reactor thermal power 2000 MW, torus radius 10.5 m, plasma radius 2.7 m, first wall radius 3.0 m, toroidal magnetic field on axis 6T, blanket fertile material Li 2 O, coolant He, structural material Mo-alloy and tritium breeding ratio 1.2

  19. A review of fast reactor progress in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Tomabechi, K [Power Reactor and Nuclear Fuel Development Corporation, Tokyo (Japan)

    1978-07-01

    The fast reactor development project in Japan is continuing at a slightly increased scale of effort in budget. The total budget for LMFBR development for fiscal year 1978 was 24 billion yen. In August 1977 major industries engaged in LMFBR have set up an office where design work can be jointly conducted. Highlights and topics of the fast reactor development activities cover description of JOYO reactor, its first criticality experiment, and the prototype fast breeder MONJU. Research and development programmes dealt with fission products release and its possible interaction with the soodium coolant, inspection of reactor components, experiments simulating sodium leakage, development of steam generator.

  20. 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

  1. A nuclear power reactor

    International Nuclear Information System (INIS)

    Borrman, B.E.; Broden, P.; Lundin, N.

    1979-12-01

    The invention consists of shock absorbing support beams fastened to the underside of the reactor tank lid of a BWR type reactor, whose purpose is to provide support to the steam separator and dryer unit against accelerations due to earthquakes, without causing undue thermal stresses in the unit due to differential expansion. (J.I.W.)

  2. Conceptual design report on advanced marine reactor MRX of Japan

    International Nuclear Information System (INIS)

    Wang Shengguo

    1995-01-01

    Design studies on the advanced marine reactors have been done continuously since 1983 at Japan Atomic Energy Institute (JAERI) in order to develop attractive marine reactors for the next generation. At present, two concepts of marine reactor are being formulated. One is 100 MWt MRX (marine Reactor X) for the marine reactor and the other is 150 kWe DRX (Deep Sea-Reactor X) for a deep-sea research vessel. They are characterized by an integral type PWR, built-type control rod drive mechanisms, a water-filled container and a passive decay heat removal system, which realize highly passive safe and compact reactors. The paper is a report about all major results of the MRX design study

  3. Demonstration tokamak-power-plant study (DEMO)

    International Nuclear Information System (INIS)

    1982-09-01

    A study of a Demonstration Tokamak Power Plant (DEMO) has been completed. The study's objective was to develop a conceptual design of a prototype reactor which would precede commercial units. Emphasis has been placed on defining and analyzing key design issues and R and D needs in five areas: noninductive current drivers, impurity control systems, tritium breeding blankets, radiation shielding, and reactor configuration and maintenance features. The noninductive current drive analysis surveyed a wide range of candidates and selected relativistic electron beams for the reference reactor. The impurity control analysis considered both a single-null poloidal divertor and a pumped limiter. A pumped limiter located at the outer midplane was selected for the reference design because of greater engineering simplicity. The blanket design activity focused on two concepts: a Li 2 O solid breeder with high pressure water cooling and a lead-rich Li-Pb eutectic liquid metal breeder (17Li-83Pb). The reference blanket concept is the Li 2 O option with a PCA structural material. The first wall concept is a beryllium-clad corrugated panel design. The radiation shielding effort concentrated on reducing the cost of bulk and penetration shielding; the relatively low-cost outborad shield is composed of concrete, B 4 C, lead, and FE 1422 structural material

  4. Statement to International Conference on Fast Reactors and Related Fuel Cycles: Challenges and Opportunities, 7 December 2009, Kyoto, Japan

    International Nuclear Information System (INIS)

    Amano, Yukiya

    2009-01-01

    Full text: Distinguished Guests, Ladies and Gentlemen, It is my honour to address participants at this opening session of the International Conference on Fast Reactors and Related Fuel Cycles: Challenges and Opportunities, organized by the IAEA and hosted by the Japan Atomic Energy Agency. Fast reactor technology has the potential to ensure that energy resources which would last hundreds of years with the technology we are using today will actually last several thousand years. In other words it can withstand enormous increases in demand. This innovative technology also reduces the risk to the environment and helps to limit the burden that will be placed on future generations in the form of waste products. The coming year will be an exciting one for the development of fast-spectrum nuclear reactors. We expect to reach many important milestones: - the first criticality of the China Experimental Fast Reactor; - the restart of the Monju prototype fast reactor in Japan; and - the new insights we will gain through the end-of-life studies at the Phenix reactor in France. In the near future, new fast reactors will be commissioned: the 500MW(e) Prototype Fast Breeder Reactor in India, the first in a series of five of the same type, and the BN-800 reactor in the Russian Federation. Moreover, France, Japan, India, China and the Republic of Korea are preparing advanced prototypes, demonstration or commercial reactors for the 2020-2030 period. Nuclear power is set to be an increasingly important part of the global energy mix in the coming decades as demand for energy grows. Scores of countries in both the developed and developing world have told the IAEA that they are interested in introducing nuclear power. The 30 countries which already have nuclear power reactors are set to build more. This trend is likely to be accompanied by accelerated deployment of fast reactors. Continued advances in research and technology development are necessary to ensure improved economics and

  5. Status of reactor pressure vessel embrittlement study in Japan

    International Nuclear Information System (INIS)

    Sasajima, H.

    1997-01-01

    Since the construction of Japanese first commercial nuclear power plant in 1966, 52 nuclear power plants have been commissioned in Japan to commercial operation. Japanese first nuclear power plant has now been service for 30 years and the aging of nuclear power plants is steadily progressing in general. Under these circumstances, the Japan Power Engineering and Inspection Corporation (JAPEIC) is executing, under consignment by the Ministry of International Trade and Industry (MITI), the development and verification test programs for plant integrity evaluation technology by which nuclear power plant aging can be appropriately handled. This paper shows the outline of study dealing with embrittlement of RPV caused by neutron irradiation, as one of the activity of JAPEIC. The embrittlement of RPV caused by neutron irradiation is manifested as a shift of transition temperature and as a reduction in Upper Shelf Energy (USE). In JAPEIC, the study dealing with a shift of transition temperature was conducted in the ''Reactor Pressure Vessel Pressurized Thermal Shock Test Project (the PTS Project)'', and the study dealing with a reduction in USE has been conducted in the ''Nuclear Power Plant Life Management Technology (the PLIM Project)''. And the reconstitution technology of surveillance test specimen has been conducted in PLIM Project as one of the measures to improve monitoring above material characteristic changes. The integrity evaluation under the Pressurized Thermal Shock (PTS) events including the effect of neutron irradiation embrittlement was initiated in 1983 FY as the PTS Project and was completed in the 1991 FY. The study verified that plant integrity could be assured at not only the end of design life, but also an extended service life even when the severest PTS events were postulated. The PLIM Project, designed to develop and verify the integrity evaluation technology dealing with reduction of USE by neutron irradiation, was started in the 1996 FY as a 10

  6. Progress report on fast breeder reactor development in Japan

    International Nuclear Information System (INIS)

    1979-01-01

    The experimental fast reactor ''Joyo'' will be tested at 75 MW output, starting in April, 1980. In connection with the accident in the Three Mile Island plant, the reexamination of the plant safety and the rechecking-up of the maintenance control system were carried out, and the special inspection by the Science and Technology Agency was executed from May 21 to 23, 1979. Thereafter, the preparation for raising the power output was completed. The periodical inspection after the completion of 50 MW operation is being carried out. The state of progress of various equipments and the codes for core characteristic analysis is reported. The construction preliminary design (2) of the prototype reactor ''Monju'' is examined, and the same design (3) is prepared. The analysis of the decay heat in the prototype reactor is carried on for the safety licensing. The technological investigation of LMFBRs in foreign countries is under way. The preliminary design (4) of the demonstration reactor is under examination, and the technical specifications of the conceptual design (1) are prepared. The researches and developments of reactor physics, structural components, instrumentation and control, sodium technology, fuel materials, structures and materials, safety and steam generators are reported. (Kako, I.)

  7. Reactor theory and power reactors. 1. Calculational methods for reactors. 2. Reactor kinetics

    International Nuclear Information System (INIS)

    Henry, A.F.

    1980-01-01

    Various methods for calculation of neutron flux in power reactors are discussed. Some mathematical models used to describe transients in nuclear reactors and techniques for the reactor kinetics' relevant equations solution are also presented

  8. Development of Next-Generation LWR (Light Water Reactor) in Japan

    International Nuclear Information System (INIS)

    Yamamoto, T.; Kasai, S.

    2011-01-01

    The Next-Generation Light Water Reactor development program was launched in Japan in April 2008. The primary objective of the program is to cope with the need to replace existing nuclear power plants in Japan after 2030. The reactors to be developed are also expected to be a global standard design. Several innovative features are envisioned, including a reactor core system with uranium enrichment above 5%, a seismic isolation system, the use of long-life materials and innovative water chemistry, innovative construction techniques, safety systems with the best mix of passive and active concepts, and innovative digital technologies to further enhance reactor safety, reliability, economics, etc. In the first 3 years, a plant design concept with these innovative features is established and the effectiveness of the program is reevaluated. The major part of the program will be completed in 2015. (author)

  9. Fractals in Power Reactor Noise

    International Nuclear Information System (INIS)

    Aguilar Martinez, O.

    1994-01-01

    In this work the non- lineal dynamic problem of power reactor is analyzed using classic concepts of fractal analysis as: attractors, Hausdorff-Besikovics dimension, phase space, etc. A new non-linear problem is also analyzed: the discrimination of chaotic signals from random neutron noise signals and processing for diagnosis purposes. The advantages of a fractal analysis approach in the power reactor noise are commented in details

  10. Fiscal 1999 research result report on energy and environment technology demonstration research support project (International joint demonstration research project). Japan- Russia joint demonstration research on large-capacity long- distant DC power transmission technology; 1999 nendo daiyoryo denryoku no chokyori chokuryu soden gijutsu ni kansuru Russia kenkyu kikan tono kyodo jissho kenkyu seika hokokusho

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-03-01

    Under the assumption of a large-capacity DC power transmission model project in the Far East and Siberia districts, technical study was made on the basic design of the project, considering selection of concrete power generation and consumption sites and power transmission routes, power transmission/transformation equipment, cables, and environmental impact. Study was also made on its applicability to similar projects in Japan. The model project aims at integration of Eastern Integrated Power System in the Far East and Russian Unified Power System, and development of abundant undeveloped hydraulic and tidal power generation in the Far East. The study result showed that (1) construction of the high-voltage DC power transmission (HVDC) system of model project class voltage and capacity in Eastern Siberia is possible technically enough, (2) the total construction cost of the model project scheduled to be put into operation in 2025 amounts to nearly $4.7 billion, and (3) the model project is environment-friendly without any CO{sub 2} gas emission because of hydraulic and tidal power generation. (NEDO)

  11. Power reactors in Member States. 1978 edition

    International Nuclear Information System (INIS)

    1978-01-01

    The computer-based reactor listing gives information on reactor core characteristics and plant systems for all power reactors in operation under construction and planned. The following two tables are included to give a general picture of the overall situation: Reactor types and net electrical power; Reactor units and net electrical power by country and cumulated by year

  12. Power reactor information system (PRIS)

    International Nuclear Information System (INIS)

    1989-06-01

    Since the very beginning of commercial operation of nuclear power plants, the nuclear power industry worldwide has accumulated more than 5000 reactor years of experience. The IAEA has been collecting Operating Experience data for Nuclear Power Plants since 1970 which were computerized in 1980. The Agency has undertaken to make Power Reactor Information System (PRIS) available on-line to its Member States. The aim of this publication is to provide the users of PRIS from their terminals with description of data base and communication systems and to show the methods of accessing the data

  13. Electric-power economy of Japan

    International Nuclear Information System (INIS)

    Dobrochotov, V.I.; Wolfberg, D.B.

    1975-01-01

    This is a survey on a) development and present capacity of electricity-supply companies in Japan, b) the structural shift in the capacity of power plants which took place from 1966 until 1974, arranged according to thermal, nuclear and hydraulic power stations, c) the structural shift in the use of fossile fuels, also from 1966 until 1974, d) the major thermal and nuclear power stations and pump storage plants under construction and in operation, e) interconnected operation. The survey ends with the development study of the Japanese Government being outlined. (GG/LN) [de

  14. Power oscillations in BWR reactors

    International Nuclear Information System (INIS)

    Espinosa P, G.

    2002-01-01

    One of the main problems in the operation of BWR type reactors is the instability in power that these could present. One type of oscillations and that is the objective of this work is the named density wave, which is attributed to the thermohydraulic processes that take place in the reactor core. From the beginnings of the development of BWR reactors, the stability of these has been an important aspect in their design, due to its possible consequences on the fuel integrity. The reactor core operates in two phase flow conditions and it is observed that under certain power and flow conditions, power instabilities appear. Studying this type of phenomena is complex, due to that a reactor core is constituted approximately by 27,000 fuel bars with different distributions of power and flow. The phenomena that cause the instability in BWR reactors continue being matter of scientific study. In the literature mainly in nuclear subject, it can be observed that exist different methods and approximations for studying this type of phenomena, nevertheless, their results are focused to establish safety limits in the reactor operation, instead of studying in depth of the knowledge about. Also in this line sense of the reactor data analysis, the oscillations characteristic frequencies are obtained for trying to establish if the power is growing or decreasing. In addition to that before mentioned in this paper it is presented a rigorous study applying the volumetric average method, for obtaining the vacuum waves propagation velocities and its possible connection with the power oscillations. (Author)

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

    International Nuclear Information System (INIS)

    Gauche, F.

    2013-01-01

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

  16. The integral fast reactor and its role in a new generation of nuclear power plants, Tokai, Japan, November 19-21, 1986

    International Nuclear Information System (INIS)

    Smith, R.R.

    1986-01-01

    This report presents information on the Integral Fast Reactor and its role in the future. Information is presented in the areas of: inherent safety; other virtues of sodium-cooled breeder; and solving LWR fuel cycle problems with IFR technologies

  17. Present status and future outlook of nuclear power generation in Japan

    International Nuclear Information System (INIS)

    Kunikazu Aisaka

    1987-01-01

    The structure of energy consumption in Japan is heavily dependent on imported oil, therefore Japan has been making its greatest effort in developing nuclear power among other alternatives of oil. The capacity factor of the nuclear power plants in Japan marked 76% in FY 1986, exceeding 70% level for the past several years. The share of nuclear power is expected to increase steadily in the future. Future scale of the nuclear power generation is projected as 62,000 MW in year 2000 and as 137,000 MW in 2030. Nuclear power is expected to produce 58% of the nation's total power generation in 2030. Under the present circumstances, Janpan is executing a nuclear energy policy based on the following guidelines: 1. Promoting the safety advancement program; 2. Improving LWR technologies; 3. Program on use of plutonium in thermal reactors; 4. Advanced thermal reactors (ATRs); 5. Promotion of FBR development; 6. Nuclear fuel cycle. (Liu)

  18. Performance indicators for power reactors

    International Nuclear Information System (INIS)

    Gillies, C.; White, M.

    1995-11-01

    A review of Canadian and worldwide performance indicator definitions and data was performed to identify a set of indicators that could be used for comparison of performance among nuclear power plants. The results of this review are to be used as input to an AECB team developing a consistent set of performance indicators for measuring Canadian power reactor safety performance. To support the identification of performance indicators, a set of criteria was developed to assess the effectiveness of each indicator for meaningful comparison of performance information. The project identified a recommended set of performance indicators that could be used by AECB staff to compare the performance of Canadian nuclear power plants among themselves, and with international performance. The basis for selection of the recommended set and exclusion of others is provided. This report provides definitions and calculation methods for each recommended performance indicator. In addition, a spreadsheet has been developed for comparison and trending for the recommended set of indicators. Example trend graphs are included to demonstrate the use of the spreadsheet. (author). 50 refs., 11 tabs., 3 figs

  19. Power Reactor Information System (PRIS)

    International Nuclear Information System (INIS)

    Spiegelberg, R.

    1992-01-01

    The IAEA has been collecting Operating Experience data for Nuclear Power Plants of the IAEA Member States since 1970. In order to facilitate an analysis of nuclear power plant performance as well as to produce relevant publications, all previously collected data supplied from the questionnaires were computerized in 1980 and the Power Reactor Information System was implemented. PRIS currently contains production records for the years up to and including 1990 and about 98% of the reactors-years operating experience in the world is contained in PRIS. (orig.)

  20. The status and prospects for the fossil-fired and nuclear power industry in Japan

    International Nuclear Information System (INIS)

    Miyahara, S.

    1994-01-01

    Power plant capacity in Japan amounts to about 200 GW, of which 180 GW belong to the electricity supply industry. 60% are installed in fossil-fired power stations, 19% in nuclear power stations and 21% in hydro-electric power stations. Key engineering techniques for power production from fossil fuels are supercritical steam conditions and combined cycle power plant technology. Crucial points for nuclear power generation are the development of the advanced light water reactor, the commericialization of the fast breeder reactor and the installation of a closed nuclear fuel cycle. (orig.) [de

  1. Hybrid reactors: recent progress of a demonstration pilot

    International Nuclear Information System (INIS)

    Billebaud, Annick

    2006-12-01

    Accelerator driven sub-critical reactors are subject of many research programmes since more than ten years, with the aim of testing the feasibility of the concept as well as their efficiency as a transmutation tool. Several key points like the accelerator, the spallation target, or neutronics in a subcritical medium were investigated extensively these last years, allowing for technological choices and the design of a low power European demonstration ADS (a few tens of MWth). Programmes dedicated to subcritical reactor piloting proposed a monitoring procedure to be validated in forthcoming experiments. Accelerator R and D provided the design of a LINAC for an ADS and research work on accelerator reliability is going on. A spallation target was operated at PSI and the design of a windowless target is in progress. All this research work converges to the design of a European demonstration ADS, the ETD/XT-ADS, which could be the Belgian MYRRHA project. (author)

  2. Revision of construction plan for advanced thermal demonstration reactor

    International Nuclear Information System (INIS)

    1996-01-01

    The Federation of Electric Power Companies demanded the revision of the construction plan for the advanced thermal demonstration reactor, which is included in the 'Long term plan on the research, development and utilization of atomic energy' decided by the Atomic Energy Commission in 1994, for economical reason. The Atomic Energy Commission carried out the deliberation on this demand. It was found that the cost of construction increases to 580 billion yen, and the cost of electric power generation increases three times as high as that of LWRs. The role as the reactor that utilizes MOX fuel can be substituted by LWRs. The relation of trust with the local town must be considered. In view of these circumstances, it is judged that the stoppage of the construction plan is appropriate. It is necessary to investigate the substitute plan for the stoppage, and the viewpoints of investigating the substitute plan, the examination of the advanced BWR with all MOX fuel core and the method of advancing its construction are considered. On the research and development related to advanced thermal reactors, the research and development contributing to the advance of nuclear fuel recycling are advanced, and the prototype reactor 'Fugen' is utilized. (K.I.)

  3. Fusion power demonstration: baseline report

    International Nuclear Information System (INIS)

    Henning, C.D.; Logan, B.G.; Neef, W.S.

    1984-01-01

    An overview is given of the FPD project. A technical discussion of the primary independent plasma parameters, the physics constraints that limited the allowed combinations of those parameters for ignition, and the basic plasma scaling laws that govern FPD plasma parameters are described. Plans are detailed for phased construction and operation of FPD. The components for FPD are based on those designed for the Mirror Advanced Reactor Study (MARS). Although FPD costs are based on the MARS first-of-a-kind estimates, FPD contains design modifications made to reduce these costs. The preliminary results of siting and safety studies are described. An essential element of FPD design is the ability to perform remote maintenance on the equipment. The key research and development issues associated with the design and implementation of an FPD facility are highlighted

  4. 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)

  5. The outlook for nuclear power development in Japan

    International Nuclear Information System (INIS)

    Hiraiwa, Gaishi

    1987-01-01

    The world economy has entered a new stage of growth--albeit low growth--following painful adjustments in the wake of past oil crises. At the same time, energy demand is expanding at an even slower rate, due to the structural changes in industry and improved efficiency in energy use. Furthermore, progress in the development of alternative energies and technical innovations in both the supply and use of energy have sharpened competition between energy sources. We also aim to improve even further the economy of nuclear power, within the bounds of safety and reliability, to minimize electric power generation costs by optimizing the total system for nuclear power generation including the nuclear fuel cycle. In Japan's long-term strategy for the development of nuclear power, our basic plan is to switch from light-water reactors to fast-breeder reactors (FBR), as the latter use plutonium most efficiently. Every effort is being made to have FBR reactors up and running at an early date. However, given the outlook for the development of their technology and the supply and demand situation for uranium, we estimate that this won't be achieved until 2020 or 2030. With this timetable in mind, it will be important to prepare for the coming age of FBR by mastering the technologies of and establishing the foundation necessary for plutonium utilization. To this end, we plan to expand our use of plutonium to an appropriate scale, at the earliest possible date. (J.P.N.)

  6. To question of NPP power reactor choice for Kazakhstan

    International Nuclear Information System (INIS)

    Batyrbekov, G.A.; Makhanov, Y.M.; Reznikova, R.A.; Sidorenco, A.V.

    2004-01-01

    in the regions of Republic of Kazakhstan occurs. Southern and western regions import electric power and capacity because of undeveloped circuit of networks. Moreover, power intensity of an industrial-agrarian complex of the country is limited transmission capacity of lines is insufficient; plenty of small consumers are removed from power supply lines. Thus, nuclear stations of medium and low power are the most acceptable for construction in Kazakhstan. Recommendations for the choice of maximum safe, reliable and economically competitive reactors for Kazakhstan have been made in result of the carried out projects' comparison of the power reactors according to 15 criteria of safety and economic competitiveness, with respect to condition and perspectives of Kazakhstan power complex development: Recommended power reactors of medium capacity: - P-600 - passive PWR, developed of the Westinghouse company, USA; - CANDU-6, developed by Atomic Energy of Canada, Limited (AECL), Canada. - MS-600 - Mitsubishi Company, Japan. Recommended reactors of low power: - IRIS - reactor of IV generation developed by the international corporation of 13 organizations from 7 countries; - NPP 'UNITERM' - development NIKIET, Moscow, Russia; - MRX - the Project of sea reactor MRX for civil applications, is developed by the Japanese Research Institute of Atomic Energy (JAERI). The most important advantages of recommended medium and low power reactors are given

  7. To question of NPP power reactor choice for Kazakhstan

    International Nuclear Information System (INIS)

    Batyrbekov, G.A.; Makhanov, Y.M.; Reznikova, R.A.; Sidorenco, A.V.

    2004-01-01

    regions of Republic of Kazakhstan occurs. Southern and western regions import electric power and capacity because of undeveloped circuit of networks. Moreover, power intensity of an industrial-agrarian complex of the country is limited transmission capacity of lines is insufficient; plenty of small consumers are removed from power supply lines. Thus, nuclear stations of medium and low power are the most acceptable for construction in Kazakhstan. Recommendations for the choice of maximum safe, reliable and economically competitive reactors for Kazakhstan have been made in result of the carried out projects' comparison of the power reactors according to 15 criteria of safety and economic competitiveness, with respect to condition and perspectives of Kazakhstan power complex development: Recommended power reactors of medium capacity: - P-600 - passive PWR, developed of the Westinghouse company, USA; - CANDU-6, developed by Atomic Energy of Canada, Limited (AECL), Canada. - MS-600 - Mitsubishi Company, Japan. Recommended reactors of low power: - IRIS - reactor of IV generation developed by the international corporation of 13 organizations from 7 countries; - NPP 'UNITERM' - development NIKIET, Moscow, Russia; - MRX - the Project of sea reactor MRX for civil applications, is developed by the Japanese Research Institute of Atomic Energy (JAERI). The most important advantages of recommended medium and low power reactors are given

  8. Performance demonstration experience for reactor pressure vessel shell ultrasonic testing

    International Nuclear Information System (INIS)

    Zado, V.

    1998-01-01

    The most ultrasonic testing techniques used by many vendors for pressurized water reactor (PWR) examinations were based on American Society of Mechanical Engineers 'Boiler and Pressurized Vessel Code' (ASME B and PV Code) Sections XI and V. The Addenda of ASME B and PV Code Section XI, Edition 1989 introduced Appendix VIII - 'Performance Demonstration for Ultrasonic Examination Systems'. In an effort to increase confidence in performance of ultrasonic testing of the operating nuclear power plants in United States, the ultrasonic testing performance demonstration examination of reactor vessel welds is performed in accordance with Performance Demonstration Initiative (PDI) program which is based on ASME Code Section XI, Appendix VIII requirements. This article provides information regarding extensive qualification preparation works performed prior EPRI guided performance demonstration exam of reactor vessel shell welds accomplished in January 1997 for the scope of Appendix VIII, Supplements IV and VI. Additionally, an overview of the procedures based on requirements of ASME Code Section XI and V in comparison to procedure prepared for Appendix VIII examination is given and discussed. The samples of ultrasonic signals obtained from artificial flaws implanted in vessel material are presented and results of ultrasonic testing are compared to actual flaw sizes. (author)

  9. New facilities in Japan materials testing reactor for irradiation test of fusion reactor components

    International Nuclear Information System (INIS)

    Kawamura, H.; Sagawa, H.; Ishitsuka, E.; Sakamoto, N.; Niiho, T.

    1996-01-01

    The testing and evaluation of fusion reactor components, i.e. blanket, plasma facing components (divertor, etc.) and vacuum vessel with neutron irradiation is required for the design of fusion reactor components. Therefore, four new test facilities were developed in the Japan Materials Testing Reactor: an in-pile functional testing facility, a neutron multiplication test facility, an electron beam facility, and a re-weldability facility. The paper describes these facilities

  10. The integral fast reactor and its role in a new generation of nuclear power plants, Tokai, Japan, November 19-21, 1986

    Energy Technology Data Exchange (ETDEWEB)

    Smith, R.R.

    1986-01-01

    This report presents information on the Integral Fast Reactor and its role in the future. Information is presented in the areas of: inherent safety; other virtues of sodium-cooled breeder; and solving LWR fuel cycle problems with IFR technologies. (JDB)

  11. A review of fast reactor program in Japan. April 1997 - March 1998

    International Nuclear Information System (INIS)

    1998-01-01

    This report describes the development and activities on fast reactor in Japan for the period of April 1997 - March 1998. During this period, two important results were drawn by the Special Committee on Fast Breeder Reactors (FBRs) and the Power Reactor and Nuclear Fuel Development Corporation (PNC) Reform Committee, respectively. The Special Committee on FBRs discussed on the future FBR development in Japan including the Prototype FBR 'Monju' operation, and proposed its conclusion as the final report to the Japan Atomic Energy Commission (JAEC) on December 1, 1997. The PNC Reform Committee reviewed PNC's management and safety assurance system, and recommended to reform PNC to a new organization. Each committee result is outlined in this report. The Experimental Fast Reactor 'Joyo' operated 30th - 32nd cycle. In parallel with the operation, the Joyo Upgrading Program (MK-III program) is in progress. Five MK-III driver fuel subassemblies were loaded to the core in the 32nd cycle. Monju comprehensive safety review, which was started in December 1996, was continued through 1997, and was completed in March 1998. The DFBR Plant Optimization (phase 2) design study was launched by the Japan Atomic Power Company (JAPC) with goal of constructing FBR plant that achieves both reliability and economy from FY 1997 for three years. Research and development works are underway under the discussion and coordination of the Japanese FBR R and D Steering Committee, which is composed of PNC, JAPC, Japan Atomic Energy Research Institute (JAERI) and Central Research Institute of Electric Power Industry (CRIEPI). (author)

  12. Cascade ICF power reactor

    International Nuclear Information System (INIS)

    Hogan, W.J.; Pitts, J.H.

    1986-01-01

    The double-cone-shaped Cascade reaction chamber rotates at 50 rpm to keep a blanket of ceramic granules in place against the wall as they slide from the poles to the exit slots at the equator. The 1 m-thick blanket consists of layers of carbon, beryllium oxide, and lithium aluminate granules about 1 mm in diameter. The x rays and debris are stopped in the carbon granules; the neutrons are multiplied and moderated in the BeO and breed tritium in the LiAlO 2 . The chamber wall is made up of SiO tiles held in compression by a network of composite SiC/Al tendons. Cascade operates at a 5 Hz pulse rate with 300 MJ in each pulse. The temperature in the blanket reaches 1600 K on the inner surface and 1350 K at the outer edge. The granules are automatically thrown into three separate vacuum heat exchangers where they give up their energy to high pressure helium. The helium is used in a Brayton cycle to obtain a thermal-to-electric conversion efficiency of 55%. Studies have been done on neutron activation, debris recovery, vaporization and recondensation of blanket material, tritium control and recovery, fire safety, and cost. These studies indicate that Cascade appears to be a promising ICF reactor candidate from all standpoints. At the 1000 MWe size, electricity could be made for about the same cost as in a future fission reactor

  13. An inherently safe power reactor module

    International Nuclear Information System (INIS)

    Salerno, L.N.

    1985-01-01

    General Electric's long participation in liquid metal reactor technology has led to a Power Reactor Inherently Safe Module (PRISM) concept supported by DOE contract DE-AC06-85NE37937. The reactor module is sized to maximize inherent safety features. The small size allows factory fabrication, reducing field construction and field QA/QC labor, and allows safety to be demonstrated in full scale, to support a pre-licensed standard commercial product. The module is small enough to be placed underground, and can be combined with steam and electrical generating equipment to provide a complete electrical power producing plant in the range of 400-1200 MWe. Initial assessments are that the concept has the potential to be economically competitive with existing methods of power production used by the utility industry

  14. Nuclear reactor power control device

    International Nuclear Information System (INIS)

    Koshi, Yuji; Sakata, Akira; Karatsu, Hiroyuki.

    1987-01-01

    Purpose: To control abrupt changes in neutron fluxes by feeding back a correction signal obtained from a deviation between neutron fluxes and heat fluxes for changing the reactor core flow rate to a recycling flow rate control system upon abrupt power change of a nuclear reactor. Constitution: In addition to important systems, that is, a reactor pressure control system and a recycling control system in the power control device of a BWR type power plant, a control circuit for feeding back a deviation between neutron fluxes and heat fluxes to a recycling flow rate control system is disposed. In the suppression circuit, a deviation signal is prepared in an adder from neutron flux and heat flux signals obtained through a primary delay filter. The deviation signal is passed through a dead band and an advance/delay filter into a correction signal, which is adapted to be fed back to the recycling flow rate control system. As a result, the reactor power control can be conducted smoothly and it is possible to effectively suppress the abrupt change or over shoot of the neutron fluxes and abrupt power change. (Kamimura, M.)

  15. Development of fresh fuel packaging for ATR demonstration reactor

    International Nuclear Information System (INIS)

    Kurakami, J.; Kurita, I.

    1993-01-01

    Related to development of the demonstration advanced thermal reactor, it is necessary and important to develop transport packaging which is used for transporting fresh fuel assemblies. Therefore, the packaging is now being developed in Power Reactor and Nuclear Fuel Development Corporation (PNC). Currently, PNC is fabricating two prototype packagings based on the final design, and land cruising and vibration tests, handling performance tests and prototype packaging tests will be executed with prototype packagings in order to experimentally confirm the soundness of packaging and its contents and the propriety of design technique. This paper describes the summary of general specifications and structures of this packaging and the summary of preliminary safety analysis of package. (J.P.N.)

  16. Power control system in BWR type reactors

    International Nuclear Information System (INIS)

    Nishizawa, Yasuo.

    1980-01-01

    Purpose: To control the reactor power so that the power distribution can satisfy the limiting conditions, by regulating the reactor core flow rate while monitoring the power distribution in the reactor core of a BWR type reactor. Constitution: A power distribution monitor determines the power distribution for the entire reactor core based on the data for neutron flux, reactor core thermal power, reactor core flow rate and control rod pattern from the reactor and calculates the linear power density distribution. A power up ratio computing device computes the current linear power density increase ratio. An aimed power up ratio is determined by converting the electrical power up ratio transferred from a load demand input device into the reactor core thermal power up ratio. The present reactor core thermal power up ratio is subtracted from the limiting power up ratio and the difference is sent to an operation amount indicator and the reactor core flow rate is changed in a reactor core flow rate regulator, by which the reactor power is controlled. (Moriyama, K.)

  17. Seismic resistance design of nuclear power plant building structures in Japan

    International Nuclear Information System (INIS)

    Kitano, Takehito

    1997-01-01

    Japan is one of the countries where earthquakes occur most frequently in the world and has incurred a lot of disasters in the past. Therefore, the seismic resistance design of a nuclear power plant plays a very important role in Japan. This report describes the general method of seismic resistance design of a nuclear power plant giving examples of PWR and BWR type reactor buildings in Japan. Nuclear facilities are classified into three seismic classes and is designed according to the corresponding seismic class in Japan. Concerning reactor buildings, the short-term allowable stress design is applied for the S1 seismic load and it is confirmed that the structures have a safety margin against the S2 seismic load. (J.P.N.)

  18. Seismic resistance design of nuclear power plant building structures in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Kitano, Takehito [Kansai Electric Power Co., Inc., Osaka (Japan)

    1997-03-01

    Japan is one of the countries where earthquakes occur most frequently in the world and has incurred a lot of disasters in the past. Therefore, the seismic resistance design of a nuclear power plant plays a very important role in Japan. This report describes the general method of seismic resistance design of a nuclear power plant giving examples of PWR and BWR type reactor buildings in Japan. Nuclear facilities are classified into three seismic classes and is designed according to the corresponding seismic class in Japan. Concerning reactor buildings, the short-term allowable stress design is applied for the S1 seismic load and it is confirmed that the structures have a safety margin against the S2 seismic load. (J.P.N.)

  19. Low power unattended defense reactor

    International Nuclear Information System (INIS)

    Kirchner, W.L.; Meier, K.L.

    1984-01-01

    A small, low power, passive, nuclear reactor electric power supply has been designed for unattended defense applications. Through innovative utilization of existing proven technologies and components, a highly reliable, ''walk-away safe'' design has been obtained. Operating at a thermal power level of 200 kWt, the reactor uses low enrichment uranium fuel in a graphite block core to generate heat that is transferred through heat pipes to a thermoelectric (TE) converter. Waste heat is removed from the TEs by circulation of ambient air. Because such a power supply offers the promise of minimal operation and maintenance (OandM) costs as well as no fuel logistics, it is particularly attractive for remote, unattended applications such as the North Warning System

  20. Low power unattended defense reactor

    International Nuclear Information System (INIS)

    Kirchner, W.L.; Meier, K.L.

    1984-01-01

    A small, low power, passive, nuclear reactor electric power supply has been designed for unattended defense applications. Through innovative utilization of existing proven technologies and components, a highly reliable, walk-away safe design has been obtained. Operating at a thermal power level of 200 kWt, the reactor uses low enrichment uranium fuel in a graphite block core to generate heat that is transferred through heat pipes to a thermoelectric (TE) converter. Waste heat is removed from the TEs by circulation of ambient air. Because such a power supply offers the promise of minimal operation and maintenance (O and M) costs as well as no fuel logistics, it is particularly attractive for remote, unattended applications such as the North Warning System

  1. Power Reactor Embrittlement Data Base

    International Nuclear Information System (INIS)

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

    1990-01-01

    Regulatory and research evaluations of embrittlement predication models and of pressure vessel integrity can be greatly expedited by the use of a well-designed, computerized data base. The Power Reactor Embrittlement Data Base (PR-EDB) is such a comprehensive collection of data for US commercial nuclear reactors. The Nuclear Regulatory Commission (NRC) has provided financial support, and the Electric Power Research Institute (EPRI) has provided technical assistance in the quality assurance (QA) of the data to establish an industry-wide data base that will be maintained and updated on a long-term basis. Successful applications of the data base to several of NRC's evaluations have received favorable response and support for its continuation. The future direction of the data base has been designed to include the test reactor and other types of data of interest to the regulators and the researchers. 1 ref

  2. Power reactor core safety research

    International Nuclear Information System (INIS)

    Rim, C.S.; Kim, W.C.; Shon, D.S.; Kim, J.

    1981-01-01

    As a part of nuclear safety research program, a project was launched to develop a model to predict fuel failure, to produce the data required for the localizaton of fuel design and fabrication technology, to establish safety limits for regulation of nuclear power plants and to develop reactor operation method to minimize fuel failure through the study of fuel failure mechanisms. During 1980, the first year of this project, various fuel failure mechanisms were analyzed, an experimental method for out-of-pile tests to study the stress corrosion cracking (SCC) behaviour of Zircaloy cladding underiodine environment was established, and characteristics of PWR and CANDU Zircaloy specimens were examined. Also developed during 1980 were the methods and correlations to evaluate fuel failures in the reactor core based on operating data from power reactors

  3. Experimental power reactor

    International Nuclear Information System (INIS)

    Anon.

    1977-01-01

    The following five topics are discussed using figures and diagrams: (1) energy storage and transfer program, (2) thermomechanical analysis, (3) a steam dual-cycle power conversion system for the EPR, (4) EPR tritium facility scoping studies, and (5) vacuum systems

  4. Consideration of the opinions and others in the public hearing on the alteration in reactor installation (addition of Unit 2) in the Tsuruga Power Station of the Japan Atomic Power Company

    International Nuclear Information System (INIS)

    1982-01-01

    A public hearing was held in Tsuruga City, Fukui Prefecture, on the alteration in reactor installation, i.e., the addition of Unit 2 in the Tsuruga Power Station, JAPC, on November 20, 1980, by the Nuclear Safety Commission. The opinions and others stated by the local people were taken into consideration in the governmental examinations on the installation, etc. The considerations of such opinions principally in the examinations by NSC are explained in the form of questions (i.e. opinion, etc.) and answers (i.e. consideration) as follows: site conditions (site, earthquakes, ground, meteorology, siting situation, etc.), the safety design of the reactor facilities (overall plant, aseismic design, the teaching by the TMI accident in U.S., ECCS, pre-stressed concrete containment vessel, radioactive waste release, etc.), radioactive waste management, radiation exposure relation, the technical capabilities of personnel (operation, etc.). (J.P.N.)

  5. Past and present role of fast breeder reactors in Japan

    International Nuclear Information System (INIS)

    Tomabechi, K.

    1978-01-01

    To meet increasing future energy demand in Japan utilization of fission energy should be promoted. In particular it is of vital importance to develop and utilize FBRs as soon as possible in order to save the natural uranium needed. If one considers the commercial introduction of FBRs in the mid-1990s in Japan, a delay of only one year will eventually result in an additional demand for natural uranium of more than 20,000 tonnes, because several LWRs will have to be installed instead. Ten years have passed since the development of FBRs in Japan was initiated as a national project with the highest priority and now the experimental fast reactor JOYO is successfully being commissioned and the prototype fast breeder reactor MONJU has reached the stage of proceeding to construction. Experience gained from past development work encourages further strong effort towards the commercialization of FBRs. This paper briefly reviews the important role to be played by FBRs, the development programme for FBRs, and the experience gained so far from the programme in Japan. (author)

  6. Research and development of super light water reactors and super fast reactors in Japan

    International Nuclear Information System (INIS)

    Oka, Y.; Morooka, S.; Yamakawa, M.; Ishiwatari, Y.; Ikejiri, S.; Katsumura, Y.; Muroya, Y.; Terai, T.; Sasaki, K.; Mori, H.; Hamamoto, Y.; Okumura, K.; Kugo, T.; Nakatsuka, T.; Ezato, K.; Akasaka, N.; Hotta, A.

    2011-01-01

    Super Light Water Reactors (Super LWR) and Super Fast Reactors (Super FR) are the supercritical- pressure light water cooled reactors (SCWR) that are developed by the research group of University of Tokyo since 1989 and now jointly under development with the researchers of Waseda University, University of Tokyo and other organizations in Japan. The principle of the reactor concept development, the results of the past Super LWR and Super FR R&D as well as the R&D program of the Super FR second phase project are described. (author)

  7. Reactor power peaking information display

    International Nuclear Information System (INIS)

    Book, T.L.; Kochendarfer, R.A.

    1986-01-01

    This patent describes a system for monitoring operating conditions within a nuclear reactor. The system consists of a method for measuring the operating parameters within the nuclear reactor, including the position of axial power shaping rods and regulating control rod. It also includes a method for determining from the operating parameters the operating limits before a power peaking condition exists within the nuclear reactor, and a method for displaying the operating limits which consists of a visual display permitting the continuous monitoring of the operating conditions within the nuclear reactor as a graph of the shaping rod position vs the regulating rod position having a permissible area and a restricted area. The permissible area is further divided into a recommended operating area for steady state operation and a cursor located on the graph to indicate the present operating condition of the nuclear reactor to allow an operator to view any need for corrective action based on the movement of the cursor out of the recommended operating area and to take any corrective transient action within the permissible area

  8. Reactor power region measuring device

    International Nuclear Information System (INIS)

    Kashiwa, Takao.

    1996-01-01

    The device of the present invention can rapidly detect abnormality of a local power region monitor (LPRM) even at a low power region caused such as upon start-up of a BWR type reactor. Namely, the present invention comprises (1) an LPRM detector for measuring neutron fluxes in the reactor, (2) a gamma thermo detector for calibrating the sensitivity of the LPRM detector, (3) a comparison circuit for comparing the detected values of the detectors (1) and (2), and (4) an alarm circuit for outputting an alarm when the comparative difference of the output of the circuit (3) exceeds a predetermined value. Signals of an alarm for a lower limit of the LPRM detector have been issued continuously upon start-up and shut down of the reactor since neutron fluxes in the reactor are reduced. However, the gamma thermo detector is always secured in the inside of the reactor different from a travelling-type incore probe monitor (TIP) disposed so far for the same purpose. Accordingly, the alarm generated upon usual start-up can be eliminated by comparing the detected values of the detector (2) and abnormality of the detector (1) can be rapidly detected by judging the abnormality of the comparative difference. (I.S.)

  9. An approach of raising the low power reactor trip block (P-7) in Maanshan Power Plant

    International Nuclear Information System (INIS)

    Wang, L.C.

    1984-01-01

    The technical specification for the Maanshan Nuclear Power Station (FSAR Table 16.2.2-3) requires that with an increasing reactor power level above the setpoint of low power reactor trip block (P-7), a turbine trip shall initiate a reactor trip. This anticipatory reactor trip on turbine trip prevents the pressurizer PORV from openning during turbine trip event. In order to reduce unnecessary reactor trip due to turbine trip on low reactor power level during Maanshan start-up stage, Taiwan Power Company performed a transient analysis for turbine trip event by using RETRAN code. The highest reactor power level at which a turbine trip will not open the pressurizer PORV is searched. The results demonstrated that this power level can be increased from the original value-10% of the rated thermal power-to about 48% of the rated thermal power

  10. Development of Probabilistic Safety Assessment with respect to the first demonstration nuclear power plant of high temperature gas cooled reactor in China

    International Nuclear Information System (INIS)

    Tong Jiejuan; Zhao Jun; Liu Tao; Xue Dazhi

    2012-01-01

    Due to the unique concept of HTR-PM (High Temperature Gas Cooled Reactor-Pebble Bed Module) design, Chinese nuclear authority has anticipated that HTR-PM will bring challenge to the present regulation. The pilot use of PSA (Probabilistic Safety Assessment) during HTR-PM design and safety review is deemed to be the necessary and efficient tool to tackle the problem, and is actively encouraged as indicated in the authority's specific policy statement on HTR-PM project. The paper summarizes the policy statement to set up the base of PSA development and application activities. The up-to-date status of HTR-PM PSA development and the risk-informed application activities are introduced in this paper as the follow-up response to the policy statement. For open discussion, the paper hereafter puts forward several technical issues which have been encountered during HTR-PM PSA development. Since HTR-PM PSA development experience has the general conclusion that many of the PSA elements can be and have been implemented successfully by the traditional PSA techniques, only the issues which extra innovative efforts may be needed are highlighted in this paper. They are safety goal and risk metrics, PSA modeling framework for the non-water reactors, passive system reliability evaluation, initiating events frequencies and component reliability data estimation techniques for the new reactors and so on. The paper presents the way in which the encountered technical issues were or will be solved, although the proposed way may not be the ultimate best solution. The paper intends to express the standpoint that although the PSA of new reactor has the inherent weakness due to the insufficient information and larger data uncertainty, the problem of component reliability data is much less severe than people have conceived. The unique design conception and functional features of the reactors can influence the results more significantly than the component reliability data. What we are benefited

  11. Time-optimal control of reactor power

    International Nuclear Information System (INIS)

    Bernard, J.A.

    1987-01-01

    Control laws that permit adjustments in reactor power to be made in minimum time and without overshoot have been formulated and demonstrated. These control laws which are derived from the standard and alternate dynamic period equations, are closed-form expressions of general applicability. These laws were deduced by noting that if a system is subject to one or more operating constraints, then the time-optimal response is to move the system along these constraints. Given that nuclear reactors are subject to limitations on the allowed reactor period, a time-optimal control law would step the period from infinity to the minimum allowed value, hold the period at that value for the duration of the transient, and then step the period back to infinity. The change in reactor would therefore be accomplished in minimum time. The resulting control laws are superior to other forms of time-optimal control because they are general-purpose, closed-form expressions that are both mathematically tractable and readily implanted. Moreover, these laws include provisions for the use of feedback. The results of simulation studies and actual experiments on the 5 MWt MIT Research Reactor in which these time-optimal control laws were used successfully to adjust the reactor power are presented

  12. Study of lower hybrid current drive for the demonstration reactor

    Energy Technology Data Exchange (ETDEWEB)

    Molavi-Choobini, Ali Asghar [Dept. of Physics, Faculty of Engineering, Islamic Azad University, Shahr-e-kord Branch, Shahr-e-kord (Iran, Islamic Republic of); Naghidokht, Ahmed [Dept. of Physics, Urmia University, Urmia (Iran, Islamic Republic of); Karami, Zahra [Dept. of Engineering, Islamic Azad University, Zanjan Branch, Zanjan (Iran, Islamic Republic of)

    2016-06-15

    Steady-state operation of a fusion power plant requires external current drive to minimize the power requirements, and a high fraction of bootstrap current is required. One of the external sources for current drive is lower hybrid current drive, which has been widely applied in many tokamaks. Here, using lower hybrid simulation code, we calculate electron distribution function, electron currents and phase velocity changes for two options of demonstration reactor at the launched lower hybrid wave frequency 5 GHz. Two plasma scenarios pertaining to two different demonstration reactor options, known as pulsed (Option 1) and steady-state (Option 2) models, have been analyzed. We perceive that electron currents have major peaks near the edge of plasma for both options but with higher efficiency for Option 1, although we have access to wider, more peripheral regions for Option 2. Regarding the electron distribution function, major perturbations are at positive velocities for both options for flux surface 16 and at negative velocities for both options for flux surface 64.

  13. Radioisotope production by reactors and cyclotrons in Japan

    International Nuclear Information System (INIS)

    Murakami, Yukio

    1978-01-01

    Present status of radioisotope production in Japan and the increasing demand from various fields are generally reviewed. Future problems associated with the shortage of economical supply are also discussed. The first half of this report is devoted to general review of the increasing demand for various radioisotopes from increasing number of users. The present status and future trends of the distribution of users of specific radioisotopes and their demands are shown. The remaining half of this report reviews the production with reactors and cyclotrons. The Japanese reactors producing radioisotopes are limited to low flux (10 13 ) research reactors at JAERI. Some problems associated with the improvement of availability and with the organizational structure are discussed. As for the production with cyclotrons, available facilities and the method of production are explained in detail. For clinical use, especially for the production of short lived radioisotopes, the advantage of a small special purpose cyclotron at each medical organization is emphasized. (Aoki, K.)

  14. Energy problems and nuclear power in Japan

    International Nuclear Information System (INIS)

    Shirasawa, T.

    1980-01-01

    International petroleum situation maintains the balance between demand and supply for the time being, but hereafter, it seems to be more serious and uncertain. Japanese economy tided over the first oil crisis with difficulty, and moreover, responded to the second oil crisis after the Iranian revolution somehow or other. But oil price has continued to rise, and the acceleration of inflation, the serious depression of businesses and electric power crisis are feared. In Japan where the dependence on imported petroleum is as high as 75%, it is necessary to establish the long term energy policy making energy saving and the development of substitute energy as its mainstay. In August, 1979, the report concerning the interim prospect of long term energy demand and supply was made. Largest efforts will be exerted to reduce the oil import. Then the total demand of energy in 1985 will be 582 million kl calculated in terms of petroleum. The law concerning energy saving was enacted in June, 1979. As the substitute energy, imported coal, LNG and nuclear power generation should be adopted. However, in order to put these energies in practical use, many problems to be solved remain. 21 nuclear power plants of 14.9 million kW capacity are in operation, and provide with 12% of total power generation installations. 30 million kW of nuclear power generation will be attained by 1985. (Kako, I.)

  15. Fusion Power Demonstrations I and II

    International Nuclear Information System (INIS)

    Doggett, J.N.

    1985-01-01

    In this report we present a summary of the first phase of the Fusion Power Demonstration (FPD) design study. During this first phase, we investigated two configurations, performed detailed studies of major components, and identified and examined critical issues. In addition to these design specific studies, we also assembled a mirror-systems computer code to help optimize future device designs. The two configurations that we have studied are based on the MARS magnet configuration and are labeled FPD-I and FPD-II. The FPD-I configuration employs the same magnet set used in the FY83 FPD study, whereas the FPD-II magnets are a new, much smaller set chosen to help reduce the capital cost of the system. As part of the FPD study, we also identified and explored issues critical to the construction of an Engineering Test Reactor (ETR). These issues involve subsystems or components, which because of their cost or state of technology can have a significant impact on our ability to meet FPD's mission requirements on the assumed schedule. General Dynamics and Grumman Aerospace studied two of these systems, the high-field choke coil and the halo pump/direct converter, in great detail and their findings are presented in this report

  16. Reactor cold neutron source facility, the first in Japan

    International Nuclear Information System (INIS)

    Utsuro, Masahiko; Maeda, Yutaka; Kawai, Takeshi; Tashiro, Tameyoshi; Sakakibara, Shoji; Katada, Minoru.

    1986-01-01

    In the Research Reactor Institute, Kyoto University, the first cold neutron source facility for the reactor in Japan was installed, and various tests are carried out outside the reactor. Nippon Sanso K.K. had manufactured it. After the prescribed tests outside the reactor, this facility will be installed soon in the reactor, and its outline is described on this occasion. Cold neutrons are those having very small energy by being cooled to about-250 deg C. Since the wavelength of the material waves of cold neutrons is long, and their energy is small, they are very advantageous as an experimental means for clarifying the structure of living body molecules and polymers, the atom configuration in alloys, and atomic and molecular movements by neutron scattering and neutron diffraction. The basic principle of the cold neutron source facility is to irradiate thermal neutrons on a cold moderator kept around 20 K, and to moderate and cool the neutrons by nuclear scattering to convert to cold neutrons. The preparatory research on cold neutrons and hydrogen liquefaction, the basic design to put the cold neutron source facility in the graphite moderator facility, the safety countermeasures, the manufacture and quality control, the operation outside the reactor and the performance are reported. The cold neutron source facility comprises a cold moderator tank and other main parts, a deuterium gas tank, a helium refrigerator and instrumentation. (Kako, I.)

  17. Progress on Fast Reactor Development in Japan

    International Nuclear Information System (INIS)

    Ohira, Hiroaki; Uto, Nariaki

    2012-01-01

    Situation of National Policy Making and FaCT Project: On July 19, 2011, JAEC decided to continue the FR cycle technology development program in the limited range of activities to contribute to international standardization (ex. safety criteria) and to maintain the technology base level until the determination of new nuclear energy policy. On Sept. 27, 2011, JAEC restarted the deliberation process for new Framework for Nuclear Energy Policy. - The process was suspended after the Fukushima NPS accidents. - Major issues: Safety, Cost, Nuclear Power and Fuel Cycle Options, Waste Management, International Perspectives, R&D planning, etc. - The process has been carried out to determine the new Framework with the relationship to new governmental energy/environmental policy making. On Dec. 21, 2011, Energy and Environment Council complied Basic guideline toward Presentation of Alternatives regarding the Strategy for Energy and the Environment. In the FaCT project, focus has been on further improvement on safety of next generation SFRs based on lessons learned from the Fukushima NPS accidents

  18. Public acceptance of nuclear power in Japan

    International Nuclear Information System (INIS)

    Iguchi, T.

    1995-01-01

    Japan has a fragile energy supply structure, with 84% of its energy depending on import; for example, 99.6% of the oil comes from overseas, which makes Japan's economic base rather vulnerable. In order to ensure constant energy supply, it is indispensable to diversify the energy sources and to create indigenous energy. In view of this, nuclear energy is considered to be the main alternative to crude oil because it has several advantages over other energy sources, such as stable supply of uranium and the fact that it is compatible with efforts to find solutions to global environmental problems. However, since the general public is not familiar with nuclear technology, it is difficult to get the understanding and co-operation of people. In view of this, public relations activities providing information on the need and safety of nuclear power generation have been performed. As a result, in recent years, about 70% of the people came to recognize the need for nuclear power generation. Although people's recognition of this need has increased substantially, it is still difficult for them to accept the construction of nuclear facilities, because of their anxiety regarding the safety of such plants and the lack of information by the government and electric utilities. This makes the acquisition of new sites for nuclear power plants difficult, so that the time required for developing such plants becomes longer. In order to eliminate people's anxieties, both the government and electric utilities should provide accurate information, at the proper time and using a method that makes it easy for the people to understand the problems involved. It is also important for the government and the electric utilities to listen carefully to the opinions and questions of people and to increase friendly communications with them. The government, electric utilities and constructors of nuclear facilities have to co-operate in order to improve the measures taken to gain public acceptance of

  19. Power reactor pressure vessel benchmarks

    International Nuclear Information System (INIS)

    Rahn, F.J.

    1978-01-01

    A review is given of the current status of experimental and calculational benchmarks for use in understanding the radiation embrittlement effects in the pressure vessels of operating light water power reactors. The requirements of such benchmarks for application to pressure vessel dosimetry are stated. Recent developments in active and passive neutron detectors sensitive in the ranges of importance to embrittlement studies are summarized and recommendations for improvements in the benchmark are made. (author)

  20. Configuration and layout of the tandem mirror Fusion Power Demonstrator

    International Nuclear Information System (INIS)

    Clarkson, I.R.; Neef, W.S.

    1983-01-01

    Studies have been performed during the past year to determine the configuration of a tandem mirror Fusion Power Demonstrator (FPD) machine capable of producing 1750 MW of fusion power. The FPD is seen as the next logical step after the Mirror Fusion Test Facility-B (MFTF-B) toward operation of a power reactor. The design of the FPD machine allows a phased construction: Phase I, a hydrogen or deuterium checkout machine; Phase 2, a DT breakeven machine; Phase 3, development of the Phase 2 machine to provide net power and act as a reactor demonstrator. These phases are essential to the development of remote handling equipment and the design of components that will ultimately be remotely handled. Phasing also permits more modes funding early in the program with some costs committed only after reaching major milestones

  1. Conceptual design study for the demonstration reactor of JSFR. (3) Safety design and evaluation

    International Nuclear Information System (INIS)

    Tani, Akihiro; Shimakawa, Yoshio; Kubo, Shigenobu; Fujimura, Ken; Yamano, Hidemasa

    2011-01-01

    This paper describes the result of conceptual safety design and evaluation for the demonstration plant of Japan sodium-cooled fast reactor (JSFR), which was preliminarily conducted for providing information necessary to decide the plant specification for further design study. The plant major specifications except for output power and safety design concept are almost the same as those of the commercial JSFR. A set of safety evaluation for typical design basis events (DBEs) is mainly focused here, which was conducted for the 750 MWe design. Safety analyses for DBEs evaluation were performed on the basis of conservative assumptions using a one-dimensional flow network code with point kinetics. For representative DBEs, transient over power type events and loss of flow type events were analyzed. The long-term loss-of-offsite power event was also calculated to evaluate the natural circulation decay heat removal system. All analytical results showed to meet tentative safety criteria, thus it was confirmed that the safety design concept of JSFR is feasible against DBEs. (author)

  2. The Optimization of power reactor control system

    International Nuclear Information System (INIS)

    Danupoyo, S.D.

    1997-01-01

    A power reactor is an important part in nuclear powered electrical plant systems. Success in controlling the power reactor will establish safety of the whole power plant systems. Until now, the power reactor has been controlled by a classical control system that was designed based on output feedback method. To meet the safety requirements that are now more restricted, the recently used power reactor control system should be modified. this paper describes a power reactor control system that is designed based on a state feedback method optimized with LQG (Linear-quadrature-gaussian) method and equipped with a state estimator. A pressurized-water type reactor has been used as the model. by using a point kinetics method with one group delayed neutrons. the result of simulation testing shows that the optimized control system can control the power reactor more effective and efficient than the classical control system

  3. The IAEA power reactor information system - PRIS

    International Nuclear Information System (INIS)

    Laue, H.J.; Qureshi, A.; Skjoeldebrand, R.; White, D.

    1983-01-01

    The IAEA Power Reactor Information System, PRIS, is based on a collection of basic design data and operating experience data which the IAEA started in 1970. PRIS is used for annual publications on 'Power Reactors in Member States', 'Operating Experience with Nuclear Power Stations in Member States', which gives annual operating information for individual plants, and a 'Performance Analysis Report' summarizing each year's and earlier experience. Since 1973 information has been collected in a systematic manner on significant plant outages (= more than 10 full power hours). There is now information on more than 10,000 outages in the system which permits some conclusions to be drawn both in regard to individual plants and to categories of plants on the significance of different outage reasons and different types of equipment failures. PRIS has not been intended to be a component reliability information system as an international data collection must stop short of the level of detail which would be needed for that purpose. The objectives of PRIS have been to provide a factual background for assumptions on parameters which are essential for economic evaluations and for systems operation planning (load factor and availability). The outage information does, however, lend itself to conclusions about generic problems in different categories of plants and it can be used by an individual operator to find other plants where information about particular problems can be obtained. It would also now be possible to use PRIS for setting availability goals based on experience and not only on theoretical design considerations. The paper demonstrates the conclusions which can be drawn from 662 reactor years of operation of light and heavy water pressurized reactors and 390 reactor years of boiling water reactors and, in particular, the role that the main heat removal system and its components have played in the equipment failure category

  4. Dose estimates in Japan following the Chernobyl reactor accident

    International Nuclear Information System (INIS)

    Togawa, Orihiko; Homma, Toshimitsu; Iijima, Toshinori; Midorikawa, Yuji.

    1988-02-01

    Estimates have been made of the maximum individual doses and the collective doses in Japan following the Chernobyl reactor accident. Based on the measured data of ground deposition and radionuclide concentrations in air, raw milk, milk on sale and leafy vegetables, the doses from some significant radionuclides were calculated for 5 typical exposure pathways; cloudshine, groundshine, inhalation, ingestion of milk and leafy vegetables. The maximum effective dose equivalents for hypothetical individuals were calculated to be 1.8 mrem for adults, 3.7 mrem for children and 6.0 mrem for infants. The collective effective dose equivalent in Japan was estimated to be 5.8 x 10 4 man · rem; 0.50 mrem of the average dose per capita. (author)

  5. Power reactors in Member States. 1979 edition

    International Nuclear Information System (INIS)

    1979-01-01

    This is the fifth issue of a periodic computer-based listing of nuclear power reactors, presenting the situation as of 1 May 1979. The basic design data for all reactors in operation, under construction, planned and shut down have been included. The following two tables are included to give a general picture of the overall situation: Table I: Reactor types and net electrical power. Table II: Reactor units and net electrical powered by country cummulated by year

  6. Computerized reactor power regulation with logarithmic controller

    International Nuclear Information System (INIS)

    Gossanyi, A.; Vegh, E.

    1982-11-01

    A computerized reactor control system has been operating at a 5 MW WWR-SM research reactor in the Central Research Institute for Physics, Budapest, for some years. This paper describes the power controller used in the SPC operating mode of the system, which operates in a 5-decade wide power range with +-0.5% accuracy. The structure of the controller easily limits the minimal reactor period and produces a reactor transient with constant period if the power demand changes. (author)

  7. A review of fast reactor program in Japan. April 2000 - March 2001

    International Nuclear Information System (INIS)

    Nagata, Takashi; Yamashita, Hidetoshi

    2001-01-01

    This report describes the development and activities on fast reactors in Japan thru April 2000 to March 2001. During this period, the most important result of the Japanese Fast Reactor Project was that the first phase 'Feasibility Study on Commercialized Fast Reactor Cycle Systems' was completed at the end of March 2001, and the second phase study has just started in order to narrow down the candidate concepts selected in the first phase for next stage. In the Experimental Fast Reactor 'Joyo', the 35 th rated power operation was completed by the end of May 2000. The 13 th periodical inspection and reconstruction works for the Joyo upgrading program (MK-III) were started on the beginning of June 2000. The modification of the cooling system is underway. In the Prototype Fast Breeder Reactor 'Monju', countermeasures against sodium leakage have already been drawn up based on 'Monju' comprehensive safety review. The Japan Atomic Energy Commission (JAEC) has issued a new 'Long-term Program for Research, Development and Utilization of Nuclear Energy' in November 2000. (author)

  8. Tritium management and anti-permeation strategies for three different breeding blanket options foreseen for the European Power Plant Physics and Technology Demonstration reactor study

    Energy Technology Data Exchange (ETDEWEB)

    Demange, D., E-mail: david.demange@kit.edu [Karlsruhe Institute of Technology, Institute for Technical Physics, Tritium Laboratory Karlsruhe, Herrmann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Boccaccini, L.V.; Franza, F. [Karlsruhe Institute of Technology, Institute for Neutron Physics and Reactor Technology, Herrmann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Santucci, A.; Tosti, S. [Associazione ENEA-Euratom sulla Fusione, C.R. ENEA Frascati, Via E. Fermi 45, 00044 Frascati (RM) (Italy); Wagner, R. [Karlsruhe Institute of Technology, Institute for Technical Physics, Tritium Laboratory Karlsruhe, Herrmann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany)

    2014-10-15

    In DT fusion reactors like DEMO, the commonly accepted tritium (T) losses through the steam generator (SG) shall not exceed about 2 mg/d that are more than 5 orders of magnitude lower than the T production rate of about 360 g/d in the breeding blanket (BB). A very effective mitigation strategy is required balancing the size and efficiency of the processes in the breeding and cooling loops, and the availability and efficiency of anti-permeation barriers. A numerical study is presented using the T permeation code FUS-TPC that computes all T flows and inventories considering the design and operation of the BB, the SG, and the T systems. Many scenarios are numerically analyzed for three breeding blankets concepts – helium cooled pebbles bed (HCPB), helium cooled lithium lead (HCLL), and water cooled lithium lead (WCLL) – varying the T processes throughput and efficiency, and the permeation regimes through the BB and SG to be either surface-limited or diffusion-limited with possible permeation reduction factor. For each BB concept, we discuss workable operation scenarios and suggest specific anti-permeation strategies.

  9. Advance Power Technology Demonstration on Starshine 3

    Science.gov (United States)

    Jenkins, Phillip; Scheiman, David; Wilt, David; Raffaelle, Ryne; Button, Robert; Smith, Mark; Kerslake, Thomas; Miller, Thomas

    2002-01-01

    The Starshine 3 satellite will carry several power technology demonstrations. Since Starshine 3 is primarily a passive experiment and does not need electrical power to successfully complete its mission, the requirement for a highly reliable power system is greatly reduced. This creates an excellent opportunity to test new power technologies. Several government and commercial interests have teamed up to provide Starshine 3 with a small power system using state-of-the-art components. Starshine 3 will also fly novel integrated microelectronic power supplies (IMPS) for evaluation.

  10. A prospect of fast reactor and related fuel cycle in Japan

    International Nuclear Information System (INIS)

    Nagata, Takashi

    2009-01-01

    JAEA has launched a new project 'Fast Reactor Cycle Technology Development'(FaCT) in cooperation with electric utilities. In this FaCT project, a combination of 'the Japanese sodium cooled loop type fast reactor with oxide fuel, the advanced aqueous reprocessing, and the simplified palletizing fuel fabrication systems' is adopted, where many innovative technologies with technical challenging issues are actively used in order to provide significant improvements in economic competitiveness, and enhancement of safety and reliability, sustainability, and nonproliferation. Fast reactor cycle technology will provide harmonic solutions for global issues of energy resources and environments, and is expected to contribute to sustainable development of the future society. Therefore, it was selected as one of key technologies of national importance in the third term (JPY2006-2010) 'Science and Technology Basic Plan' in March 2006 in Japan. The 'Nuclear Energy National Plan' in August 2006 states start up of a demonstration FR by around 2025 and deployment of a commercial FR before 2050, and start operating fuel cycle facilities when these reactors achieve consistency. Accordingly, we will decide about the adoption of innovative technologies by judging their applicability by 2010, and present the conceptual designs of commercial and demonstration FR cycle facilities by 2015 with the R and D plans to realize. In developing the FR cycle, 5 Party council, which consists of MEXt, MITI, electricity utilities, manufacturers, and JAEA, was established in July 2006 for moving forward on the commercialization smoothly. In this framework, users' requirements for the future R and D, a scenario of transition from light water reactor cycle to sodium cooled FR cycle, international collaboration, development schedule, demonstration steps, and so on are discussed. In this presentation, a prospect concerning the system design features of JSFR and a summary of the above R and D progresses for

  11. SP-100 space reactor power system readiness

    International Nuclear Information System (INIS)

    Josloff, A.T.; Matteo, D.N.; Bailey, H.S.

    1992-01-01

    This paper discusses the SP-100 Space Reactor Power System which is being developed by GE, under contract to the U.S. Department of Energy, to provide electrical power in the range of 10's to 100's of kW. The system represents an enabling technology for a wide variety of earth orbital and interplanetary science missions, nuclear electric propulsion (NEP) stages, and lunar/Mars surface power for the Space Exploration Initiative (SEI). The technology and design is now at a state of readiness to support the definition of early flight demonstration missions. Of particular importance is that SP-100 meets the demanding U.S. safety performance, reliability and life requirements. The system is scalable and flexible and can be configured to provide 10's to 100's of kWe without repeating development work and can meet DoD goals for an early, low-power demonstration flight in the 1996-1997 time frame

  12. Current approaches to nuclear power plant life management in Japan

    International Nuclear Information System (INIS)

    Noda, T.; Tajima, K.; Ishikawa, M.; Koyama, M.

    2002-01-01

    Full text: Some of Japan's commercial light water reactors (LWRs) have been operating for more than 30 years. The more progress in ageing, the more increasing concerns of the public will grow about such nuclear power plants. In order to develop basic policies regarding countermeasures against ageing on nuclear power plants, in 1996, the Ministry of International Trade and Industry (MITI) summarized a report entitled 'Basic Policy on Aged Nuclear Power Plants'. The MITI also indicated that following 30 years' commercial operation of these plants, the electric utility companies should conduct technical evaluations for the ageing of all the components in the plants and to prepare detailed maintenance plans for the future. The Nuclear Safety Commission (NSC) accepted the MITI's report as appropriate in November 1998. The Commission also recommended the addition of effective countermeasures against ageing to the Periodical Safety Review and the evaluation of activities in response to ageing in order to implement such activities regularly and systematically in the future. The MITI reviewed the ageing countermeasures conducted by the electric utility companies and issued the second report entitled 'Evaluation of Countermeasures for ageing Conducted by Electric Utility Companies and Future Plans to cope with ageing'. The evaluation was made for Tsuruga Power Station Unit 1, Mihama Power Station Unit 1, and Fukushima Daiichi Nuclear Power Station Unit 1. At the same time, the MITI determined to incorporate the technical evaluations of ageing and the preparation of long-term maintenance plans into the periodical safety review in the future. The Kansai Electric Power Co., Inc., and Tokyo Electric Power Co. conducted the technical evaluations in their periodical safety reviews concerning the ageing phenomena of all their safety-related components/structures of Mihama Power Station Unit 2 and Fukushima Daiichi Nuclear Power Station Unit 2. Also, concerning ageing, they

  13. On the definition of a DEMO (demonstration) reactor

    International Nuclear Information System (INIS)

    Cole, H.C.; Challender, R.S.

    1987-01-01

    The paper is an appendix to a study of the reactor relevance of the NET design concept. The latter study examines whether the technologies and design principles proposed for NET can be directly extrapolated to a demonstration (DEMO) reactor. The authors have suggested a definition of a DEMO, and listed what they considered to be the most important implications of this definition. A table of parameters is included comparing published DEMO's with typical commercial reactor and 'pre-DEMO' studies. (U.K.)

  14. Effort to grapple with improvement of security and reliability of nuclear power plant. Actions of the Japan Atomic Power Company

    International Nuclear Information System (INIS)

    Ishiguma, Kazuo

    2012-01-01

    Following the Great Tohoku Earthquake in 2011, Tokai No.2 reactor was shut down automatically. Three of emergency diesel generators worked automatically at loss-of-offsite-power and began to work the cooling system of reactor. The reactor could be kept stable and safe in cold state by management of power from the gas turbine electric generator and power source car. Actions of Japan Atomic Power Company (JAPC) for cold shutdown and Tsunami were stated. Inspection results after the earthquake and testimony of staff was described. Countermeasure of improvement of safety of nuclear power station is explained by ensuring of power source and water supply, crisis management system, countermeasure of accident, ensuring, and training of workers, and action for better understanding of reliance. (S.Y.)

  15. Operational power reactor health physics

    International Nuclear Information System (INIS)

    Watson, B.A.

    1987-01-01

    Operational Health Physics can be comprised of a multitude of organizations, both corporate and at the plant sites. The following discussion centers around Baltimore Gas and Electric's (BG and E) Calvert Cliffs Nuclear Power Plant, located in Lusby, Maryland. Calvert Cliffs is a twin Combustion Engineering 825 MWe pressurized water reactor site with Unit I having a General electric turbine-generator and Unit II having a Westinghouse turbine-generator. Having just completed each Unit's ten-year Inservice Inspection and Refueling Outge, a total of 20 reactor years operating health physics experience have been accumulated at Calvert Cliffs. Because BG and E has only one nuclear site most health physics functions are performed at the plant site. This is also true for the other BG and E nuclear related organizations, such as Engineering and Quality Assurance. Utilities with multiple plant sites have corporate health physics entity usually providing oversight to the various plant programs

  16. The characteristics of the prestressed concrete reactor vessel of the HHT demonstration plant

    International Nuclear Information System (INIS)

    Schoening, J.; Schwiers, H.G.

    1979-01-01

    The paper concentrates on the design studies of the HTGR prestressed concrete reactor vessel (PCRV) for the HHT Demonstration Plant. The multi-cavity reactor pressure vessel accommodates all components carrying primary gas, including heat exchangers and gas turbine. For reasons of economics and availability of the reactor plant, generic requirements are made for the PCRV. A short description of the power plant is also presented

  17. Nuclear energy in Japan

    International Nuclear Information System (INIS)

    Guillemard, B.

    1978-01-01

    After having described the nuclear partners in Japan, the author analyzes the main aspects of Japan's nuclear energy: nuclear power plants construction program; developping of light water reactors; fuel cycle politics [fr

  18. Power Reactor Embrittlement Data Base

    International Nuclear Information System (INIS)

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

    1990-01-01

    Regulatory and research evaluations of embrittlement prediction models and of vessel integrity under load can be greatly expedited by the use of a well designed, computerized embrittlement data base. The Power Reactor Embrittlement Data Base (PR-EDB) is a comprehensive collection of data from surveillance reports and other published reports of commercial nuclear reactors. The uses of the data base require that as many different data as available are collected from as many sources as possible with complete references and that subsets of relevant data can be easily retrieved and processed. The objectives of this NRC-sponsored program are the following: (1) to compile and to verify the quality of the PR-EDB; (2) to provide user-friendly software to access and process the data; (3) to explore or confirm embrittlement prediction models; and (4) to interact with standards organizations to provide the technical bases for voluntary consensus standards that can be used in regulatory guides, standard review plans, and codes. To achieve these goals, the data base architecture was designed after much discussion and planning with prospective users, namely, material scientists and members of the research staff. The current compilation of the PR-EDB (Version 1) contains results from surveillance capsule reports of 78 reactors with 381 data points for 110 different irradiated base materials and 161 data points for 79 different welds. Results from heat-affected zone materials are also listed. The time and effort required to process and evaluate different types of data in the PR-EDB have been drastically reduced from previous data bases. The Electric Power Research Institute (EPRI), reactor vendors, and utilities are in the process of providing back-up quality assurance checks of PR-EDB and will be supplementing the data base with additional data and documentation

  19. Reactor power reduction system and method

    International Nuclear Information System (INIS)

    Bruno, S.J.; Dunn, S.A.; Raber, M.

    1978-01-01

    A method of operating a nuclear power reactor is disclosed which enables an accelerated power reduction of the reactor without completely shutting the reactor down. The method includes monitoring the incidents which, upon their occurrence, would require an accelerated power reduction in order to maintain the reactor in a safe operation mode; calculating the power reduction required on the occurrence of such an incident; determining a control rod insertion sequence for the normal operation of the reactor, said sequence being chosen to optimize reactor power capability; selecting the number of control rods necessary to respond to the accelerated power reduction demand, said selection being made according to a priority determined by said control rod insertion sequence; and inserting said selected control rods into the reactor core. 11 claims, 13 figures

  20. Present state and future prospect of development of high temperature gas-cooled reactors in Japan

    International Nuclear Information System (INIS)

    Sanokawa, Konomo

    1994-01-01

    High temperature gas-cooled reactors can supply the heat of about 1000degC, and the high efficiency and the high rate of heat utilization can be attained. Also they have the features of excellent inherent safety, the easiness of operation, the high burnup of fuel and so on. The heat utilization of atomic energy in addition to electric power generation is very important in view of the protection of global environment and the diversification of energy supply. Japan Atomic Energy Research Institute has advanced the construction of the high temperature engineering test and research reactor (HTTR) of 30 MW thermal output, aiming at attaining the criticality in 1998. The progress of the development of a high temperature gas-cooled reactor is described. For 18 years, the design study of the reactor was advanced together with the research and development of the reactor physics, fuel and materials, high temperature machinery and equipment and others, and the decision of the design standard and the development of computation codes. The main specification and the construction schedule are shown. The reactor building was almost completed, and the reactor containment vessel was installed. The plan of the research and development by using the HTTR is investigated. (K.I.)

  1. Study on reactor vessel replacement (RVR) for 1100 MW class BWR plants in Japan

    International Nuclear Information System (INIS)

    Mizutani, J.; Kawamura, S.; Aoki, M.; Mori, T.

    2001-01-01

    Plant Life Management (PLM) is being studied in Japan, and reactor vessel replacement (RVR) is being considered as one option. Since reactor internals, except for reusable parts, and the reactor pressure vessel (RPV) are replaced, the RVR provides an effective technology for extending the service life of nuclear power plants substantially. At ICONE 7, we reported on the technical viability of the RVR for BWR4-type 800 MWe class plants. This time, we rationalized the RVR method through a study for BWR5-type 1100 MWe class plants to reduce the RVR duration and evaluated the technical viability and the economic efficiency of the method. In addition, we discuss how to dispose of the RPV to complete a scenario of the process from the RVR to its final disposal. (author)

  2. Physical protection of power reactors

    International Nuclear Information System (INIS)

    Darby, J.L.

    1979-01-01

    Sandia Laboratories has applied a systematic approach to designing physical protection systems for nuclear facilities to commercial light-water reactor power plants. A number of candidate physical protection systems were developed and evaluated. Focus is placed on the design of access control subsystems at each of three plant layers: the protected area perimeter, building surfaces, and vital areas. Access control refers to barriers, detectors, and entry control devices and procedures used to keep unauthorized personnel and contraband out of the plant, and to control authorized entry into vital areas within the plant

  3. 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

  4. Safety design study of fast breeder reactors in Japan

    International Nuclear Information System (INIS)

    Miura, M.; Inagaki, T.

    1992-01-01

    This paper reports on two fast breeder reactor (FBR) concepts, the tank type and the loop type, that have been studied as possible reactor designs to be used for a demonstration FBR (DFBR). The basic principle fo the DFBR design is to ensure plant safety through a defense-in-depth methodology. Improvements in the seismic and thermal stress designs have been attempted for both reactor concepts. The system design study strives to maximize the reliability of the safety-related systems and to rationalize commercialization of the plant

  5. Feasible reactor power cutback logic development for an integral reactor

    International Nuclear Information System (INIS)

    Han, Soon-Kyoo; Lee, Chung-Chan; Choi, Suhn; Kang, Han-Ok

    2013-01-01

    Major features of integral reactors that have been developed around the world recently are simplified operating systems and passive safety systems. Even though highly simplified control system and very reliable components are utilized in the integral reactor, the possibility of major component malfunction cannot be ruled out. So, feasible reactor power cutback logic is required to cope with the malfunction of components without inducing reactor trip. Simplified reactor power cutback logic has been developed on the basis of the real component data and operational parameters of plant in this study. Due to the relatively high rod worth of the integral reactor the control rod assembly drop method which had been adapted for large nuclear power plants was not desirable for reactor power cutback of the integral reactor. Instead another method, the control rod assembly control logic of reactor regulating system controls the control rod assembly movements, was chosen as an alternative. Sensitivity analyses and feasibility evaluations were performed for the selected method by varying the control rod assembly driving speed. In the results, sensitivity study showed that the performance goal of reactor power cutback system could be achieved with the limited range of control rod assembly driving speed. (orig.)

  6. 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)

  7. Activities on Fast Reactor Knowledge Preservation in Japan

    International Nuclear Information System (INIS)

    Uto, Nariaki; Ohira, Hiroaki

    2013-01-01

    Japanese FRs are stepping up from Joyo, Monju to JSFR with efforts to overcome the Fukushima no. 1 Nuclear Power Station Accidents, including focus on ensuring and enhancing safety of these FRs. Japan (JAEA) has made continuous and contributive efforts to supply FR information to IAEA/INIS. JAEA has also contributed to a global dissemination of its R&D activities through JOPSS English version, which will be beneficial to worldwide FR development

  8. Nuclear power plant with several reactors

    Energy Technology Data Exchange (ETDEWEB)

    Grishanin, E I; Ilyunin, V G; Kuznetsov, I A; Murogov, V M; Shmelev, A N

    1972-05-10

    A design of a nuclear power plant suggested involves several reactors consequently transmitting heat to a gaseous coolant in the joint thermodynamical circuit. In order to increase the power and the rate of fuel reproduction the low temperature section of the thermodynamical circuit involves a fast nuclear reactor, whereas a thermal nuclear reactor is employed in the high temperature section of the circuit for intermediate heating and for over-heating of the working body. Between the fast nuclear and the thermal nuclear reactors there is a turbine providing for the necessary ratio between pressures in the reactors. Each reactor may employ its own coolant.

  9. Nuclear reactor instrumentation power monitor

    International Nuclear Information System (INIS)

    Suzuki, Shigeru.

    1989-01-01

    The present invention concerns a nuclear reactor instrumentation power monitor that can be used in, for example, BWR type nuclear power plants. Signals from multi-channel detectors disposed on field units are converted respectively by LPRM signal circuits. Then, the converted signals are further converted by a multiplexer into digital signals and transmitted as serial data to a central monitor unit. The thus transmitted serial data are converted into parallel data in the signal processing section of the central monitor unit. Then, LPRM signals are taken out from each of channel detectors to conduct mathematical processing such as trip judgment or averaging. Accordingly, the field unit and the central monitor unit can be connected by way of only one data transmission cable thereby enabling to reduce the number of cables. Further, since the data are transmitted on digital form, it less undergoes effect of noises. (I.S.)

  10. IAEA Expert Team Completes Mission to Review Japan's Nuclear Power Plant Safety Assessment Process, 31 January 2012, Tokyo, Japan

    International Nuclear Information System (INIS)

    2012-01-01

    Full text: A team of international nuclear safety experts today completed a review of Japan's two-stage process for assessing nuclear safety at the nation's nuclear power plants. The team began its work on 23 January and delivered a Preliminary Summary Report to Japanese officials today and plans to finish the final report by the end of February. National safety assessments and their peer review by the IAEA are a key component of the IAEA's Action Plan on Nuclear Safety, which was approved by the Agency's 152 Member States following last year's nuclear accident at Fukushima Daiichi Nuclear Power Station. At the request of the Government of Japan, the International Atomic Energy Agency (IAEA) organized a 10-person team to review the Japanese Nuclear and Industrial Safety Agency's (NISA) approach to the Comprehensive Assessments for the Safety of Existing Power Reactor Facilities and how NISA examines the results submitted by nuclear operators. The IAEA safety review mission consisted of five IAEA and three international nuclear safety experts. To help its review, the team held meetings in Tokyo with officials from NISA, the Japanese Nuclear Energy Safety (JNES) Organization, and the Kansai Electric Power Company (KEPCO), and the team visited the Ohi Nuclear Power Station to see an example of how Japan's Comprehensive Safety Assessment is being implemented by nuclear operators. 'We concluded that NISA's instructions to power plants and its review process for the Comprehensive Safety Assessments are generally consistent with IAEA Safety Standards', said team leader James Lyons, director of the IAEA's Nuclear Installation Safety Division. In its Preliminary Summary Report delivered today, the team highlighted a number of good practices and identified some improvements that would enhance the overall effectiveness of the Comprehensive Safety Assessment process. Good practices identified by the mission team include: Based on NISA instructions and commitments of the

  11. Wave Power Demonstration Project at Reedsport, Oregon

    Energy Technology Data Exchange (ETDEWEB)

    Mekhiche, Mike [Principal Investigator; Downie, Bruce [Project Manager

    2013-10-21

    Ocean wave power can be a significant source of large‐scale, renewable energy for the US electrical grid. The Electrical Power Research Institute (EPRI) conservatively estimated that 20% of all US electricity could be generated by wave energy. Ocean Power Technologies, Inc. (OPT), with funding from private sources and the US Navy, developed the PowerBuoy to generate renewable energy from the readily available power in ocean waves. OPT's PowerBuoy converts the energy in ocean waves to electricity using the rise and fall of waves to move the buoy up and down (mechanical stroking) which drives an electric generator. This electricity is then conditioned and transmitted ashore as high‐voltage power via underwater cable. OPT's wave power generation system includes sophisticated techniques to automatically tune the system for efficient conversion of random wave energy into low cost green electricity, for disconnecting the system in large waves for hardware safety and protection, and for automatically restoring operation when wave conditions normalize. As the first utility scale wave power project in the US, the Wave Power Demonstration Project at Reedsport, OR, will consist of 10 PowerBuoys located 2.5 miles off the coast. This U.S. Department of Energy Grant funding along with funding from PNGC Power, an Oregon‐based electric power cooperative, was utilized for the design completion, fabrication, assembly and factory testing of the first PowerBuoy for the Reedsport project. At this time, the design and fabrication of this first PowerBuoy and factory testing of the power take‐off subsystem are complete; additionally the power take‐off subsystem has been successfully integrated into the spar.

  12. Reactor power control method upon accidents of electrical power system

    International Nuclear Information System (INIS)

    Hirose, Masao.

    1983-01-01

    Purpose: To enable to continue the operation of a BWR type reactor by avoiding the scram while suppressing the reactor power, just after the external disturbance such as earth-trouble in power-transmission network. Method: Steep power drop of an electrical generator is to be detected not only by a current-type power-load-unbalance relay but also with a power-type power-load-unbalance-relay. If steep power-drop was detected by the latter relay, a previously selected control rod is rapidly inserted into the reactor. In this way, in the case where there is a possibility of the reactor scram, the scram can be avoided by suppressing the reactor power, thus the reactor operation can be continued. (Kamimura, M.)

  13. Power market restructuring in Asia : Russia, China, India, and Japan

    International Nuclear Information System (INIS)

    Hammons, T.J.; Zhong, J.; Mukhopadhyay, S.; Kurihara, I.

    2008-01-01

    Many countries are now in the process of deregulating their power industries in order to promote growth and competitiveness. This paper discussed power market restructuring activities in Russia, China, India and Japan. Economic convergence points in Russian and Asian power markets were reviewed. The state of Russia and China's power industry after the implementation of recent power restructuring initiatives was discussed. Technical characteristics of the industries were evaluated, and market development plans were outlined. Regional electricity markets in Asia were discussed, as well as issues related to domestic and foreign investment. Institutional reforms were reviewed, and individual outlines of revisions for the power industries of the 4 countries were presented. The study demonstrated that structuring processes vary from country to country. Differences in restructuring patterns were attributed to economic differences; country-specific features established within the electric power industry; and attitudes towards deregulation. It was concluded that the reforms adopted by the countries will lead to the expansion of national electric power systems. 23 refs., 4 tabs., 6 figs

  14. Research and development for high temperature gas cooled reactor in Japan

    International Nuclear Information System (INIS)

    Taketani, K.

    1978-01-01

    The paper describes the current status of High Temperature Gas Cooled Reactor research and development work in Japan, with emphasis on the Experimental Very High Temperature Reactor (Exp. VHTR) to be built by Japan Atomic Energy Research Institute (JAERI) before the end of 1985. The necessity of construction of Exp. VHTR was explained from the points of Japanese energy problems and resources

  15. BN-1200 Reactor Power Unit Design Development

    International Nuclear Information System (INIS)

    Vasilyev, B.A.; Shepelev, S.F.; Ashirmetov, M.R.; Poplavsky, V.M.

    2013-01-01

    Main goals of BN-1200 design: • Develop a reliable new generation reactor plant for the commercial power unit with fast reactor to implement the first-priority objectives in changing over to closed nuclear fuel cycle; • Improve technical and economic indices of BN reactor power unit to the level of those of Russian VVER of equal power; • Enhance the safety up to the level of the requirements for the 4th generation RP

  16. High power ring methods and accelerator driven subcritical reactor application

    Energy Technology Data Exchange (ETDEWEB)

    Tahar, Malek Haj [Univ. of Grenoble (France)

    2016-08-07

    High power proton accelerators allow providing, by spallation reaction, the neutron fluxes necessary in the synthesis of fissile material, starting from Uranium 238 or Thorium 232. This is the basis of the concept of sub-critical operation of a reactor, for energy production or nuclear waste transmutation, with the objective of achieving cleaner, safer and more efficient process than today’s technologies allow. Designing, building and operating a proton accelerator in the 500-1000 MeV energy range, CW regime, MW power class still remains a challenge nowadays. There is a limited number of installations at present achieving beam characteristics in that class, e.g., PSI in Villigen, 590 MeV CW beam from a cyclotron, SNS in Oakland, 1 GeV pulsed beam from a linear accelerator, in addition to projects as the ESS in Europe, a 5 MW beam from a linear accelerator. Furthermore, coupling an accelerator to a sub-critical nuclear reactor is a challenging proposition: some of the key issues/requirements are the design of a spallation target to withstand high power densities as well as ensure the safety of the installation. These two domains are the grounds of the PhD work: the focus is on the high power ring methods in the frame of the KURRI FFAG collaboration in Japan: upgrade of the installation towards high intensity is crucial to demonstrate the high beam power capability of FFAG. Thus, modeling of the beam dynamics and benchmarking of different codes was undertaken to validate the simulation results. Experimental results revealed some major losses that need to be understood and eventually overcome. By developing analytical models that account for the field defects, one identified major sources of imperfection in the design of scaling FFAG that explain the important tune variations resulting in the crossing of several betatron resonances. A new formula is derived to compute the tunes and properties established that characterize the effect of the field imperfections on the

  17. Some problems on domestic technology development from a point of fabricator of nuclear power plant. [Japan

    Energy Technology Data Exchange (ETDEWEB)

    Watamori, T [Hitachi Ltd., Tokyo (Japan)

    1976-06-01

    During past 20 years, the nuclear power industry in Japan has introduced foreign technology, digested it in a short period, and continued to research and develop domestic technology. Now, 95% of the machinery and equipments for nuclear power generation with light water reactors can be produced domestically, and some technologies are going to be exported. However, the nuclear power industry is still in a severe environment. The progress of the development of nuclear power plants passed the periods of organizational preparation, the construction of research reactors, the import of foreign technologies and reactors for practical use, and the construction of domestically produced reactors for practical use. The supplying capacity of the nuclear power industry in Japan reached 6 units of 1,000 MW yearly, but in order to meet the long term plan of nuclear power generation, this capacity must be further enhanced. The problems in the promotion of domestic production are the establishment of independent technologies, the promotion of standardization, the strengthening of business basis, the upbringing of relating enterprises, and the acceleration of national projects. Since the energy crisis, the trend of filling up energy demand with nuclear power generation became conspicuous, but for the expansion of export, the problems of safety guarantee, nuclear fuel cycle, and financial measures must be solved with government aid.

  18. Progress Report on Sodium Cooled Fast Breeder Reactor Development in Japan, April 1975

    International Nuclear Information System (INIS)

    Tomabechi, K.

    1975-01-01

    The progress of the sodium cooled fast Breeder Reactor development in Japan in the past 12 months can be summarized as follows. Installation of all the components of the Experimental Fast Reactor, ''JOYO'', was completed in the end of the last year and various commissioning tests of the reactor began in January 1975. It is planned to charge sodium into the reactor in coming fall and the first criticality experiment is currently planned in the summer 1976. Most of the research and development works for ''JOYO'' are nearing completion. These include an endurance test of 3 prototype primary sodium pump for 12,000 hours. 86 core fuel subassemblies and 220 blanket subassemblies, a sufficient number for composing the initial core, have already been fabricated. Concerning the Prototype Fast Breeder Reactor, ''MONJU'', design activity as well as relevant research and development works are continued. A siting problem exists and it is hoped to be resolved soon. Of the research and development works, a significant achievement in the past 12 months can be a successful operation at full power of the 50 MW Steam Generator Test Facility. This facility was put into operation at full power in June 1974. No leak of water into sodium has been experienced with operation of the steam generator tested. The steam generator is being dismantled for a detailed inspection originally planned

  19. Early hydrogen water chemistry in the boiling water reactor: industry-first demonstration

    International Nuclear Information System (INIS)

    Garcia, Susan E.; Odell, Andrew D.; Giannelli, Joseph F.

    2012-09-01

    Hydrogen injection into the BWR feedwater during power operation has resulted in significant IGSCC reductions. Further, noble metal application (NMCA) during shutdown or On-line NobleChem TM (OLNC) during power operation has greatly reduced the required hydrogen injection rate by catalyzing the hydrogen-oxygen reaction on the metal surfaces, reducing the electrochemical corrosion potential (ECP) at operating temperature to well below the mitigation ECP of -230 mV (SHE) at reactor water hydrogen to oxidant (O 2 + H 2 O 2 ) molar ratios of ≥2. Since IGSCC rates increase markedly at reduced temperature, and the potential for crack initiation exists, additional crack mitigation was desired. To close this gap in mitigation, the EPRI BWR Startup ECP Reduction research and development program commenced in 2008 to undertake laboratory and feasibility studies for adding a reductant to the reactor water system during start-ups. Under this program, ECP reductions of noble metal treated stainless steel sufficient to mitigate IGSCC at startup temperatures were achieved in the laboratory in the absence of radiation at hydrogen, hydrazine and carbohydrazide to oxygen molar ratios of ≥ 2, ≥1.5 and ≥0.7, respectively. Based on the familiarity of operating BWRs with using hydrogen, a demonstration of hydrogen injection during the startup of an actual BWR using noble metals was planned. This process, named EHWC (Early Hydrogen Water Chemistry), differs from the HDS (Hydrogen During Startup) approach that has been successful in Japan in that HDS injects sufficient hydrogen for bulk oxidant reduction whereas EHWC injects a smaller amount of hydrogen, sufficient to achieve a hydrogen:oxidant molar ratio of at least two at noble metal treated surfaces. The industry-first EHWC demonstration was performed at Exelon's Peach Bottom 3 nuclear power plant in October 2011. Prior to EHWC, Peach Bottom 3 had one NMCA (October 1999) and five annual OLNC applications (starting in 2007

  20. Reactor technology: power conversion systems and reactor operation and maintenance

    International Nuclear Information System (INIS)

    Powell, J.R.

    1977-01-01

    The use of advanced fuels permits the use of coolants (organic, high pressure helium) that result in power conversion systems with good thermal efficiency and relatively low cost. Water coolant would significantly reduce thermal efficiency, while lithium and salt coolants, which have been proposed for DT reactors, will have comparable power conversion efficiencies, but will probably be significantly more expensive. Helium cooled blankets with direct gas turbine power conversion cycles can also be used with DT reactors, but activation problems will be more severe, and the portion of blanket power in the metallic structure will probably not be available for the direct cycle, because of temperature limitations. A very important potential advantage of advanced fuel reactors over DT fusion reactors is the possibility of easier blanket maintenance and reduced down time for replacement. If unexpected leaks occur, in most cases the leaking circuit can be shut off and a redundant cooling curcuit will take over the thermal load. With the D-He 3 reactor, it appears practical to do this while the reactor is operating, as long as the leak is small enough not to shut down the reactor. Redundancy for Cat-D reactors has not been explored in detail, but appears feasible in principle. The idea of mobile units operating in the reactor chamber for service and maintenance of radioactive elements is explored

  1. Summary of Great East Japan Earthquake response at Onagawa Nuclear Power Station and further safety improvement measures

    International Nuclear Information System (INIS)

    Sato, Toru

    2013-01-01

    A large earthquake occurred on March 11, 2011 and tsunami was generated following it. The East Japan suffered serious damage by the earthquake and tsunami. This is called the Great East Japan Earthquake. Onagawa Nuclear Power Station (NPS) is located closest to the epicenter of Great East Japan Earthquake. We experienced intense shake by the earthquake and some flooding from the tsunami, however, we have succeeded safely cold shutdown of the reactors. In this paper, we introduce summary of Great East Japan Earthquake response a Onagawa NPS and safety improvement measures which are based on both experience of Onagawa NPS and lesson from Fukushima Daiichi NPS accident. (author)

  2. Public acceptance of nuclear power development in Japan

    International Nuclear Information System (INIS)

    Ohori, H.

    1977-01-01

    Although the Japanese set out to achieve the peaceful uses of atomic energy in 1956, the question of public acceptance took on serious proportions only as the development of nuclear power moved toward commercial application. A string of reactor troubles over the past few years complicated the question apparently to the point where it could scarcely be worse. It is not possible to deal with opposition movements in Japan without taking into account the background of the special national sentiment born of the people's experience of the atomic bombings, but it is also true that the people's deep-going fears of atomic energy have been increased by sensational newspaper reports, as well as internetional attacks by the opponents of nuclear development. Added to this, the ''Mutsu'' incident and other troubles have given the people distrust of the whole nuclear administration and those responsible for nuclear regulation. but, at the same time, the oil crisis of 1973 brought about an awakening of the people to the need for the development of nuclear power to solve Japan's energy problems, for Japan is seriously lacking in natural resources. An influential newspaper took samplings of public opinion in 1975 which revealed that, while 48 percent of the people expressed fears of atomic energy, 70 percent, including those who had some misgivings but still took the need for granted, said that Japan has no choice but to depend on nuclear power. The Government and industry have made long-range projections on nuclear power generation, forecasting that it will expand to 25 percent of all power generating plants by 1985, and to 35 percent by 1990. The gravest problem to be solved if this projected scale of nuclear development is to be achieved is the shortage of adequate plant sites. This can not be solved unless every effort is made to dispel the general feeling of mistrust mentioned, and to make sure that the development of nuclear power is socially accepted. It is hoped that the

  3. Power reactor embrittlement data base

    International Nuclear Information System (INIS)

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

    1989-01-01

    Regulatory and research evaluations of embrittlement prediction models and of vessel integrity under load can be greatly expedited by the use of a well-designed, computerized embrittlement data base. The Power Reactor Embrittlement Data Base (PR-EDB) is a comprehensive collection of data from surveillance reports and other published reports of commercial nuclear reactors. The uses of the data base require that as many different data as available are collected from as many sources as possible with complete references and that subsets of relevant data can be easily retrieved and processed. The objectives of this NRC-sponsored program are the following: to compile and to verify the quality of the PR-EDB; to provide user-friendly software to access and process the data; to explore or confirm embrittlement prediction models; and to interact with standards organizations to provide the technical bases for voluntary consensus standards that can be used in regulatory guides, standard review plans, and codes. 9 figs

  4. Preparations for the Integral Fast Reactor fuel cycle demonstration

    International Nuclear Information System (INIS)

    Lineberry, M.J.; Phipps, R.D.

    1989-01-01

    Modifications to the Hot Fuel Examination Facility-South (HFEF/S) have been in progress since mid-1988 to ready the facility for demonstration of the unique Integral Fast Reactor (IFR) pyroprocess fuel cycle. This paper updates the last report on this subject to the American Nuclear Society and describes the progress made in the modifications to the facility and in fabrication of the new process equipment. The IFR is a breeder reactor, which is central to the capability of any reactor concept to contribute to mitigation of environmental impacts of fossil fuel combustion. As a fast breeder, fuel of course must be recycled in order to have any chance of an economical fuel cycle. The pyroprocess fuel cycle, relying on a metal alloy reactor fuel rather than oxide, has the potential to be economical even at small-scale deployment. Establishing this quantitatively is one important goal of the IFR fuel cycle demonstration

  5. Status of reduced enrichment program for research reactors in Japan

    International Nuclear Information System (INIS)

    Kaieda, Keisuke; Baba, Osamu; Nagaoka, Yoshiharu; Kanda, Keiji; Nakagome, Yoshihiro

    1999-01-01

    The reduced enrichment programs for the JRR-3M, JRR-4 and JMTR of Japan Atomic Energy Research Institute (JAERI) have been completed. The KUR of Kyoto University Research Reactor Institute (KURRI) has been partially completed and is still in progress under the Joint Study Program with Argonne National Laboratory (ANL). The JRR-3M commenced using LEU silicide fuel elements instead of LEU aluminide fuel elements in September, 1999. The Japanese Government approved a cancellation of the KUHFR Project in February 1991, and April 1994 the U.S. Government gave an approval to utilize HEU fuel in the KUR instead of the KUHFR. Therefore, the KUR will be operated with HEU fuel until March 2004, then the full core conversion with LEU silicide will be done. The first shipment of spent fuels since 1974 was done in August, 1999. (author)

  6. Basic studies for molten-salt reactor engineering in Japan

    International Nuclear Information System (INIS)

    Ishiguro, R.; Sugiyama, K.; Sakashita, H.

    1985-01-01

    A research project of nuclear engineering for the molten-salt reactor is underway which is supported by the Grant-in-Aid for Scientific Research of the Ministry of Education of Japan. At present, the major effort is devoted only to basic engineering problems because of the limited amount of the grant. The reporters introduce these and related studies that have been carrying out in Japanese universities. Discussions on the following four subjects are summerized in this report: a) Vapour explosion when hight temperature molten-salts are brought into direct contact with water. b) Measurements of exact thermophysical properties of molten-salt. c) Free convection heat transfer with uniform internal heat generation and a constant heating rate from the bottem. d) Stability of frozen salt film on the container surface. (author)

  7. Summary of the progress of reactor physics in Japan reviewing the activities related to NEA Committee on Reactor Physics

    International Nuclear Information System (INIS)

    Hirota, Jitsuya

    1984-09-01

    The progress of fast and thermal reactor physics, fusion neutronics and shielding researches in these twenty years can be clearly recognized in the reviews of reactor physics activities in Japan which had been perpared by the Special Committee on Reactor Physics: the joint committee under Atomic Energy Society of Japan and JAERI. Many topics of those discussed at the NEACRP meetings concerned fast reactor physics. Information exchange on the topics such as adjustment of group cross sections by integral data, central worth discrepancy, sodium void effect and heterogeneous core stimulated the researches in Japan. And achievements in Japan including those in the JAERI Fast Critical Facility FCA were reported and contributed largely to the international co-operation. In addition, the contribution from Japan was also made concerning a study of fusion blanket. Among various specialists' meetings recommended by NEACRP, those on nuclear data and benchmarks for reactor shielding were often held since 1973 and helpful to the progress of shielding researches in Japan. The Third Specialists' Meeting on Reactor Noise (SMORN-III) was held in Tokyo in 1981, indicating the recent progress in safety-related applications of reactor noise analysis. The NEACRP benchmark tests were quite useful to the progress of reactor physics in Japan, which included the benchmark calculations of BWR lattice cell, key parameters and burn-up characteristics of a large LMFBR, FBR and PWR shielding, and so on. It may be noted that the benchmark test on reactor noise analysis methods was successfully conducted by Japan in connection with SMORN-III. In addition, the co-operation was positively made to the compilation of light water lattice data, and the preparation of reviews on actinide production and burn-up, and blanket physics. (J.P.N.)

  8. Historical construction costs of global nuclear power reactors

    International Nuclear Information System (INIS)

    Lovering, Jessica R.; Yip, Arthur; Nordhaus, Ted

    2016-01-01

    The existing literature on the construction costs of nuclear power reactors has focused almost exclusively on trends in construction costs in only two countries, the United States and France, and during two decades, the 1970s and 1980s. These analyses, Koomey and Hultman (2007); Grubler (2010), and Escobar-Rangel and Lévêque (2015), study only 26% of reactors built globally between 1960 and 2010, providing an incomplete picture of the economic evolution of nuclear power construction. This study curates historical reactor-specific overnight construction cost (OCC) data that broaden the scope of study substantially, covering the full cost history for 349 reactors in the US, France, Canada, West Germany, Japan, India, and South Korea, encompassing 58% of all reactors built globally. We find that trends in costs have varied significantly in magnitude and in structure by era, country, and experience. In contrast to the rapid cost escalation that characterized nuclear construction in the United States, we find evidence of much milder cost escalation in many countries, including absolute cost declines in some countries and specific eras. Our new findings suggest that there is no inherent cost escalation trend associated with nuclear technology. - Highlights: •Comprehensive analysis of nuclear power construction cost experience. •Coverage for early and recent reactors in seven countries. •International comparisons and re-evaluation of learning. •Cost trends vary by country and era; some experience cost stability or decline.

  9. Transmutation of actinides in power reactors.

    Science.gov (United States)

    Bergelson, B R; Gerasimov, A S; Tikhomirov, G V

    2005-01-01

    Power reactors can be used for partial short-term transmutation of radwaste. This transmutation is beneficial in terms of subsequent storage conditions for spent fuel in long-term storage facilities. CANDU-type reactors can transmute the main minor actinides from two or three reactors of the VVER-1000 type. A VVER-1000-type reactor can operate in a self-service mode with transmutation of its own actinides.

  10. Power controlling method for BWR type reactors

    International Nuclear Information System (INIS)

    Yoshida, Kenji.

    1983-01-01

    Purpose: To enable reactor operation exactly following after an aimed curve in the high power resuming and maintaining period without failures in cladding tubes. Method: Upon recovery of the reactor power to a high power level after changing the reactor power from the high power to the low power level, control rod is operated under such conditions that the linear power density after operation of the control rod does not exceed the PC envelope in the low power period, and the core flow rate is coordinated to the control rod operation. The linear power density can be suppressed within an allowable linear power density by the above operation during high power resuming and maintaining period and, as the result, PCI failures can be prevented. (Kamimura, M.)

  11. Future of nuclear power in Japan - Development of next Generation LWRs

    Energy Technology Data Exchange (ETDEWEB)

    Yamada, Eiji; Yamamoto, T.; Kurosaki, K.; Ohga, Y.; Tsuzuki, K.; Kasai, S.; Tanaka, T.

    2010-09-15

    Japan's energy policies have been to decrease the oil portion and dependence on the Middle East for energy security, as well as satisfy environmental requirement. The report of 2008 targeted reducing GHG emission by 60-80% before 2050, and highlighted ''Cool Earth-Innovative Energy Technology Program'' featuring 21 innovative technologies. In this context nuclear power is expected as a core power source. In April 2008, ''Next Generation Light Water Power Reactor Development Program'' was launched with the IAE as the core organization in alliance with Japan's major vendors and in collaboration with METI and power utilities for the future of nuclear power.

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

    International Nuclear Information System (INIS)

    Paranjpe, S.R.

    1992-01-01

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

  13. Power generator in BWR type reactors

    International Nuclear Information System (INIS)

    Yoshida, Kenji.

    1984-01-01

    Purpose: To enable to perform stable and dynamic conditioning operation for nuclear fuels in BWR type reactors. Constitution: The conditioning operation for the nuclear fuels is performed by varying the reactor core thermal power in a predetermined pattern by changing the predetermined power changing pattern of generator power, the rising rate of the reactor core thermal power and the upper limit for the rising power of the reactor core thermal power are calculated and the power pattern for the generator is corrected by a power conditioning device such that the upper limit for the thermal power rising rate and the upper limit for the thermal power rising rate are at the predetermined levels. Thus, when the relation between the reactor core thermal power and the generator electrical power is fluctuated, the fluctuation is detected based on the variation in the thermal power rising rate and the limit value for the thermal power rising rate, and the correction is made to the generator power changing pattern so that these values take the predetermined values to thereby perform the stable conditioning operation for the nuclear fuels. (Moriyama, K.)

  14. Operational method for demonstrating fuel loading integrity in a reactor having accessible 235U fuel

    International Nuclear Information System (INIS)

    Ward, D.R.

    1979-07-01

    The Health Physics Research Reactor is a small pulse reactor at the Oak Ridge National Laboratory. It is desirable for the operator to be able to demonstrate on a routine basis that all the fuel pieces are present in the reactor core. Accordingly, a technique has been devised wherein the control rod readings are recorded with the reactor at delayed critical and corrections are made to compensate for the effects of variations in reactor height above the floor, reactor power, core temperature, and the presence of any massive neutron reflectors. The operator then compares these readings with the values expected based on previous operating experience. If this routine operational check suggests that the core fuel loading might be deficient, a more rigorous follow-up may be made

  15. Impacts on power reactor health physics programs

    International Nuclear Information System (INIS)

    Meyer, B.A.

    1991-01-01

    The impacts on power reactor health physics programs form implementing the revised 10 CFR Part 20 will be extensive and costly. Every policy, program, procedure and training lesson plan involving health physics will require changes and the subsequent retraining of personnel. At each power reactor facility, hundreds of procedures and thousands of people will be affected by these changes. Every area of a power reactor health physics program will be affected. These areas include; ALARA, Respiratory Protection, Exposure Control, Job Coverage, Dosimetry, Radwaste, Effluent Accountability, Emergency Planning and Radiation Worker Training. This paper presents how power reactor facilities will go about making these changes and gives possible examples of some of these changes and their impact on each area of power reactor health physics program

  16. Fast reactors: potential for power

    International Nuclear Information System (INIS)

    1983-02-01

    The subject is discussed as follows: basic facts about conventional and fast reactors; uranium economy; plutonium and fast reactors; cooling systems; sodium coolant; safety engineering; handling and recycling plutonium; safeguards; development of fast reactors in Britain and abroad; future progress. (U.K.)

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

    International Nuclear Information System (INIS)

    Vilim, R.B.; Park, Y.S.; Grandy, C.; Belch, H.; Dworzanski, P.; Misterka, J.

    2012-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-07-30

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

  19. 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.)

  20. A review of fast reactor program in Japan - April 1984

    International Nuclear Information System (INIS)

    Matsuno, Y.

    1984-01-01

    The fast breeder reactor development project in PNC has been in progress steadily in these eighteen years. Concerning the experimental fast reactor, JOYO, the MK-II core attained criticality on November 22, 1982 with 51 fuel assemblies, and received the ''Certificate of Inspection before Operation'' from Government Authority on March 31, 1983, after 100 hours operation with the rated output of 100 MW. Since then, the core has been utilized to implement irradiation bed characteristics test, and to irradiate fuels and structural materials especially for the prototype reactor MONJU. With respect to the prototype reactor MONJU, the installation permit was issued on May 27, 1983, from the prime minister, and the contracts of the first stage between PNC and fabricators were made recently. At the same time, almost all the licenses of preparatory construction works were issued by March 1983, and preparatory construction works were started in April 1983. On the other hand, conceptual design of a demonstration reactor is now under way in a close cooperation with concerned authorities and utilities, as well as investigations of the way of conducting necessary research and development

  1. Conceptual design of a Demonstration Tokamak Hybrid Reactor (DTHR), September 1978

    International Nuclear Information System (INIS)

    Kelley, J.L.

    1978-12-01

    The flexibility of the fusion hybrid reactor to function as a fuel production facility, power plant, waste disposal burner or combinations of all of these, as well as the reactor's ability to use proliferation resistant fuel cycles, has provided the incentive to assess the feasibility of a near-term demonstration plant. The goals for a Demonstration Tokamak Hybrid Reactor (DTHR) were established and an initial conceptual design was selected. Reactor performance and economics were evaluated and key developmental issues were assessed. The study has shown that a DTHR is feasible in the late 1980's, a significant quantity of fissile fuel could be produced from fertile thorium using present day fission reactor blanket technology, and a large number of commercially prototypical components and systems could be developed and operationally verified. The DTHR concept would not only serve as proof-of-principle for hybrid technology, but could be operated in the ignited mode and provide major advancements for pure fusion technology

  2. 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.

  3. Method and device for controlling reactor power

    International Nuclear Information System (INIS)

    Oohashi, Masahisa; Masuda, Hiroyuki.

    1982-01-01

    Purpose: To enable load following-up operation of a reactor adapted to perform power conditioning by the control of the liquid poison density in the core and by the control rods. Constitution: In a case where the reactor power is repeatedly changed in a reactor having a liquid poison density control device and control rods, the time period for the power control is divided depending on the magnitude of the change with time in the reactivity and the optimum values are set for the injection and removal amount of the liquid poison within the divided period. Then, most parts of the control required for the power change are alloted to the liquid poison that gives no effect on the power distribution while minimizing the movement of the control rods, whereby the power change in the reactor as in the case of the load following-up operation can be practiced with ease. (Kawakami, Y.)

  4. Concentrating Solar Power Gen3 Demonstration Roadmap

    Energy Technology Data Exchange (ETDEWEB)

    Mehos, Mark [National Renewable Energy Lab. (NREL), Golden, CO (United States); Turchi, Craig [National Renewable Energy Lab. (NREL), Golden, CO (United States); Vidal, Judith [National Renewable Energy Lab. (NREL), Golden, CO (United States); Wagner, Michael [National Renewable Energy Lab. (NREL), Golden, CO (United States); Ma, Zhiwen [National Renewable Energy Lab. (NREL), Golden, CO (United States); Ho, Clifford [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Kolb, William [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Andraka, Charles [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Kruizenga, Alan [Sandia National Lab. (SNL-CA), Livermore, CA (United States)

    2017-01-01

    Today's power-tower concentrating solar power (CSP) technology exists in large part as a result of Department of Energy (DOE) and utility industry funding of demonstration systems in the 1980s and 1990s. Today's most advanced towers are integrated with molten-salt thermal energy storage, delivering thermal energy at 565 degrees C for integration with conventional steam-Rankine cycles. The supercritical carbon dioxide power cycle has been identified as a likely successor to the steam-Rankine power cycle due to its potential for high efficiency when operating at elevated temperatures of 700 degrees C or greater. Over the course of the SunShot Initiative, DOE has supported a number of technology pathways that can operate efficiently at these temperatures and that hold promise to be reliable and cost effective. Three pathways - molten salt, particle, and gaseous - were selected for further investigation based on a two-day workshop held in August of 2016. The information contained in this roadmap identifies research and development challenges and lays out recommended research activities for each of the three pathways. DOE foresees that by successfully addressing the challenges identified in this roadmap, one or more technology pathways will be positioned for demonstration and subsequent commercialization within the next ten years. Based on current knowledge of the three power tower technologies, all three have the potential to achieve the SunShot goal of 6 cents/kilowatt-hour. Further development, modeling, and testing are now required to bring one or more of the technologies to a stage where integrated system tests and pilot demonstrations are feasible.

  5. Nuclear power reactors of new generation

    International Nuclear Information System (INIS)

    Ponomarev-Stepnoi, N.N.; Slesarev, I.S.

    1988-01-01

    The paper presents discussions on the following topics: fuel supply for nuclear power; expansion of the sphere of nuclear power applications, such as district heating; comparative estimates of power reactor efficiencies; safety philosophy of advanced nuclear plants, including passive protection and inherent safety concepts; nuclear power unit of enhanced safety for the new generation of nuclear power plants. The emphasis is that designers of new generation reactors face a complicated but technically solvable task of developing highly safe, efficient, and economical nuclear power sources having a wide sphere of application

  6. A prospect of fast reactor and related fuel cycle in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Nagata, Takashi [Japan Atomic Energy Agency, Ibaraki (Japan)

    2009-04-15

    JAEA has launched a new project 'Fast Reactor Cycle Technology Development'(FaCT) in cooperation with electric utilities. In this FaCT project, a combination of 'the Japanese sodium cooled loop type fast reactor with oxide fuel, the advanced aqueous reprocessing, and the simplified palletizing fuel fabrication systems' is adopted, where many innovative technologies with technical challenging issues are actively used in order to provide significant improvements in economic competitiveness, and enhancement of safety and reliability, sustainability, and nonproliferation. Fast reactor cycle technology will provide harmonic solutions for global issues of energy resources and environments, and is expected to contribute to sustainable development of the future society. Therefore, it was selected as one of key technologies of national importance in the third term (JPY2006-2010) 'Science and Technology Basic Plan' in March 2006 in Japan. The 'Nuclear Energy National Plan' in August 2006 states start up of a demonstration FR by around 2025 and deployment of a commercial FR before 2050, and start operating fuel cycle facilities when these reactors achieve consistency. Accordingly, we will decide about the adoption of innovative technologies by judging their applicability by 2010, and present the conceptual designs of commercial and demonstration FR cycle facilities by 2015 with the R and D plans to realize. In developing the FR cycle, 5 Party council, which consists of MEXt, MITI, electricity utilities, manufacturers, and JAEA, was established in July 2006 for moving forward on the commercialization smoothly. In this framework, users' requirements for the future R and D, a scenario of transition from light water reactor cycle to sodium cooled FR cycle, international collaboration, development schedule, demonstration steps, and so on are discussed. In this presentation, a prospect concerning the system design features of JSFR and a

  7. A report of the overall working group of the AEC Committee on Development of Advanced Power Reactors

    International Nuclear Information System (INIS)

    1979-01-01

    The AEC Committee on Development of Advanced Power Reactors was set up in April, 1978, following on the previous AEC Special Committee on Development of Advanced Power Reactors, in order to study on the complementary power reactors between current LWRs and future FBRs. The subjects of study by the overall working group are the status of advanced power reactors in views of the nuclear fuel cycle, the impacts on industries, the selection of reactor types under present international circumstances, and the evaluation of advanced power reactors in their technology and economy. The following matters are described: evaluations in view of the nuclear fuel cycle, i.e. the features of the ATR of Japan and CANDU reactors of Canada; international problems concerning nuclear nonproliferation and securing of uranium; problems in the diversification of power reactor types concerning the expenditure and technology; problems of technology in the ATR of Japan, CANDU reactors of Canada and Pu utilization for LWRs; and the economy of D 2 O power reactors, i.e. the ATR of Japan and CANDU reactors of Canada. (J.P.N.)

  8. Power supplyer for reactor coolant recycling pump

    International Nuclear Information System (INIS)

    Nara, Hiroshi; Okinaka, Yo.

    1991-01-01

    The present invention concerns a variable voltage/variable frequency static power source (static power source) used as a power source for a coolants recycling pump motor of a nuclear power plant. That is, during lower power operation such as start up or shutdown in which stoppage of the power source gives less effect to a reactor core, power is supplied from a power system, a main power generator connected thereto or a high voltage bus in the plant or a common high voltage bus to the static power source. However, during rated power operation, power is supplied from the output of an axially power generator connected with a main power generator having an extremely great inertia moment to the static power device. With such a constitution, the static power device is not stopped by the lowering of the voltage due to a thunderbolt falling accident or the like to a power-distribution line suddenly occurred in the power system. Accordingly, reactor core flowrate is free from rapid decrease caused by the reduction of rotation speed of the recycling pump. Accordingly, disadvantgages upon operation control in the reactor core is not caused. (I.S.)

  9. Problems in Siting Nuclear Power Plants in Japan and Efforts to Solve Them

    Energy Technology Data Exchange (ETDEWEB)

    Inouye, T. [Ministry of International Trade and Industry, Tokyo (Japan)

    1967-09-15

    The rapidly growing demand for energy in Japan will require a total capacity of 30 to 40 thousand MW(e) in nuclear power by 1985. Materialization of this development programme must naturally be supported by securing the requisite sites for the nuclear power plants. The following factors make siting of nuclear power plants more difficult in Japan than in any other country: a small, densely populated territory with little level land, that is already completely utilized for agricultural and/or industrial purposes; small rivers and an active marine-product industry developed along most of the seacoasts, both of which create difficult cooling-water problems; frequent earthquakes; and the fear of possible radioactivity, which prevails in the only nation in the world to have suffered from the atomic bomb. There are at present four nuclear power plants in operation or under construction in Japan with a total capacity of about 1.3 thousand MW(e). However, the plants in these construction programmes have been sited on the basis of taking the easiest course available although there were several possible solutions to choose from. It is pointed out here that the long-range nuclear power development programme will call for a fundamental solution to enable siting a large number of power plants under the adverse conditions in Japan. Accordingly, a study was made, which included quantitative analyses of reactor siting factors and suggested measures for solving the siting problems. The analyses were based on nuclear power plant sites assumed to be located on the seacoast and characterized by low-population density, desirable geology and favourable topography. It was assumed that seacoast siting was more economical than inland siting. Although the study was made by a general survey using maps, it was shown that approximately 10% of the total coastline areas would be eligible for reactor siting, but most of these areas in this case are located in the northern part of Japan, far from

  10. Development status on hydrogen production technology using high-temperature gas-cooled reactor at JAEA, Japan

    International Nuclear Information System (INIS)

    Shiozawa, Shusaku; Ogawa, Masuro; Hino, Ryutaro

    2006-01-01

    The high-temperature gas-cooled reactor (HTGR), which is graphite-moderated and helium-cooled, is attractive due to its unique capability of producing high temperature helium gas and its fully inherent reactor safety. In particular, hydrogen production using the nuclear heat from HTGR (up to 900 deg. C) offers one of the most promising technological solutions to curb the rising level of CO 2 emission and resulting risk of climate change. The interests in HTGR as an advanced nuclear power source for the next generation reactor, therefore, continue to rise. This is represented by the Japanese HTTR (High-Temperature Engineering Test Reactor) Project and the Chinese HTR-10 Project, followed by the international Generation IV development program, US nuclear hydrogen initiative program, EU innovative HTR technology development program, etc. To enhance nuclear energy application to heat process industries, the Japan Atomic Energy Agency (JAEA) has continued extensive efforts for development of hydrogen production system using the nuclear heat from HTGR in the framework of the HTTR Project. The HTTR Project has the objectives of establishing both HTGR technology and heat utilization technology. Using the HTTR constructed at the Oarai Research and Development Center of JAEA, reactor performance and safety demonstration tests have been conducted as planned. The reactor outlet temperature of 950 deg. C was successfully achieved in April 2004. For hydrogen production as heat utilization technology, R and D on thermo-chemical water splitting by the 'Iodine-Sulfur process' (IS process) has been conducted step by step. Proof of the basic IS process was made in 1997 on a lab-scale of hydrogen production of 1 L/h. In 2004, one-week continuous operation of the IS process was successfully demonstrated using a bench-scale apparatus with hydrogen production rate of 31 L/h. Further test using a pilot scale facility with greater hydrogen production rate of 10 - 30 m 3 /h is planned as

  11. Fusion neutronics plan in the development of fusion reactor. With the aim of realizing electric power

    Energy Technology Data Exchange (ETDEWEB)

    Nakamura, Hiroo; Morimoto, Yuichi; Ochiai, Kentarou; Sugimoto, Masayoshi; Nishitani, Takeo; Takeuchi, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2000-10-01

    On June 1992, Atomic Energy Commission in Japan has settled Third Phase Program of Fusion Research and Development to achieve self-ignition condition, to realize long pulse burning plasma and to establish basis of fusion engineering for demonstration reactor. This report describes research plan of Fusion Neutron Laboratory in JAERI toward a development of fusion reactor with an aim of realizing electric power. The fusion neutron laboratory has a fusion neutronics facility (FNS), intense fusion neutron source. The plan includes research items in the FNS; characteristics of shielding and breeding materials, nuclear characteristics of materials, fundamental irradiation process of insulator, diagnostics materials and structural materials, and development of in-vessel diagnostic technology. Upgrade of the FNS is also described. Also, the International Fusion Material Irradiation Facility (IFMIF) for intense neutron source to develop fusion materials is described. (author)

  12. Ten years of KRB Gundremmingen demonstration power station

    International Nuclear Information System (INIS)

    Facius, H. von; Ettemeyer, R.

    1976-01-01

    In August 1976 the first large nuclear power station in the Federal Republic, the KRB Gundremmingen plant with a net power of 237 MWe, has been in operation ten years. The construction of KRB as a demonstration plant was a major step forward on the way to the economic utilization of nuclear power for German utilities. Design and operation of the plant have decisively influenced the further development of the technology of light water reactors in the Federal Republic. Unlike the Kahl Experimental Nuclear Power Station (VAK), which was a test facility designed to generate experience and to train personnel, the decision to build KRB from the outset was conditional upon the fulfillment of economic criteria. Here are some of the aspects in which KRB has greatly influenced the development of nuclear power station technology: first application of internal steam-water separation instead of a steam drum with a water content of the steam of less than 1%; construction of a reactor buildung with all the necessary safety factors; solution of the corrosion and erosion problems linked with the use of a saturated steam turbine; special measures taken to prevent the turbine from speeding up due to post-evaporation effects after shutdown. Detailed comments are devoted to the subjects of availability, causes of failure and repair work. (orig.) [de

  13. Small and medium power reactors 1987

    International Nuclear Information System (INIS)

    1987-12-01

    This TECDOC follows the publication of TECDOC-347 Small and Medium Power Reactors Project Initiation Study - Phase I published in 1985 and TECDOC-376 Small and Medium Power Reactors 1985 published in 1986. It is mainly intended for decision makers in Developing Member States interested in embarking on a nuclear power programme. It consists of two parts: 1) Guidelines for the Introduction of Small and Medium Power Reactors in Developing Countries. These Guidelines were established during the Advisory Group Meeting held in Vienna from 11 to 15 May 1987. Their purpose is to review key aspects relating to the introduction of Small and Medium Power Reactors in developing countries; 2) Up-dated Information on SMPR Concepts Contributed by Supplier Industries. According to the recommendations of the Second Technical Committee Meeting on SMPRs held in Vienna in March 1985, this part contains the up-dated information formerly published in Annex I of the above mentioned TECDOC-347. Figs

  14. Small and medium power reactors 1987

    Science.gov (United States)

    1987-12-01

    This TECDOC follows the publication of TECDOC-347: Small and Medium Power Reactors (SMPR) Project Initiation Study, Phase 1, published in 1985 and TECDOC-376: Small and Medium Power Reactors 1985 published in 1986. It is mainly intended for decision makers in Developing Member States interested in embarking on a nuclear power program. It consists of two parts: (1) guidelines for the introduction of small and medium power reactors in developing countries. These Guidelines were established during the Advisory Group Meeting held in Vienna from 11 to 15 May 1987. Their purpose is to review key aspects relating to the introduction of small and medium power reactors in developing countries; (2) up-dated information on SMPR Concepts Contributed by Supplier Industries. According to the recommendations of the Second Technical Committee Meeting on SMPRs held in Vienna in March 1985, this part contains the up-dated information formerly published in Annex 1 of the above mentioned TECDOC-347.

  15. Physics design of experimental metal fuelled fast reactor cores for full scale demonstration

    International Nuclear Information System (INIS)

    Devan, K.; Bachchan, Abhitab; Riyas, A.; Sathiyasheela, T.; Mohanakrishnan, P.; Chetal, S.C.

    2011-01-01

    Highlights: → In this study we made physics designs of experimental metal fast reactor cores. → Aim is for full-scale demonstration of fuel assemblies in a commercial power reactor. → Minimum power with adequate safety is considered. → In addition, fuel sustainability is also considered in the design. → Sodium bonded U-Pu-6%Zr and mechanically bonded U-Pu alloys are used. - Abstract: Fast breeder reactors based on metal fuel are planned to be in operation for the year beyond 2025 to meet the growing energy demand in India. A road map is laid towards the development of technologies required for launching 1000 MWe commercial metal breeder reactors with closed fuel cycle. Construction of a test reactor with metallic fuel is also envisaged to provide full-scale testing of fuel sub-assemblies planned for a commercial power reactor. Physics design studies have been carried out to arrive at a core configuration for this experimental facility. The aim of this study is to find out minimum power of the core to meet the requirements of safety as well as full-scale demonstration. In addition, fuel sustainability is also a consideration in the design. Two types of metallic fuel pins, viz. a sodium bonded ternary (U-Pu-6% Zr) alloy and a mechanically bonded binary (U-Pu) alloy with 125 μm thickness zirconium liner, are considered for this study. Using the European fast reactor neutronics code system, ERANOS 2.1, four metallic fast reactor cores are optimized and estimated their important steady state parameters. The ABBN-93 system is also used for estimating the important safety parameters. Minimum achievable power from the converter metallic core is 220 MWt. A 320 MWt self-sustaining breeder metal core is recommended for the test facility.

  16. Improvement of nuclear power plant operation and maintenance in Japan

    International Nuclear Information System (INIS)

    Kazushige Hamazaki

    1987-01-01

    Following the inauguration of commercial nuclear power generation in Japan in 1966, capacity factors were held in the relatively low level until around 1975 due to initial-period troubles. With subsequent improvement, however, capacity factors have climbed steadily and recently been sustaining more than 70%. To obtain this successful result, a various kind of improvement have been made not only for the operation management area but also for the maintenance management area in conjunction with the successive effort to reflect the operating experiences to the early stage design. Nowadays nuclear generation has assumed increasing importance for Japan's electrical power needs, and is making a great contribution to stabilizing power supply costs. (author)

  17. Cermet-fueled reactors for multimegawatt space power applications

    International Nuclear Information System (INIS)

    Cowan, C.L.; Armijo, J.S.; Kruger, G.B.; Palmer, R.S.; Van Hoomisson, J.E.

    1988-01-01

    The cermet-fueled reactor has evolved as a potential power source for a broad range of multimegawatt space applications. In particular, the fast spectrum reactor concept can be used to deliver 10s of megawatts of electric power for continuous, long term, unattended operation, and 100s of megawatts of electric power for times exceeding several hundred seconds. The system can also be utilized with either a gas coolant in a Brayton power conversion cycle, or a liquid metal coolant in a Rankine power conversion cycle. Extensive testing of the cermet fuel element has demonstrated that the fuel is capable of operating at very high temperatures under repeated thermal cycling conditions, including transient conditions which approach the multimegawatt burst power requirements. The cermet fuel test performance is reviewed and an advanced cermet-fueled multimegawatt nuclear reactor is described in this paper

  18. Multiple microprocessor based nuclear reactor power monitor

    International Nuclear Information System (INIS)

    Lewis, P.S.; Ethridge, C.D.

    1979-01-01

    The reactor power monitor is a portable multiple-microprocessor controlled data acquisition device being built for the International Atomic Energy Association. Its function is to measure and record the hourly integrated operating thermal power level of a nuclear reactor for the purpose of detecting unannounced plutonium production. The monitor consists of a 3 He proportional neutron detector, a write-only cassette tape drive and control electronics based on two INTEL 8748 microprocessors. The reactor power monitor operates from house power supplied by the plant operator, but has eight hours of battery backup to cover power interruptions. Both the hourly power levels and any line power interruptions are recorded on tape and in memory. Intermediate dumps from the memory to a data terminal or strip chart recorder can be performed without interrupting data collection

  19. Power distribution forecasting device for reactors

    International Nuclear Information System (INIS)

    Tsukii, Makoto

    1981-01-01

    Purpose: To save expensive calculations on the forecasting of reactor power distribution. Constitution: Core status (CSD) such as entire coolant flow rate, pressures in the reactor, temperatures at the outlet and inlet and positions for control rods are inputted into a power distribution calculation device to calculate the power distribution based on physical models intermittently. Further, present power distribution is calculated based on in-core neutron flux measured values and CSD in a process control computer. Further, the ratio of the calculation results of the latter to those of the former is calculated, stored and inputted into a correction device to correct the forecast power distribution obtained by the power distribution calculation device. This enables to forecast the power distribution with excellent responsivity in the reactor site. (Furukawa, Y.)

  20. Inspection, evaluation and maintenance guidelines for reactor vessel internals in JAPAN

    International Nuclear Information System (INIS)

    Sakashita, Akihiro; Goto, Tomoya; Hirano, Shinro; Dozaki, Koji

    2010-01-01

    Inspection and Evaluation Guidelines for reactor internals has been taken into the Rules on Fitness-for–Service for Nuclear Power Plants of The Japan Society of Mechanical Engineers. It is a base of the maintenance plan of each Nuclear Power Plant. A plant maintenance methodology will have more importance to maintain the plant safety and stable plant operation. This paper introduces the systematization of the maintenance such as repair, replacement, preventive maintenance in these guidelines. Maintenance methodologies are classified follows. Repair: methodology to reinforce degraded parts by some methods or prevent progress of degradation of without replacement of the existing structure when the degradation of structure is actualized. Replacement: methodology to replace the existing structure with new one when the degradation of structure is actualized. Preventive maintenance : methodology to mitigate the damaged condition. When the maintenance methodologies are implemented in the actual plant, we have to consider the feedback of the inspection program and plant life management. (author)

  1. Power distribution monitor in a nuclear reactor

    International Nuclear Information System (INIS)

    Uematsu, Hitoshi

    1983-01-01

    Purpose: To enable accurate monitoring for the reactor power distribution within a short time in a case where abnormality occurs in in-core neutron monitors or in a case where the reactor core state changes after the calibration for the neutron monitors. Constitution: The power distribution monitor comprises a power distribution calculator adapted to be inputted counted values from a reactor core present state data instruments and calculate the neutron flux distribution in the reactor core and the power distribution based on previously incorporated physical models, an RCF calculator adapted to be inputted with the counted values from the in-core neutron monitors and the neutron flux distribution and the power distribution calculated in the power distribution calculator and compensate the counted errors included in the counted values form the in-core neutron monitors and the calculation errors included in the power distribution calculated in the power distribution calculator to thereby calculate the power distribution within the reactor core, and an input/output device for the input of the data required for said power distribution calculator and the display for the calculation result calculated in the RCF calculator. (Ikeda, J.)

  2. Power conditioning system for a nuclear reactor

    International Nuclear Information System (INIS)

    Higashigawa, Yuichi; Joge, Toshio.

    1981-01-01

    Purpose: To provide a power conditioning system for a BWR type reactor which has a function to be automatically operated within a range that the relationship between the heat power of the reactor and the electric power of an electric generator does not lose the safety of fuel by eliminating the unnecessary fluctuation of the power of the reactor. Constitution: A load request error signal fed from a conventional turbine control system to recirculation flow regulator is eliminated, and a reactor power conditioning system is newly provided, to which an electric generator power signal, a reactor average power area monitor signal and a load request signal are inputted. Thus, the load request signal is compared directly with the electric power of the electric generator, the recirculation flow rate is controlled by the compared result, and whether the correlation between the heat power of the reqctor and the electric power of the generator satisfies the correlation determined to prove the safety of fuel or not is checked. If this correlation is satisfied, the recirculation flow rate is merely automatically controlled. (Yoshino, Y.)

  3. 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)

  4. Identification of fast power reactivity effect in nuclear power reactor

    International Nuclear Information System (INIS)

    Efanov, A.I.; Kaminskas, V.A.; Lavrukhin, V.S.; Rimidis, A.P.; Yanitskene, D.Yu.

    1987-01-01

    A nuclear power reactor is an object of control with distributed parameters, characteristics of which vary during operation time. At the same time the reactor as the object of control has internal feedback circuits, which are formed as a result of the effects of fuel parameters and a coolant (pressure, temperature, steam content) on the reactor breeding properties. The problem of internal feedback circuit identification in a nuclear power reactor is considered. Conditions for a point reactor identification are obtained and algorithms of parametric identification are constructed. Examples of identification of fast power reactivity effect for the RBMK-1000 reactor are given. Results of experimental testing have shown that the developed method of fast power reactivity effect identification permits according to the data of normal operation to construct adaptive models for the point nuclear reactor, designed for its behaviour prediction in stationary and transition operational conditions. Therefore, the models considered can be used for creating control systems of nuclear power reactor thermal capacity (of RBMK type reactor, in particular) which can be adapted to the change in the internal feedback circuit characteristics

  5. Fuel combustion in thermal power plants in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Kotler, V.R.

    1983-11-01

    The position of black coal in the energy balance of Japan is discussed. About 75% of electric energy is produced by thermal power plants. Eighty-five per cent of electricity is produced by power plants fired with liquid fuels and 3% by coal fired plants. Coal production in Japan, the forecast coal import to the country by 1990 (132 Mt/year), proportion of coal imported from various countries, chemical and physical properties of coal from Australia, China and Japan are discussed. Coal classification used in Japan is evaluated. The following topics associated with coal combustion in fossil-fuel power plants in Japan are discussed: coal grindability, types of pulverizing systems, slagging properties of boiler fuel in Japan, systems for slag removal, main types of steam boilers and coal fired furnaces, burner arrangement and design, air pollution control from fly ash, sulfur oxides and nitrogen oxides, utilization of fly ash for cement production, methods for removal of nitrogen oxides from flue gas using ammonia and catalysts or ammonia without catalysts, efficiency of nitrogen oxide control, abatement of nitrogen oxide emission from boilers by flue gas recirculation and reducing combustion temperatures. The results of research into air pollution control carried out by the Nagasaki Technical Institute are reviewed.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-08-01

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

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

    International Nuclear Information System (INIS)

    Kanda, Keiji; Nishihara, Hedeaki; Shirai, Eiji; Oyamada, Rokuro; Sanokawa, Konomo

    1997-01-01

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

  8. MODERATOR ELEMENTS FOR UNIFORM POWER NUCLEAR REACTOR

    Science.gov (United States)

    Balent, R.

    1963-03-12

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

  9. A nuclear power reactor concept for Brazil

    International Nuclear Information System (INIS)

    Sefidvash, F.

    1980-01-01

    For the purpose of developing an independent national nuclear technology and effective manner of transferring such a technology, as well as developing a modern reactor, a new nuclear power reactor concept is proposed which is considered as a suitable and viable project for Brazil to support its development and finally construct its prototype as an indigeneous venture. (Author) [pt

  10. Conceptual design for Japan sodium-cooled fast reactor. (1) Current status of system design for JSFR

    International Nuclear Information System (INIS)

    Uto, Nariki; Sakai, Takaaki; Mihara, Takatsugu; Kotake, Shoji; Aoto, Kazumi; Toda, Mikio

    2009-01-01

    Japan Atomic Energy Agency is now conducting 'Fast Reactor Cycle Technology Development (FaCT)' project. In the FaCT project, the system design for JSFR has been carried out along the design categories such as safety design, reactor system, heat transport system, etc., together with research and developments (R and Ds) on innovative technologies to be adopted to JSFR for achieving economic competitiveness, enhanced safety and reliability. This paper describes the system design features of JSFR and a summary of the progresses of the design and R and Ds concerned with a compact reactor vessel, an innovative containment vessel, etc. The approach for the commercialization of fast reactors including discussion on a demonstration reactor for JSFR is also briefly described. (author)

  11. Progress report on fast breeder reactor development in Japan

    International Nuclear Information System (INIS)

    1976-05-01

    Following the completion of the in-argon, high temperature test, the in-sodium functional test of Joyo has set in. The fabrication of the equipments for monitoring the flow rate and temperature in the center channel and the power distribution was finished. The modification of design of the prototype fast breeder reactor Monju came into its phase 3. The interim report on the check-up and review of the Monju project by the Government is now ready. Various calculation codes were developed or are in development stage. The mock-up assembly FCA 7-1 has been built, which consists of the Pu-fueled sector region simulating the Monju core and the U-235-fueled driver region. Various reactor physics experiments have been carried out in this assembly. Also, the calculation methods for reactor physics parameters have been developed, and the detailed calculation on the main shield of Joyo was performed. The situation of the developments of the components for Joyo and Monju and the measuring and control systems is shown. Almost all the existing sodium test facilities in Oarai Engineering Center were in service without any trouble, and the new test facility named ''Carbon transfer test loop'' was commissioned. The progress in the fields of sodium technology, fuel and material, safety and steam generators is reported. (Kako, I.)

  12. Reactor experiments, workshops, and human resource development education simulating the Great East Japan Earthquake

    International Nuclear Information System (INIS)

    Horiguchi, Tetsuo; Yamamoto, Tomosada

    2012-01-01

    Kinki University Atomic Energy Research Institute has been implementing a social education program such as reactor experiments and training sessions for junior and senior high school teachers since 1987, and in recent years, it has been implementing an education program for common citizens. However, the Great East Japan Earthquake has made it necessary to consider not only the dissemination of accurate knowledge, but also responding to the anxiety on nuclear power. This paper explains the contents of the social contribution activities and workshops conducted at Kinki University Atomic Energy Research Institute, after the Great East Japan Earthquake and the Fukushima Daiichi Nuclear Power Station accident. As the activities that are carried out in addition to training sessions, it introduces the implementation state of telephone consultation about nuclear power, and earthquake reconstruction assistance advisory at Kawamata Town, Date-gun, Fukushima Prefecture. As workshop support, it reports human resource development education in the nuclear field at the university, activities at the workshops for junior/senior high school teachers and general public, and questionnaire survey at the time of the workshops. (A.O.)

  13. SEISMIC DESIGN CRITERIA FOR NUCLEAR POWER REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    Williamson, R. A.

    1963-10-15

    The nature of nuclear power reactors demands an exceptionally high degree of seismic integrity. Considerations involved in defining earthquake resistance requirements are discussed. Examples of seismic design criteria and applications of the spectrum technique are described. (auth)

  14. Recent activities of SOFC research development and demonstration in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Yokokawa, Harumi [National Institute of Advanced Industrial Science and Technology (AIST), Tsukuba, Ibaraki (Japan); Tokyo City Univ. (Japan). Advanced Research Labs.

    2010-07-01

    Currently Japanese efforts in developing the SOFC systems have two major targets; one is small SOFC cogeneration systems for residential houses or small business sites, the other being the SOFC-GT hybrid systems with an aim at larger stationary applications. The former activity exhibits impressively rapid progress in system development and demonstration in actual residential environment. On the other hand, the development of hybrid systems is not so rapid but successfully has continued to test the operation of the hybrid system in a 200 kW class. The common requirement for both applications is high durability such as 40,000-100,000 of life. To achieve a long life simultaneously with reasonably low price and high efficiency the NEDO project is going on to promote the cooperation among the stack developers, national institute and universities. To achieve well organized cooperation, it is highly required to create mutual reliance between industry and academic organizations. After several years' experience, progress has been made in maturing cooperation and in leading to many new insights into physicochemical understanding of degradation phenomena. Some of them will be reported. (orig.)

  15. Reactor power control method and device

    International Nuclear Information System (INIS)

    Fushimi, Atsushi; Ishii, Yoshihiko; Miyamoto, Yoshiyuki; Ishii, Kazuhiko; Kiyoharu, Norihiko; Aizawa, Yuko.

    1997-01-01

    The present invention provides a method and a device suitable to rise the temperature and increase the pressure of the reactor to an aimed pressure in accordance with an aimed value for a reactor water temperature changing rate in the course of rising temperature and increasing pressure of the reactor upon start up of a BWR type power plant. Namely, neutron fluxes in the reactor and the temperature of reactor water are detected respectively. The maximum value among the detected values for the neutron fluxes is detected. The reactor water temperature changing rate is calculated based on the detected values of the reactor water temperature, from which the maximum value of the reactor water temperature changing rate is detected. An aimed value for the neutron flux is calculated in accordance with both detected maximum values and the aimed value of the reactor water temperature changing rate. The position of control rods is adjusted in accordance with the aimed value for the calculated neutron flux. Then, an aimed value for the neutron flux for realizing the aimed value for the reactor water temperature changing rate can be obtained accurately with no influence of the sensitivity of the detected values of the neutron fluxes and the time delay of the reactor water temperature changing rate. (I.S.)

  16. Engineering experiences through nuclear power development in Japan

    International Nuclear Information System (INIS)

    Uchida, Hideo

    2004-01-01

    This keynote paper deals with: energy issues and nuclear power development in Japan, problems of radiation protection, licensing and safety regulations, research on LOCA and ECCS, stress corrosion cracks related to pressure vessels, nuclear fuel failures, steam generators, incidents, waste management and fuel cycle facilities. In conclusion it is stated that: on order to cope with global matters vitally affecting the electricity generation, taking into consideration Japanese specific energy issues, the nuclear power development has been an indispensable policy of Japan. In order to proceed with further development of nuclear power plants, it is necessary to obtain proper understanding by the public, showing assurance of the safety and reliable operation of nuclear power plants through daily plant operation. The nuclear safety issues should be considered from a global point of view. It is necessary to establish common safety standards which could harmonize the safety level of nuclear power plants in the world. The safety goal concerning severe accidents should be established as an internationally agreeable one. Japan has accumulated highly technological experience in maintenance of nuclear power plants. It is believed that the cumulative experiences in Japan can contribute to the further improvement of safety of nuclear power plants throughout the world, and for this aim a mutual information exchange should be encouraged

  17. Low power reactor for remote applications

    International Nuclear Information System (INIS)

    Meier, K.L.; Palmer, R.G.; Kirchner, W.L.

    1985-01-01

    A compact, low power reactor is being designed to provide electric power for remote, unattended applications. Because of the high fuel and maintenance costs for conventional power sources such as diesel generators, a reactor power supply appears especially attractive for remote and inaccessible locations. Operating at a thermal power level of 135 kWt, the power supply achieves a gross electrical output of 25 kWe from an organic Rankine cycle (ORC) engine. By intentional selection of design features stressing inherent safety, operation in an unattended mode is possible with minimal risk to the environment. Reliability is achieved through the use of components representing existing, proven technology. Low enrichment uranium particle fuel, in graphite core blocks, cooled by heat pipes coupled to an ORC converter insures long-term, virtually maintenance free, operation of this reactor for remote applications. 10 refs., 7 figs., 3 tabs

  18. MIT research reactor. Power uprate and utilization

    Energy Technology Data Exchange (ETDEWEB)

    Hu, Lin-Wen [Nuclear Reactor Laboratory, Massachusetts Inst. of Technology, Cambridge, MA (United States)

    2012-03-15

    The MIT Research Reactor (MITR) is a university research reactor located on MIT campus. and has a long history in supporting research and education. Recent accomplishments include a 20% power rate to 6 MW and expanding advanced materials fuel testing program. Another important ongoing initiative is the conversion to high density low enrichment uranium (LEU) monolithic U-Mo fuel, which will consist of a new fuel element design and power increase to 7 MW. (author)

  19. reactor power control using fuzzy logic

    International Nuclear Information System (INIS)

    Ahmed, A.E.E.

    2001-01-01

    power stabilization is a critical issue in nuclear reactors. convention pd- controller is currently used in egypt second testing research reactor (ETRR-2). two fuzzy controllers are proposed to control the reactor power of ETRR-2 reactor. the design of the first one is based on a set of linguistic rules that were adopted from the human operators experience. after off-line fuzzy computations, the controller is a lookup table, and thus, real time controller is achieved. comparing this f lc response with the pd-controller response, which already exists in the system, through studying the expected transients during the normal operation of ETRR-2 reactor, the simulation results show that, fl s has the better response, the second controller is adaptive fuzzy controller, which is proposed to deal with system non-linearity . The simulation results show that the proposed adaptive fuzzy controller gives a better integral square error (i se) index than the existing conventional od controller

  20. Reactor Power Meter type SG-8

    Energy Technology Data Exchange (ETDEWEB)

    Glowacki, S W

    1981-01-01

    The report describes the principle and electronic circuits of the Reactor Power Meter type SG-8. The gamma radiation caused by the activity of the reactor first cooling circuit affectes the ionization chamber being the detector of the instrument. The output detector signal direct current is converted into the frequency of electric pulses by means of the current-to-frequency converter. The output converter frequency is measured by the digital frequency meter: the number of measured digits in time unit is proportional to the reactor power.

  1. A review on the utilization of the Japan materials testing reactor (JMTR)

    Energy Technology Data Exchange (ETDEWEB)

    Kim, D. H.; Kang, Y. H.; Kim, B. G.; Choo, K. N.; Oh, J. M.; Park, S. J.; Shin, Y. T

    1999-04-01

    The HANARO has possessed the potential capability for the testing of materials and fuels since the beginning of its operation in 1995. Recently, this reactor has contributed to various activities in nuclear power research in Korea. We need the recent technical data of developed countries to support these activities in nuclear power. Most of the developed countries in nuclear power have more than thirty years' experience in the irradiation test of nuclear fuel and material for performing their complicated in-core measurements of the change of material properties. They also have developed various types of sensors, equipment and techniques. This report describes the status of utilization of the irradiation facilities of the Japan Materials Testing Reactor(JMTR). It also describes the recent efforts of the JMTR in order to develop new irradiation test techniques. It will be our great pleasure for this report to help a broad range of people understand the generic contents (JMTR utilization, new techniques) of the JMTR. (author)

  2. A review on the utilization of the Japan materials testing reactor (JMTR)

    International Nuclear Information System (INIS)

    Kim, D. H.; Kang, Y. H.; Kim, B. G.; Choo, K. N.; Oh, J. M.; Park, S. J.; Shin, Y. T.

    1999-04-01

    The HANARO has possessed the potential capability for the testing of materials and fuels since the beginning of its operation in 1995. Recently, this reactor has contributed to various activities in nuclear power research in Korea. We need the recent technical data of developed countries to support these activities in nuclear power. Most of the developed countries in nuclear power have more than thirty years' experience in the irradiation test of nuclear fuel and material for performing their complicated in-core measurements of the change of material properties. They also have developed various types of sensors, equipment and techniques. This report describes the status of utilization of the irradiation facilities of the Japan Materials Testing Reactor(JMTR). It also describes the recent efforts of the JMTR in order to develop new irradiation test techniques. It will be our great pleasure for this report to help a broad range of people understand the generic contents (JMTR utilization, new techniques) of the JMTR. (author)

  3. New generation of reactors for space power

    International Nuclear Information System (INIS)

    Boudreau, J.E.; Buden, D.

    1982-01-01

    Space nuclear reactor power is expected to enable many new space missions that will require several times to several orders of magnitude anything flown in space to date. Power in the 100-kW range may be required in high earth orbit spacecraft and planetary exploration. The technology for this power system range is under development for the Department of Energy with the Los Alamos National Laboratory responsible for the critical components in the nuclear subsystem. The baseline design for this particular nuclear sybsystem technology is described in this paper; additionally, reactor technology is reviewed from previous space power programs, a preliminary assessment is made of technology candidates covering an extended power spectrum, and the status is given of other reactor technologies

  4. Demonstrated operational and inherent safety of the prototype fast reactor (PFR)

    International Nuclear Information System (INIS)

    Smedley, J.A.; Gregory, C.V.; Judd, A.M.

    1983-01-01

    The Prototype Fast Reactor (PFR) is sited at Dounreay, on the north coast of Scotland in the United Kingdom, and has been in operation since 1974. Three aspects of the safety of the reactor are described, including the all-important practical consideration of operational safety, a demonstration of the limited consequences of a sodium/water reaction in a steam generator and the ability of the reactor to protect itself against highly improbable incidents. Attention is drawn to the low radiation levels in the plant and the correspondingly low dose rate to personnel. A feature of PFR operation has been the stable and predictable behaviour of its core together with the high degree of reliability exhibited by the engineered safety system. No failures have occurred within the standard driver charge but two experimental fuel pins suffered cladding failure, which was detected easily by the fission gas and delayed neutron detection systems. In the steam generating units sodium and water are separated by the single steel wall of the steam tubes. Although no under-sodium leak has occurred, an experimental programme is continuing and demonstrates that were any such leak to occur its consequences would be containable and would not result in the release of sodium to the environment or any breach of the reactor containment. The final section describes the inherent safety features of the reactor which enable it to survive a range of very improbable incidents even when the engineered safeguards fail. The features considered are natural circulation, which has been demonstrated by reactor experiment; the reactor's negative power coefficient, which, for example, enables the reactor to survive a complete loss of heat sink; and the durability of the fuel pins, demonstrated by a series of boiling experiments in the Dounreay Fast Reactor (DFR). (author)

  5. Compact reactor/ORC power source

    International Nuclear Information System (INIS)

    Meier, K.L.; Kirchner, W.L.; Willcutt, G.J.

    1986-01-01

    A compact power source that combines an organic Rankine Cycle (ORC) electric generator with a nuclear reactor heat source is being designed and fabricated. Incorporating existing ORC technology with proven reactor technology, the compact reactor/ORC power source offers high reliability while minimizing the need for component development. Thermal power at 125 kWt is removed from the coated particle fueled, graphite moderated reactor by heat pipes operating at 500 0 C. Outside the reactor vessel and connected to the heat pipes are vaporizers in which the toluene ORC working fluid is heated to 370 0 C. In the turbine-alternator-pump (TAP) combined-rotating unit, the thermal energy of the toluene is converted to 25 kWe of electric power. Lumped parameter systems analyses combined with a finite element thermal analysis have aided in the power source design. The analyses have provided assurance of reliable multiyear normal operation as well as full power operation with upset conditions, such as failed heat pipes and inoperative ORC vaporizers. Because of inherent high reliability, long life, and insensitivity to upset conditions, this power source is especially suited for use in remote, inaccessible locations where fuel delivery and maintenance costs are high. 10 refs

  6. A compact reactor/ORC power source

    International Nuclear Information System (INIS)

    Meier, K.L.; Kirchner, W.L.; Willcutt, G.J.

    1986-01-01

    A compact power source that combines an organic Rankine cycle (ORC) electric generator with a nuclear reactor heat source is being designed and fabricated. Incorporating existing ORC technology with proven reactor technology, the compact reactor/ORC power source offers high reliability while minimizing the need for componenet development. Thermal power at 125 kWt is removed from the coated particle fueled, graphite moderated reactor by heat pipes operating at 500 0 C. Outside the reactor vessel and connected to the heat pipes are vaporizers in which the toluene ORC working fluid is heated to 370 0 C. In the turbine-alternator-pump (TAP) combined-rotating unit, the thermal energy of the toluene is converted to 25 kWe of electric power. Lumped parameter systems analyses combined with a finite element thermal analyses combined with a finite element thermal analysis have aided in the power source design. The analysis have provided assurance of reliable multiyear normal operation as well as full power operation with upset conditions, such as failed heat pipes and inoperative ORC vaporizers. Because of inherent high reliability, long life, and insensitivity to upset conditions, this power source is especially suited for use in remote, inaccessible locations where fuel delivery and maintenance costs are high

  7. Demonstration-informative center based on research reactor IR-50 in heat regime

    International Nuclear Information System (INIS)

    Krupenina, Ph.

    2000-01-01

    Many problems exist in the nuclear field, but the most significant one is the public's mistrust of Nuclear Energy. Strong downfalls of the radiological culture affect public perception, the main paradox being the situation after Chernobyl. The task of creating a Demonstration-Informative Center (Minatom RF) on reactor IR-50 research is conducted by Research and Development Institute of Power Engineering (ENTEK). The IR-50 is situated on the grounds of the institute. It will be a unique event when the functional reactor is situated in the center of the city. The purposes of the Demonstration-Informative Center are discussed. (authors)

  8. Station Blackout Analysis of HTGR-Type Experimental Power Reactor

    Science.gov (United States)

    Syarip; Zuhdi, Aliq; Falah, Sabilul

    2018-01-01

    The National Nuclear Energy Agency of Indonesia has decided to build an experimental power reactor of high-temperature gas-cooled reactor (HTGR) type located at Puspiptek Complex. The purpose of this project is to demonstrate a small modular nuclear power plant that can be operated safely. One of the reactor safety characteristics is the reliability of the reactor to the station blackout (SBO) event. The event was observed due to relatively high disturbance frequency of electricity network in Indonesia. The PCTRAN-HTR functional simulator code was used to observe fuel and coolant temperature, and coolant pressure during the SBO event. The reactor simulated at 10 MW for 7200 s then the SBO occurred for 1-3 minutes. The analysis result shows that the reactor power decreases automatically as the temperature increase during SBO accident without operator’s active action. The fuel temperature increased by 36.57 °C every minute during SBO and the power decreased by 0.069 MW every °C fuel temperature rise at the condition of anticipated transient without reactor scram. Whilst, the maximum coolant (helium) temperature and pressure are 1004 °C and 9.2 MPa respectively. The maximum fuel temperature is 1282 °C, this value still far below the fuel temperature limiting condition i.e. 1600 °C, its mean that the HTGR has a very good inherent safety system.

  9. Demonstration tokamak fusion power plant for early realization of net electric power generation

    International Nuclear Information System (INIS)

    Hiwatari, R.; Okano, K.; Asaoka, Y.; Shinya, K.; Ogawa, Y.

    2005-01-01

    A demonstration tokamak fusion power plant Demo-CREST is proposed as the device for early realization of net electric power generation by fusion energy. The plasma configuration for Demo-CREST is optimized to satisfy the electric breakeven condition (the condition for net electric power, P e net = 0 MW) with the plasma performance of the ITER reference operation mode. This optimization method is considered to be suitable for the design of a demonstration power plant for early realization of net electric power generation, because the demonstration power plant has to ensure the net electric generation. Plasma performance should also be more reliably achieved than in past design studies. For the plasma performance planned in the present ITER programme, net electric power from 0 to 500 MW is possible with Demo-CREST under the following engineering conditions: maximum magnetic field 16 T, thermal efficiency 30%, NBI system efficiency 50% and NBI current drive power restricted to 200 MW. By replacing the blanket system with one of higher thermal efficiency, a net electric power of about 1000 MW is also possible so that the performance of the commercial plant with Demo-CREST can also be studied from the economic point of view. The development path from the experimental reactor 'ITER' to the commercial plant 'CREST' through the demonstration power plant 'Demo-CREST' is proposed as an example of the fast track concept. (author)

  10. Development step toward fusion power plant and role of experimental reactor ITER

    International Nuclear Information System (INIS)

    Hiwatari, Ryouji; Asaoka, Yoshiyuki; Okano, Kunihiko

    2005-01-01

    The development of fusion energy is going into the experimental reactor stage, and the thermal energy from the fusion reaction will be generated in a plant scale through the ITER (International Thermonuclear Experimental Reactor) project. The remaining critical issue toward the realization of fusion energy is to map out the development strategy. Recently early realization approach as for the fusion energy development is being discussed in Japan, Europe, and the United States. This approach implies that the devices for a Demo reactor and a proto-type reactor as seen in the fast breeder reactor are combined into a single device in order to advance the fusion energy development. On the other hand, a clear development road map for fusion energy hasn't been suggested yet, and whether that early realization approach is feasible or not is still ambiguous. In order to realize the fusion energy as an user-friendly energy system, the suggestion of the development missions and the road map from the user-side point of view is instructive not only to Japanese but also to other country's development policy after the ITER project. In this report, first of all, the development missions from the user's point of view have been structured. Second, the development target required to demonstrate net electric generation and to introduce the fusion energy into the market is investigated, respectively. This investigation reveals that the completion of the ITER reference operation gives the outlook toward the demonstration of net electric generation and that the completion of the ITER advanced operation gives the possibility to introduce the fusion energy into the market. At last, the electric demonstration power plant Demo-CREST and the commercial power plant CREST are proposed to construct the development road map for fusion energy. (author)

  11. Trends on R and D of the innovative nuclear reactors in Japan

    International Nuclear Information System (INIS)

    Kinoshita, Izumi

    2002-01-01

    In Japan, since LWRs introduced from U.S.A. began their business operations one by one from 1970 and 1971, their scale-up were carried out, to reach, at present, a condition on developments of ABWR-2 of 1700 MW class in output and APWR+. They are on a line of large scale LWR development aiming at further upgrading of their economical efficiency, safety, operability and maintenance by improving and developing conventional reactors. On the other hand, an innovative small scale reactor capable of siting at proximity of its markets and flexibly responsible to needs, a low decelerated spectrum reactor intending to effectively use the resources, an super-critical pressure reactor aiming at upgrading of thermal efficiency, a high temperature gas reactor aiming at hydrogen production using nuclear heat , and so on, and so forth, are investigated at a number of institutes. And, on the fast breeder reactor, some innovative investigations such as small-scale reactor, reactor using coolant except metal sodium, and so on, in addition to development of sodium cooling large-scale reactor, under the 'Actual use strategy survey research' progressed at a center of the Japan Nuclear Cycle Development Institute, are promoted. Here were outlined on trends of R and D on various innovative reactors under classification of water cooling reactor, gas cooling reactor, and liquid metal cooling reactor. (G.K.)

  12. Foundations of a long-term strategy for nuclear power development in Japan

    International Nuclear Information System (INIS)

    Murata, H.

    1975-01-01

    A long-term strategy for nuclear power developments in Japan is proposed. The situation in the world has greatly changed in the recent years due to the rise in oil prices as well as the considerable concern about the environmental problems caused by the nuclear power plants. Stress is being placed on the harmonization with the environmental protection rather than on the economical generation of the nuclear power. In order to meet the future requirements, five systems are given for the short, medium and long ranges beyond the year 2000. For the final stage a system is proposed that combines fusion-fission hybrid reactors with very high temperature gas cooled reactors to supply clean energy. (author)

  13. A Review of the Fast Reactor Programme in Japan

    International Nuclear Information System (INIS)

    Matsuno, Y.

    1987-01-01

    This report covers the activities of the experimental fast reactor JOYO from April 1985 to December 1986. Five duty cycles operation from the 8th to 12th was carried out. After the 12th duty cycle operation natural circulation test from full power level of 100 MWt. The 6th annual inspection was started from the end of December 1986. Works for maintenance or modification of the plant will be done on such system as the Cesium trap in order to enable trap of Cesium which will be released from failed fuel during reactor operation, exchange of cold trap of lry coolant system in order to get a higher trapping efficiency and reloading of the on-line irradiation test assembly (INTA). Besides the operation and maintenance works, following items were conducted during the period. 1. Development of operation and maintenance supporting systems using computers. 2. Data supply to Centralized Reliability Data Organization (CREDO) of U.S.A. based on the agreement between DOE and PNC. 3. Obtaining of the license to extend the fuel maximum burnup from 50.000 MWd/t to 75.000 MWd/t. and for increase of uranium enrichment of the fuel from. 12wt% to 18wt%. Operation history of JOYO is illustrated

  14. Reference reactor module for NASA's lunar surface fission power system

    International Nuclear Information System (INIS)

    Poston, David I.; Kapernick, Richard J.; Dixon, David D.; Werner, James; Qualls, Louis; Radel, Ross

    2009-01-01

    Surface fission power systems on the Moon and Mars may provide the first US application of fission reactor technology in space since 1965. The Affordable Fission Surface Power System (AFSPS) study was completed by NASA/DOE to determine the cost of a modest performance, low-technical risk surface power system. The AFSPS concept is now being further developed within the Fission Surface Power (FSP) Project, which is a near-term technology program to demonstrate system-level TRL-6 by 2013. This paper describes the reference FSP reactor module concept, which is designed to provide a net power of 40 kWe for 8 years on the lunar surface; note, the system has been designed with technologies that are fully compatible with a Martian surface application. The reactor concept uses stainless-steel based. UO 2 -fueled, pumped-NaK fission reactor coupled to free-piston Stirling converters. The reactor shielding approach utilizes both in-situ and launched shielding to keep the dose to astronauts much lower than the natural background radiation on the lunar surface. The ultimate goal of this work is to provide a 'workhorse' power system that NASA can utilize in near-term and future Lunar and Martian mission architectures, with the eventual capability to evolve to very high power, low mass systems, for either surface, deep space, and/or orbital missions.

  15. Thorium utilization in power reactors

    International Nuclear Information System (INIS)

    Saraceno; Marcos.

    1978-10-01

    In this work the recent (prior to Aug, 1976) literature on thorium utilization is reviewed briefly and the available information is updated. After reviewing the nuclear properties relevant to the thorium fuel cycle we describe briefly the reactor systems that have been proposed using thorium as a fertile material. (author) [es

  16. Bottom reflector for power reactors

    International Nuclear Information System (INIS)

    Elter, C.; Kissel, K.F.; Schoening, J.; Schwiers, H.G.

    1982-01-01

    In pebble bed reactors erosion and damage due fuel elements movement on the surface of the bottom reflector should be minimized. This can be achieved by chamfering and/or rounding the cover edges of the graphite blocks and the edges between the drilled holes and the surface of the graphite block. (orig.) [de

  17. 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

  18. Introduction to the neutron kinetics of nuclear power reactors

    CERN Document Server

    Tyror, J G; Grant, P J

    2013-01-01

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

  19. Parliament votes against building fifth power reactor

    International Nuclear Information System (INIS)

    Anon.

    1993-01-01

    After a heated three-day debate, Finland's parliament voted on September 24 to reject the proposal to build the country's fifth nuclear power reactor. As predicted, the vote was close: 107 voted against more nuclear power, 90 were in favor, two members of the 200-seat parliament were not present, and the speaker did not vote

  20. 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

  1. TerraPower, Bill Gates' reactor

    International Nuclear Information System (INIS)

    Guidez, J.

    2016-01-01

    TerraPower is a traveling wave reactor, it means that the reactor gradually converts non fissile material into the fuel it needs and the active part of the core progressively moves through the core leaving spent fuel behind. The last design of the TerraPower shows that it will use depleted uranium as fuel and that its core will need reloading every 10 years. Re-arrangement of the nuclear fuel will have to be made every 18 months to keep the core reactive. Metallic nuclear fuels will be used as they allow the highest breeding rates. It appears that apart from the very specific configuration of the core, the TerraPower is a reactor very similar to sodium-cooled fast reactors. Neutron transport inside traveling wave reactor core is complex and simulations show that the piling-up of fission product tends to kill the chain reaction and a continuous neutron addition may be necessary to keep the reactor going. A large part of the TerraPower feasibility studies concerns neutron transport inside its core. (A.C.)

  2. Radioactive waste management and disposal scenario for fusion power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Tabara, Takashi; Yamano, Naoki [Sumitomo Atomic Energy Industries Ltd., Tokyo (Japan); Seki, Yasushi; Aoki, Isao

    1997-10-01

    The environmental and economic impact of radioactive waste (radwaste) generated from fusion power reactors using five types of structural materials and a light water reactor (LWR) have been evaluated and compared. At first, the amount and the radioactive level of the radwaste generated in five fusion reactors ware evaluated by an activation calculation code. Next, a possible radwaste disposal scenario applicable to fusion radwaste in Japan is considered and the disposal cost evaluated under certain assumptions. The exposure doses are evaluated for the skyshine of gamma-rays during the disposal operation, groundwater migration scenario during the institutional control period of 300 years and future site use scenario after the institutional period. The radwaste generated from a typical LWR was estimated based on a literature survey and the disposal cost was evaluated using the same assumptions as for the fusion reactors. It is found that the relative cost of disposal is strongly dependent on the cost for interim storage of medium level waste of fusion reactors and the cost of high level waste for the LWR. (author)

  3. Analysis on the Electric Power Supply - Demand Measures of Japan in 2011 Summer after Earthquake and Tsunami

    International Nuclear Information System (INIS)

    Lee, Y. E.; Chang, H. S.

    2011-01-01

    Only 12 of 54 nuclear reactors are in operation as of September 1, 2011 in the wake of the earthquake and tsunami in Japan. The share of nuclear power in the nation's installation capacity fell to about 14% in August from about 30% before March 11, 2011. Government or many of research institutes estimated that the power supply system in Japan would fall to the minus reserve margin, if the nuclear power stations could not be restarted as scheduled. However, the current situation of power supply system in Japan is less severe than expected before, because the power companies and public have engaged in various diligent efforts to boost supply capacity or reduce demand in response to the electric power crisis. This paper aims to analyze the how much Japan electric power supply system depends on the nuclear power, what kinds of countermeasures of electric power supply-demand are taken by electricity companies in summer time to avoid the blackouts and why the saving electricity in Japan could be possible unlike Korea. Insights from this paper would be taken into account in the long term energy planning, even though the further study in depth should be followed

  4. Public acceptance (PA) activities of nuclear power in Japan

    International Nuclear Information System (INIS)

    Yamada, Masafumi; Iguchi, Tatsuro

    1993-10-01

    At the first part of presentation present status of nuclear power development in Japan is described. Then results of poll on nuclear energy acceptance by population are analyzed. Further, current activities and future efforts directed to broad understanding by people benefits of nuclear energy are described. 6 figs

  5. Gaseous fuel reactors for power systems

    Science.gov (United States)

    Kendall, J. S.; Rodgers, R. J.

    1977-01-01

    Gaseous-fuel nuclear reactors have significant advantages as energy sources for closed-cycle power systems. The advantages arise from the removal of temperature limits associated with conventional reactor fuel elements, the wide variety of methods of extracting energy from fissioning gases, and inherent low fissile and fission product in-core inventory due to continuous fuel reprocessing. Example power cycles and their general performance characteristics are discussed. Efficiencies of gaseous fuel reactor systems are shown to be high with resulting minimal environmental effects. A technical overview of the NASA-funded research program in gaseous fuel reactors is described and results of recent tests of uranium hexafluoride (UF6)-fueled critical assemblies are presented.

  6. Fractional power operation of tokamak reactors

    International Nuclear Information System (INIS)

    Mau, T.K.; Vold, E.L.; Conn, R.W.

    1986-01-01

    Methods to operate a tokamak fusion reactor at fractions of its rated power, identify the more effective control knobs and assess the impact of the requirements of fractional power operation on full power reactor design are explored. In particular, the role of burn control in maintaining the plasma at thermal equilibrium throughout these operations is studied. As a prerequisite to this task, the critical physics issues relevant to reactor performance predictions are examined and some insight into their impact on fractional power operation is offered. The basic tool of analysis consists of a zero-dimensional (0-D) time-dependent plasma power balance code which incorporates the most advanced data base and models in transport and burn plasma physics relevant to tokamaks. Because the plasma power balance is dominated by the transport loss and given the large uncertainty in the confinement model, the authors have studied the problem for a wide range of energy confinement scalings. The results of this analysis form the basis for studying the temporal behavior of the plasma under various thermal control mechanisms. Scenarios of thermally stable full and fractional power operations have been determined for a variety of transport models, with either passive or active feedback burn control. Important power control parameters, such as gas fueling rate, auxiliary power and other plasma quantities that affect transport losses, have also been identified. The results of these studies vary with the individual transport scaling used and, in particular, with respect to the effect of alpha heating power on confinement

  7. Power control device in nuclear reactor

    International Nuclear Information System (INIS)

    Koyama, Kazuaki.

    1981-01-01

    Purpose: To enable smooth power changes in power conditioning systems by calculating forecast values for the neutron flux distribution and power distribution and by controlling the driving speed of control rods so as to correspond the forecast values with aimed values. Constitution: Control rod position is detected by a position detector and sent to a control computer as the position information. At the same time, the neutron flux distribution information is obtained by the neutron monitors, the power distribution information is obtained by a reactor power computer and they are outputted to the control computer. The control computer calculates the forecast values for the neutron flux distribution and the reactor power distribution from the information, and compares them with the aimed values from a setter and then outputs control signals so as to correspond the forecast values with the aimed values. The control rods can be inserted in appropriate velocity by the control signals. (Horiuchi, T.)

  8. Current status and future perspective of upgrading on light water reactors in Japan

    International Nuclear Information System (INIS)

    Tateki, Hoshion

    1991-01-01

    Energy demands in the world would be expected to increase constantly furthermore. In this situation the role of nuclear energy has become more important than it used to be from the view of the countermeasure for the global warming issue resulting from the exhausted gases causing the greenhouse effect in addition to the view of the energy securities. In also Japan, the energy demand-supply perspective plan issued on October,1990 which aims to expand the capacity of nuclear power plants to 72,500MWe as a target. by 2010. (see table.l) Present Light Water Reactors(LWRs) have been experienced and this type of reactor will remain as major stream of nuclear power generation for a long time all over the world. Considering above perspective, Advisory Committee for Energy of MITI(Ministry of International Trade and Industry) has investigated the technology and technical development of coming LWRs and concluded the report on June of this year. In this paper the contents of above report are introduced

  9. Operating data monitoring and fatigue evaluation systems and findings for boiling water reactors in Japan

    International Nuclear Information System (INIS)

    Maekawa, O.; Kanazawa, Y.; Takahashi, Y.; Tani, M.

    1995-01-01

    The reactor pressure vessel (RPV) is one of the most critical components of a boiling water reactor (BWR) when utilities think about plant life extension (PLEX). Design stress analysis sometimes reports very high fatigue usage factors for such portions of RPVs as stud bolt, feedwater nozzle and support skirt.In order to evaluate design margin and to eliminate excessive conservatism in this design analysis to pave the way for PLEX, Japanese BWR utilities jointly with BWR manufacturers in Japan established a programme (1) to acquire plant operational data on line for specific parameters used in stress analysis, (2) to evaluate margin in the design using measured plant data best estimate boundary conditions for stress analysis, and (3) to establish a simplified fatigue analysis method for BWR RPV.A plant data acquisition system, named OPEDAS, has been developed and installed in Tokyo Electric Power Company's 1100MWe BWR at its Kashiwazaki-Kariwa nuclear power plant. Best estimate stress analysis using measured in-plant data has been carried out and the results show considerable margin in fatigue usage factor over the design. A simplified fatigue analysis method using in-plant data has been developed with the Green's function, although some limitations have been identified for its use. ((orig.))

  10. Decline of Civilian power in Japan's Defense Policy

    OpenAIRE

    SEBATA, TAKAO; 瀬端, 孝夫

    2006-01-01

    Over the last thirty years, military official of Japan's Self-Defense Forces (SDF) have begun voicing their opinions more aggressively, exerting influence and power in such areas as: the right of personnel management, budget formulation, organization, and defense policy decision-making. Due to the enhanced status of military officials, change of power balance between civilians and military is taking place. This paper examines civil-military relations in the above areas and discusses how this ...

  11. The Swedish Zero Power Reactor R0

    Energy Technology Data Exchange (ETDEWEB)

    Landergaard, Olof; Cavallin, Kaj; Jonsson, Georg

    1961-05-15

    The reactor R0 is a critical facility built for heavy water and natural uranium or fuel of low enrichment,, The first criticality was achieved September 25, 1959. During a first period of more than two years the R0 will be operated as a bare reactor in order to simplify interpretation of results. The reactor tank is 3. 2 m high and 2. 25 m in diameter. The fuel suspension system is quite flexible in order to facilitate fuel exchange and lattice variations. The temperature of the water can be varied between about 10 and 90 C by means of a heater and a cooler placed in the external circulating system. The instrumentation of the reactor has to meet the safety requirements not only during operation but also during rearrangements of the core in the shut-down state. Therefore, the shut-down state is always defined by a certain low 'safe' moderator level in the reactor tank. A number of safety rods are normally kept above the moderator ready for action. For manual or automatic control of the reactor power a specially designed piston pump is needed, by which the moderator level is varied. The pump speed is controlled from the reactor power error by means of a Ward-Leonard system. Moderator level measurement is made by means of a water gauge with an accuracy of {+-} 0. 1 mm.

  12. The Swedish Zero Power Reactor R0

    International Nuclear Information System (INIS)

    Landergaard, Olof; Cavallin, Kaj; Jonsson, Georg

    1961-05-01

    The reactor R0 is a critical facility built for heavy water and natural uranium or fuel of low enrichment,, The first criticality was achieved September 25, 1959. During a first period of more than two years the R0 will be operated as a bare reactor in order to simplify interpretation of results. The reactor tank is 3. 2 m high and 2. 25 m in diameter. The fuel suspension system is quite flexible in order to facilitate fuel exchange and lattice variations. The temperature of the water can be varied between about 10 and 90 C by means of a heater and a cooler placed in the external circulating system. The instrumentation of the reactor has to meet the safety requirements not only during operation but also during rearrangements of the core in the shut-down state. Therefore, the shut-down state is always defined by a certain low 'safe' moderator level in the reactor tank. A number of safety rods are normally kept above the moderator ready for action. For manual or automatic control of the reactor power a specially designed piston pump is needed, by which the moderator level is varied. The pump speed is controlled from the reactor power error by means of a Ward-Leonard system. Moderator level measurement is made by means of a water gauge with an accuracy of ± 0. 1 mm

  13. Advances in ICF power reactor design

    International Nuclear Information System (INIS)

    Hogan, W.J.; Kulcinski, G.L.

    1985-01-01

    Fifteen ICF power reactor design studies published since 1980 are reviewed to illuminate the design trends they represent. There is a clear, continuing trend toward making ICF reactors inherently safer and environmentally benign. Since this trend accentuates inherent advantages of ICF reactors, we expect it to be further emphasized in the future. An emphasis on economic competitiveness appears to be a somewhat newer trend. Lower cost of electricity, smaller initial size (and capital cost), and more affordable development paths are three of the issues being addressed with new studies

  14. Containment and surveillance techniques at power reactors

    International Nuclear Information System (INIS)

    Stirling, A.J.

    1982-01-01

    This session will provide participants with an understanding of the functions of safeguards equipment at power reactors, including equipment for fuel accounting, video and film surveillance, diversion monitoring, and containment and surveillance of irradiated fuel in storage. In addition, some appreciation of the impact that reactor safeguards have on the plant operator will be gained. From this, participants will be able to ensure that a reactor safeguards system meets their nation's international and national nonproliferation objectives with a minimum of interference to plant operations

  15. Different types of power reactors and provenness

    International Nuclear Information System (INIS)

    Goodman, E.I.

    1977-01-01

    The lecture guides the potential buyer in the selection of a reactor type. Recommended criteria regarding provenness, licensability, and contractual arrangements are defined and discussed. Tabular data summarizing operating experience and commercial availability of units are presented and discussed. The status of small and medium power reactors which are of interest to many developing countries is presented. It is stressed that each prospective buyer will have to establish his own criteria based on specific conditions which will be applied to reactor selection. In all cases it will be found that selection, either pre-selection of bidders or final selection of supplier, will be a fairly complex evaluation. (orig.) [de

  16. Safety of next generation power reactors

    International Nuclear Information System (INIS)

    Anon.

    1988-01-01

    This book is organized under the following headings: Future needs of utilities regulators, government, and other energy users, PRA and reliability, LMR concepts, LWR design, Advanced reactor technology, What the industry can deliver: advanced LWRs, High temperature gas-cooled reactors, LMR whole-core experiments, Advanced LWR concepts, LWR technology, Forum: public perceptions, What the industry can deliver: LMRs and HTGRs, Criteria and licensing, LMR modeling, Light water reactor thermal-hydraulics, LMR technology, Working together to revitalize nuclear power, Appendix A, luncheon address, Appendix B, banquet address

  17. Scoping and sensitivity analyses for the Demonstration Tokamak Hybrid Reactor (DTHR)

    International Nuclear Information System (INIS)

    Sink, D.A.; Gibson, G.

    1979-03-01

    The results of an extensive set of parametric studies are presented which provide analytical data of the effects of various tokamak parameters on the performance and cost of the DTHR (Demonstration Tokamak Hybrid Reactor). The studies were centered on a point design which is described in detail. Variations in the device size, neutron wall loading, and plasma aspect ratio are presented, and the effects on direct hardware costs, fissile fuel production (breeding), fusion power production, electrical power consumption, and thermal power production are shown graphically. The studies considered both ignition and beam-driven operations of DTHR and yielded results based on two empirical scaling laws presently used in reactor studies. Sensitivity studies were also made for variations in the following key parameters: the plasma elongation, the minor radius, the TF coil peak field, the neutral beam injection power, and the Z/sub eff/ of the plasma

  18. Space nuclear reactor power plants

    International Nuclear Information System (INIS)

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

    1980-01-01

    Requirements for electrical and propulsion power for space are expected to increase dramatically in the 1980s. Nuclear power is probably the only source for some deep space missions and a major competitor for many orbital missions, especially those at geosynchronous orbit. Because of the potential requirements, a technology program on space nuclear power plant components has been initiated by the Department of Energy. The missions that are foreseen, the current power plant concept, the technology program plan, and early key results are described

  19. Cobalt-60 production in CANDU power reactors

    International Nuclear Information System (INIS)

    Malkoske, G.R.; Norton, J.L.; Slack, J.

    2002-01-01

    MDS Nordion has been supplying cobalt-60 sources to industry for industrial and medical purposes since 1946. These cobalt-60 sources are used in many market and product segments, but are primarily used to sterilize single-use medical products including; surgical kits, gloves, gowns, drapes, and cotton swabs. Other applications include sanitization of cosmetics, microbial reduction of pharmaceutical raw materials, and food irradiation. The technology for producing the cobalt-60 isotope was developed by MDS Nordion and Atomic Energy of Canada Limited (AECL) almost 55 years ago using research reactors at the AECL Chalk River Laboratories in Ontario, Canada. The first cobalt-60 source produced for medical applications was manufactured by MDS Nordion and used in cancer therapy. The benefits of cobalt-60 as applied to medical product manufacturing, were quickly realized and the demand for this radioisotope quickly grew. The same technology for producing cobalt-60 in research reactors was then designed and packaged such that it could be conveniently transferred to a utility/power reactor. In the early 1970's, in co-operation with Ontario Power Generation (formerly Ontario Hydro), bulk cobalt-60 production for industrial irradiation applications was initiated in the four Pickering A CANDU reactors. As the demand and acceptance of sterilization of medical products grew, MDS Nordion expanded its bulk supply by installing the proprietary Canadian technology for producing cobalt-60 in additional CANDU reactors. CANDU is unique among the power reactors of the world, being heavy water moderated and fuelled with natural uranium. They are also designed and supplied with stainless steel adjusters, the primary function of which is to shape the neutron flux to optimize reactor power and fuel bum-up, and to provide excess reactivity needed to overcome xenon-135 poisoning following a reduction of power. The reactor is designed to develop full power output with all of the adjuster

  20. 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

  1. Progress report on fast breeder reactor development in Japan, July - September 1977

    International Nuclear Information System (INIS)

    1978-06-01

    As for the experimental fast breeder reactor ''Joyo'', the low power performance tests have been continued, and the measurements of reactor noise, the reactivity of fuel assemblies, power distribution and Na-void effect have been made. Efforts have been exerted to develop the required maintenance equipments, to manufacture the transfer rotor maintenance facilities, and to construct the spent fuel storing and cooling facilities. The analysis and calculation of the core characteristics have been in progress. The design work on the prototype fast breeder reactor ''Monju'' has been continued, and the development of the computer codes for the design has progressed. Informations have been gathered regarding the technological developments of LMFBRs overseas. The surveys on the site for ''Monju'' have been carried out. The design and research works on the demonstration reactor were started, and the general design factors such as the steam condition and the plant layout have been studied. As for the research and development of reactor physics, structural components, instrumentation and control, sodium technology, fuel materials, structural materials, safety, and steam generators, the progresses are reported in detail. High performance neutron detectors for nuclear instrumentation have been under development, and the tagging gas method for fuel failure detection and location system has been tested. (Kako, I.)

  2. IAEA Issues Report on Mission to Review Japan's Nuclear Power Plant Safety Assessment Process

    International Nuclear Information System (INIS)

    2012-01-01

    Full text: A team of international nuclear safety experts has delivered its report on a mission it conducted from 21-31 January 2012 to review Japan's process for assessing nuclear safety at the nation's nuclear power plants. International Atomic Energy Agency (IAEA) officials delivered the IAEA Mission Report to Japanese officials yesterday and made it publicly available today. Following the 11 March 2011 accident at TEPCO's Fukushima Daiichi Nuclear Power Station, Japan's Nuclear and Industrial Safety Agency (NISA) announced the development of a revised safety assessment process for the nation's nuclear power reactors. At the request of the Government of Japan, the IAEA organized a team of five IAEA and three international nuclear safety experts and visited Japan to review NISA's approach to the Comprehensive Assessments for the Safety of Existing Power Reactor Facilities and how NISA examines the results submitted by nuclear operators. A Preliminary Summary Report was issued on 31 January. 'The mission report provides additional information regarding the team's recommendations and overall finding that NISA's instructions to power plants and its review process for the Comprehensive Safety Assessments are generally consistent with IAEA Safety Standards', said team leader James Lyons, Director of the IAEA's Nuclear Installation Safety Division. National safety assessments and their peer review by the IAEA are a key component of the IAEA Action Plan on Nuclear Safety, which was approved by the Agency's Member States following last year's nuclear accident at Fukushima Daiichi Nuclear Power Station. The IAEA safety review mission held meetings in Tokyo with officials from NISA, the Japanese Nuclear Energy Safety Organization (JNES), and the Kansai Electric Power Company (KEPCO), and the team visited the Ohi Nuclear Power Station to see an example of how Japan's Comprehensive Safety Assessment is being implemented by nuclear operators. In its report delivered today

  3. Corrosion control in CANDU nuclear power reactors

    International Nuclear Information System (INIS)

    Lesurf, J.E.

    1974-01-01

    Corrosion control in CANDU reactors which use pressurized heavy water (PHW) and boiling light water (BLW) coolants is discussed. Discussions are included on pressure tubes, primary water chemistry, fuel sheath oxidation and hydriding, and crud transport. It is noted that corrosion has not been a significant problem in CANDU nuclear power reactors which is a tribute to design, material selection, and chemistry control. This is particularly notable at the Pickering Nuclear Generating Station which will have four CANDU-PHW reactors of 540 MWe each. The net capacity factor for Pickering-I from first full power (May 1971) to March 1972 was 79.5 percent, and for Pickering II (first full power November 1971) to March 1972 was 83.5 percent. Pickering III has just reached full power operation (May 1972) and Pickering IV is still under construction. Gentilly CANDU-BLW reached full power operation in May 1972 after extensive commissioning tests at lower power levels with no major corrosion or chemistry problems appearing. Experience and operating data confirm that the value of careful attention to all aspects of corrosion control and augur well for future CANDU reactors. (U.S.)

  4. Reactors of different types in the world nuclear power

    International Nuclear Information System (INIS)

    Simonov, K.V.

    1991-01-01

    The status of the world nuclear power is briefly reviewed. It is noted that PWR reactors have decisive significance in the world power. The second place is related to gas-cooled graphite-moderated reactors. Channel-type heavy water moderated reactors are relatively important. Nuclear power future is associated with fast liquid-metal cooled breeder reactors

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

    Science.gov (United States)

    Corradini, M L

    2007-11-01

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

  6. FY 2000 report on the demonstrative research for photovoltaic power generation system in Thailand. Demonstrative study on photovoltaic power generation grid-connected system

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-09-01

    In relation to the demonstrative study of the photovoltaic power system that is planned in Libong island, Thailand, the FY 2000 results were reported. In this R and D, construction/demonstrative operation were planned for a photovoltaic power station with a generation output of 100kW, photovoltaic power system in school facilities, and system for transmitting/distributing power to houses by connecting the power station and power system. In this fiscal year, the field survey was conducted together with the alteration from Yao Yai island, for which the demonstrative study was planned at first, to Libong island. The electric equipment was selected which met the requests from Thailand and the results of the field survey, and the basic design of the photovoltaic power generation/transmission/distribution system was completed. Based on this, the design/manufacture of photovoltaic power generation modules, power control equipment, measuring equipment, etc. were made. At the construction site of photovoltaic power station, construction work such as land formation was conducted. Further, Thai engineers who visited Japan did the following: discussions about power system, presence at test/inspection of photovoltaic power generation modules, visits to photovoltaic power stations, wind power stations, etc. (NEDO)

  7. Current Status of QA For Nuclear Power Plants in Japan

    International Nuclear Information System (INIS)

    Nagoshi, Hitohiko

    1986-01-01

    It is the current status of QA and our QA experiences with nuclear power plants against the background of the Japanese social and business environment. Accordingly, in 1972, 'The Guidance for Quality Assurance in Construction of Nuclear Power Plants' based on U. S. 10CEF50 Appendix B, was published by the Japan Electric Association. 'Jug-4101 The Guide for Quality Assurance of Nuclear Power Plants' has been prepared by referring to the IAEA QA code. The Guide has been accepted by the Japanese nuclear industry and applied to the QA programs of every organization concerned therewith. The Japanese approach to higher quality will naturally be different from that of other countries because of Japan's cultural, social, and economic conditions. Even higher quality is being aimed at through the LWR Improvement and Standardization Program and coordinated quality assurance efforts

  8. The status and future development of nuclear power in Japan

    International Nuclear Information System (INIS)

    Shoda, A.Y.

    1987-01-01

    As a result of its high dependence on imports in the energy supply sector, Japan has embarked on an extensive nuclear power program, which covers the whole nuclear fuel cycle and the construction of nuclear power plants. In 1985, 32 nuclear generating units with an aggregate power of 24.500 MW were in operation; this amounts to well over a quarter of the total generating capacity installed in Japan. Another ten units with an aggregate approx. 10,000 MW are under construction and six units with 6300 MW are being prepared for construction. After the completion of this phase of the program in 1995, the nuclear generating capacity is to be stepped up year by year by an aggregate 1500 to 2500 MW, for the time being. (orig.) [de

  9. Demonstration and information center on the basis of the research reactor IR-50

    International Nuclear Information System (INIS)

    Krupenina, F.

    2001-01-01

    Many problems exist in the nuclear field, but the most significant one is the public's mistrust of Nuclear Energy. Strong downfalls of the radiological culture affect public perception, the main paradox being the situation after Chernobyl. The task of creating a Demonstration and-Information Center (Minatom RF) on the basis of the research reactor IR-50 is conducted by Research and Development Institute of Power Engineering (ENTEK). The IR-50 is situated on the grounds of the institute. It will be a unique event when the functional reactor is situated in the center of the city (about 5 km from Kremlin). (author)

  10. Power reactor events, May-June 1986

    International Nuclear Information System (INIS)

    Massaro, S.A.

    1986-12-01

    Power Reactor Events is a bi-monthly newsletter that compiles operating experience information about commercial nuclear power plants. This includes summaries of noteworthy events and listings and/or abstracts of USNRC and other documents that discuss safety-related or possible generic issues. It is intended to feed back some of the lessons learned from operational experience to the various plant personnel, i.e., managers, licensed reactor operators, training coordinators, and support personnel. Events at the following plants are reported: McGuire Unit 1; Susquehanna Units 1 and 2; Browns Ferry Units 1, 2, and 3; and River Bend Unit 1

  11. ASTRID: Advanced Sodium Technological Reactor for Industrial Demonstration

    International Nuclear Information System (INIS)

    Vasile, A.

    2012-01-01

    Conclusions: • R&D results [CEA-AREVA-EDF] obtained from 2007 to 2009 have contributed to ASTRID mid 2010 choice of options; • ASTRID has the objective to demonstrate at the industrial scale progress in the identified domains of SFR weakness (safety, operability, economy). and to perform transmutation demonstrations; • A lot of improvements are related to safety; • The first very important milestone is 2012 (June 2006 French Act on wastes management): – ASTRID pre-conceptual design studies: 2010-2012; – First investment cost evaluation; – First safety Authorities advice on the orientations for ASTRID safety; • With the ASTRID program funded by the French government, France has the opportunity to develop a GEN IV Sodium Fast Reactor

  12. 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

  13. Reactor power system deployment and startup

    International Nuclear Information System (INIS)

    Wetch, J.R.; Nelin, C.J.; Britt, E.J.; Klein, G.; Rasor Associates, Inc., Sunnyvale, CA; California Institute of Technology, Pasadena)

    1985-01-01

    This paper addresses issues that should receive further examination in the near-term as concept selection for development of a U.S. space reactor power system is approached. The issues include: the economics, practicality and system reliability associated with transfer of nuclear spacecraft from low earth shuttle orbits to operational orbits, via chemical propulsion versus nuclear electric propulsion; possible astronaut supervised reactor and nuclear electric propulsion startup in low altitude Shuttle orbit; potential deployment methods for nuclear powered spacecraft from Shuttle; the general public safety of low altitude startup and nuclear safe and disposal orbits; the question of preferred reactor power level; and the question of frozen versus molten alkali metal coolant during launch and deployment. These issues must be considered now because they impact the SP-100 concept selection, power level selection, weight and size limits, use of deployable radiators, reliability requirements, and economics, as well as the degree of need for and the urgency of developing space reactor power systems. 5 references

  14. Environmental Remediation Activities in Japan Following the Fukushima Dai-ichi Reactor Incident - 12603

    Energy Technology Data Exchange (ETDEWEB)

    Lively, J.W.; Kelley, J.L.; Marcial, M.R. [AMEC Environment and Infrastructure (United States); Yashio, Shoko; Kuriu, Nobou; Kamijo, Hiroaki; Jotatsu, Kato [Obayashi Corporation (Japan)

    2012-07-01

    In March 2011, the Fukushima Dai-ichi reactor power plant was crippled by the Great Pacific earthquake and subsequent tsunami. Much of the focus in the news was on the reactor site itself as the utility company (TEPCO), the Japanese government, and experts from around the world worked to bring the damaged plants into a safe shutdown condition and stem the release of radioactivity to the environment. Most of the radioactivity released was carried out to sea with the prevailing winds. Still, as weather patterns changed and winds shifted, a significant plume of radioactive materials released from the plant deposited in the environment surrounding the plant, contaminating large land areas of the Fukushima Prefecture. The magnitude of the radiological impact to the surrounding environmental is so large that the Japanese government has had to reevaluate the meaning of 'acceptably clean'. In many respects, 'acceptably clean' cannot be a one-size-fits-all standard. The economics costs of such an approach would make impossible what is already an enormous and costly environmental response and remediation task. Thus, the Japanese government has embarked upon an approach that is both situation-specific and reasonably achievable. For example, the determination of acceptably clean for a nursery school or kindergarten play yard may be different from that for a parking lot. The acceptably clean level of residual radioactivity in the surface soil of a rice paddy is different from that in a forested area. The recognized exposure situation (scenario) thus plays a large role in the decision process. While sometimes complicated to grasp or implement, such an approach does prioritize national resources to address environment remediation based upon immediate and significant risks. In addition, the Japanese government is testing means and methods, including advanced or promising technologies, that could be proven to be effective in reducing the amount of radioactivity

  15. Environmental and health consequences in Japan due to the accident at Chernobyl nuclear reactor plant

    International Nuclear Information System (INIS)

    Uchiyama, Masafumi; Nakamura, Yuji; Kankura, Takako; Iwasaki, Tamiko; Fujimoto, Kenzo; Kobayashi, Sadayoshi.

    1988-03-01

    A comprehensive review was made on the results of national monitoring program for environmental radioactivity in Japan resulting from the accident at the Chernobyl nuclear power plant in USSR. Period of monitoring efforts covered by the present review is from 30th of April 1986 to 31st of May 1987. A radioactive cloud released from the Chernobyl nuclear reactor initially arrived in Japan on 30th of April 1986 as indicated by the elevated level of 131 I, 137 Cs and 134 Cs activity in the total deposition on 30th of April and also by the increased 137 Cs body burden noted on 1st of May. Almost all the radioactive nuclides detected in the European countries were also identified in Japan. For example, the observed nuclides were: 95 Zr, 95 Nb, 99m Tc, 103 Ru, 106 Ru, 110m Ag, 111 Ag, 125 Sb, 127 Sb, 129m Te, 131 I, 132 Te, 132 I, 133 I, 134 Cs, 136 Cs, 137 Cs, 140 Ba, 140 La, 141 Ce and 144 Ce. Among the above radionuclides, the country average concentration was determined for 131 I, 137 Cs and 134 Cs in various environmental materials such as air, fresh water, soil, milk, leafy and root vegetables, cereals, marine products and other foodstuffs. In contrast to the sharp decline of 131 I which was negligible after a few months, 137 Cs showed a tendency to maintain its activity in foodstuffs at an appreciable level one year later. Collective effective dose equivalent and dose equivalent to thyroid in Japanese population due to 137 Cs, 134 Cs and 131 I were estimated to be around 590 man Sv and 4760 man Sv, respectively. Corresponding values for the per caput dose equivalent are 5 μSv for whole body and 40 μSv for thyroid, respectively. (author)

  16. Systems aspects of a space nuclear reactor power system

    Science.gov (United States)

    Jaffe, L.; Fujita, T.; Beatty, R.; Bhandari, P.; Chow, E.; Deininger, W.; Ewell, R.; Grossman, M.; Bloomfield, H.; Heller, J.

    1988-01-01

    Various system aspects of a 300-kW nuclear reactor power system for spacecraft have been investigated. Special attention is given to the cases of a reusable OTV and a space-based radar. It is demonstrated that the stowed length of the power system is important to mission design, and that orbital storage for months to years may be needed for missions involving orbital assembly.

  17. Power Reactor Thoria Reprocessing Facility (PRTRF), Trombay

    International Nuclear Information System (INIS)

    Dhami, P.S; Yadav, J.S; Agarwal, K.

    2017-01-01

    Exploitation of the abundant thorium resources to meet sustained energy demand forms the basis of the Indian nuclear energy programme. To gain reprocessing experience in thorium fuel cycle, thoria was irradiated in research reactor CIRUS in early sixties. Later in eighties, thoria bundles were used for initial flux flattening in some of the pressurized heavy water reactors (PHWRs). The research reactor irradiated thoria contained small content (∼ 2-3ppm) of "2"3"2U in "2"3"3U product, which did not pose any significant radiological problems during processing in Uranium Thorium Separation Facility (UTSF), Trombay. Thoria irradiated in PHWRs on discharge contained (∼ 0.5-1.5% "2"3"3U with significant "2"3"2U content (100-500 ppm) requiring special radiological attention. Based on the experience from UTSF, a new facility viz. Power Reactor Thoria Reprocessing Facility (PRTRF), Trombay was built which was hot commissioned in the year 2015

  18. Power control device for nuclear reactors

    International Nuclear Information System (INIS)

    Kagawa, Tatsuo

    1984-01-01

    Purpose: To eliminate for requirement of control rods and movable portions, as well as ensure the safety and reliability of the operation. Constitution: A plurality of control tubes are disposed within a reactor core instead of control rods. Tubes are connected from below the reactor core to the control tubes for supplying liquid poisons such as aqueous boric acid to the inside of the control tubes. Further, tubes are connected to the upper portion of the control tubes for guiding the liquid poisons from the reactor core to the outside. The tubes for supplying and discharging the liquid poisons are introduced externally through the flange disposed at the upper portion of a pressure vessel. At the outside of the pressure vessel, are disposed a liquid poison tank, a pressurizing source, a pressure control valve, a liquid level meter and the like. The control for the reactor power is conducted by controlling the level of the liquid poisons in the control tubes. (Ikeda, J.)

  19. Review of the activities in Japan

    International Nuclear Information System (INIS)

    Otake, I.

    1982-01-01

    The fast breeder reactor development project in Japan has been in progress through.operation of the experimental fast reactor JOYO, design of the prototype fast breeder reactor MONJU and related R and D works. JOYO began operation in mid-1977, increased power from 50 MWt to 75 MWth in July 1979 and operation cycles at 75 MWth are continued at present. With respect to MONJU, which is a 300 MWe plant, progress toward construction has been made and the safety review are started by the concerned authorities. Conceptual design studies of large demonstration fast breeder reactor are also being made by PNC. It is a 1000 MWe, loop type plant

  20. Radiological consequence of Chernobyl nuclear power accident in Japan

    International Nuclear Information System (INIS)

    Uchiyama, Masafumi; Nakamura, Yuji; Kankura, Takako; Iwasaki, Tamiko; Fujimoto, Kenzo; Kobayashi, Sadayoshi.

    1988-03-01

    Two years have elapsed since the accident in Chernobyl nuclear power station shocked those concerned with nuclear power generation. The effect that this accident exerted on human environment has still continued directly and indirectly, and the reports on the effect have been made in various countries and by international organizations. In Japan, about the exposure dose of Japanese people due to this accident, the Nuclear Safety Commission and Japan Atomic Energy Research Institute issued the reports. In this report, the available data concerning the envrionmental radioactivity level in Japan due to the Chernobyl accident are collected, and the evaluation of exposure dose which seems most appropriate from the present day scientific viewpoint was attempted by the detailed analysis in the National Institute of Radiological Sciences. The enormous number of the data observed in various parts of Japan were different in sampling, locality, time and measuring method, so difficulty arose frequently. The maximum concentration of I-131 in floating dust was 2.5 Bq/m 3 observed in Fukui, and the same kinds of radioactive nuclides as those in Europe were detected. (Kako, I.)

  1. Public perception of risks from nuclear power plants in Japan, before the Great East Japan Earthquake

    International Nuclear Information System (INIS)

    Murayama, Rumiko; Nakaune, Naoko; Kishikawa, Hiroki; Uchiyama, Iwao

    2011-01-01

    On this research we aim to clarify public perception of risks from nuclear power plants before the Great East Japan Earthquake. The findings of the questionnaire survey conducted in 2010 showed that 1) about 70% of the people felt that they gained benefit from nuclear power plants and these were needed for their daily life. 2) Fifty percent respondents recognized there was danger to themselves and their family members with regards to nuclear power plants. The risks of nuclear power plants to Japanese society ware estimated higher than that risk to individuals of Japanese public. 3) Perception of risks from nuclear power plants to individual Japanese tended to be slightly lower between 1999 and 2010. (author)

  2. Radiation streaming in power reactors. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Lahti, G.P.; Lee, R.R.; Courtney, J.C. (eds.)

    1979-02-01

    Separate abstracts are included for each of the 14 papers given at a special session on Radiation Streaming in Power Reactors held on November 15 at the American Nuclear Society 1978 Winter Meeting in Washington, D.C. The papers describe the methods of calculation, the engineering of shields, and the measurement of radiation environments within the containments of light water power reactors. Comparisons of measured and calculated data are used to determine the accuracy of computer predictions of the radiation environment. Specific computational and measurement techniques are described and evaluated. Emphasis is on radiation streaming in the annular region between the reactor vesel and the primary shield and its resultant environment within the primary containment.

  3. Advances in neutronics calculation of fast neutron reactors - Demonstration on Super-Phenix reactor

    International Nuclear Information System (INIS)

    Czernecki, Sebastien

    1998-01-01

    The fast reactor european neutronics calculations system, ERANOS, has integrated recent improvements both in nuclear data, with the use of the adjusted nuclear library ERALIB 1 from the JEF2.2 library, and calculation methods, with the use of the new european cell code, ECCO, and the deterministic code, TGV/VARIANT. This code performs full 3-D reactor calculation in the transport theory with variational method. The aim of this work is to create and validate a new calculational scheme for fast spectrum systems offering good compromise between accuracy and running time. The new scheme is based on these improvements plus a special procedure accounting for control rod heterogeneity, which uses a reactivity equivalence homogenization. The new scheme has been validated by means of experiment/calculation comparisons, using the extensive start-up program measurements performed in Super-Phenix reactor. The validation uses also recent measurements performed in the Phenix reactor. The results are very satisfactory and show a significant improvement for almost all core parameters, especially for critical mass, control rod worth and radial subassembly power distribution. A detailed analysis of the discrepancies between the old scheme and the new one for this parameter allows to understand the separate effects of methods and nuclear data on the radial power distribution shape. (author) [fr

  4. High Flux Isotope Reactor power upgrade status

    International Nuclear Information System (INIS)

    Rothrock, R.B.; Hale, R.E.; Cheverton, R.D.

    1997-01-01

    A return to 100-MW operation is being planned for the High Flux Isotope Reactor (HFIR). Recent improvements in fuel element manufacturing procedures and inspection equipment will be exploited to reduce hot spot and hot streak factors sufficiently to permit the power upgrade without an increase in primary coolant pressure. Fresh fuel elements already fabricated for future use are being evaluated individually for power upgrade potential based on their measured coolant channel dimensions

  5. The RERTR demonstration experiments program at the Ford Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wehe, D K; King, J S [Department of Nuclear Engineering, University of Michigan (United States)

    1983-08-01

    The purpose of this paper is to highlight a major part of the experimental work which is being carried out at the Ford Nuclear Reactor (FNR) in conjunction with the RERTR program. A demonstration experiments program has been developed to: 1) characterize the FNR in sufficient detail to discern and quantify neutronic differences between the high and low enriched cores; 2) provide the theoretical group with measurements to benchmark their calculations. As with any experimental program associated with a reactor, stringent constraints limit the experiments which can be performed. Some experiments are performed routinely on the FNR (such as control rod calibrations), and much data is already available. Unfortunately, the accuracy we demand precludes using much of this earlier data. And in many cases, the requirement of precise (and copious) data has led to either developing new techniques (as in the case of rhodium mapping and neutron diffraction) or to further refinements on existing methods (as in the case of spectral unfolding). Nevertheless, we have tried to stay within the realm of recognized, well-established experimental methods in order to assuage any doubts about measured differences between HEU and LEU core parameters. This paper describes the principal results of the experiments performed so far.

  6. Experience on the demonstration of safety for older reactors

    International Nuclear Information System (INIS)

    Facer, R.

    2001-01-01

    The UK's oldest reactors are still operating. Built during the 1950's and commissioned between 1956 and 1960, eight reactors continue to provide electricity and process steam. It is still economically justified to keep them running. In addition to the economic considerations it is also necessary to justify that they can still continue to operate safely. This paper provides a brief review of how the Operator of these stations has justified the safety of operation to date and how they expect to continue to justify their operation for several more years. It is appropriate to consider why the Operator wishes to keep the plant operating. Among the most important reasons are that: The plant is built and paid for, Running costs are relatively low process steam is available for the adjacent sites It is a commercially viable electricity producer It is a reliable electricity source The operators have developed programmes for safety review of the plant and introduced a Continuing Operation Programme which had two main requirements which were, the demonstration of continuing acceptable safety the ensurance of commercial viability. (author)

  7. State system experience with safeguarding power reactors

    International Nuclear Information System (INIS)

    Roehnsch, W.

    1982-01-01

    This session describes the development and operation of the State System of Accountancy and Control in the German Democratic Republic, and summarizes operating experience with safeguards at power reactor facilities. Overall organization and responsibilities, containment and surveillance measures, materials accounting, and inspection procedures will be outlined. Cooperation between the IAEA, State system, facility, and supplier authorities will also be addressed

  8. The program of reactors and nuclear power plants

    International Nuclear Information System (INIS)

    Calabrese, Carlos R.

    2001-01-01

    Into de framework of the program of research reactors and nuclear power plants, the operating Argentine reactors are described. The uses of the research reactors in Argentina are summarized. The reactors installed by Argentina in other countries (Peru, Algeria, Egypt) are briefly described. The CAREM project for the design and construction of an innovator small power reactor (27 MWe) is also described in some detail. The next biennial research and development program for reactor is briefly outlined

  9. Data list of nuclear power plants in Japan

    International Nuclear Information System (INIS)

    Izumi, Fumio; Nakamura, Jinichi

    1982-10-01

    This report has collected and compiled the data by December in 1981 concerning performances, equipments and installations of the nuclear power plants in Japan. The data have been modified according to the changes produced after previous publication of 1979 edition including BWR and PWR (JAERI-M 8947) and 1980 edition including PWR (JAERI-M 9629), and extended to cover the new plants developed thereafter. All data have been processed and tabulated with a data processing computer program FREP. Besides this report, user also can refer to 'Data List of Nuclear Power Plant in Japan' through terminals equipped at various places in JAERI using TSS (Time Shearing System) network of FACOM M-200, and the explanation of the usage is given in the Appendix. (author)

  10. Correlations between power and test reactor data bases

    International Nuclear Information System (INIS)

    Guthrie, G.L.; Simonen, E.P.

    1989-02-01

    Differences between power reactor and test reactor data bases have been evaluated. Charpy shift data has been assembled from specimens irradiated in both high-flux test reactors and low-flux power reactors. Preliminary tests for the existence of a bias between test and power reactor data bases indicate a possible bias between the weld data bases. The bias is nonconservative for power predictive purposes, using test reactor data. The lesser shift for test reactor data compared to power reactor data is interpreted primarily in terms of greater point defect recombination for test reactor fluxes compared to power reactor fluxes. The possibility of greater thermal aging effects during lower damage rates is also discussed. 15 refs., 5 figs., 2 tabs

  11. Monitoring device for the reactor power distribution

    International Nuclear Information System (INIS)

    Uematsu, Hitoshi; Tsuiki, Makoto

    1982-01-01

    Purpose: To enable accurate monitoring for the power distribution in a short time, as well as independent detection for in-core neutron flux detectors in abnormal operation due to failures or like other causes to thereby surely provide reliable substitute values. Constitution: Counted values are inputted from a reactor core present status data detector by a power distribution calculation device to calculate the in-core neutron flux density and the power distribution based on previously stored physical models. While on the other hand, counted value from the in-core neutron detectors and the neutron flux distribution and the power distribution calculated from the power distribution calculation device are inputted from a BCF calculation device to compensate the counting errors incorporated in the counted value from the in-core neutron flux detectors and the calculation errors incorporated in the power distribution calculated in the power distribution calculation device respectively and thereby calculate the power distribution in the reactor core. Further, necessary data are inputted to the power distribution calculation device by an input/output device and the results calculated in the BCF calculation device are displayed. (Aizawa, K.)

  12. Long-term scenarios of power reactors and fuel cycle development and the role of reduced moderation water reactors

    International Nuclear Information System (INIS)

    Sato, Osamu; Tatematsu, Kenji; Tanaka, Yoji

    2000-01-01

    Reduced moderation spectrum reactor is one of water cooled type reactors in future, which is based on the advanced technology of conventional nuclear power plants. The reduced moderation water reactor (RMWR) has various advantages, such as effective utilization of uranium resources, high conversion ratio, high burn-up, long-term cycle operation, and multiple recycle of plutonium. The RMWR is expected to be a substitute of fast breeder reactor (FBR) of which the development encounters with some technical and financial difficulties, and discontinues in many countries. The role of the RMWR on long-term scenarios of power reactor and fuel cycle development in Japan is investigated from the point of view of uranium resource needed. The consumption of natural uranium needed up to the year 2200 is calculated on various assumptions for the following three cases: (1) no breeder reactor; plutonium-thermal cycle in conventional light water reactor, (2) introduction of the FBR, and (3) introduction of the RMWR. The amounts of natural uranium consumption depends largely on the conversion ratio and plutonium quantity needed of a reactor type. The RMWR has a possibility as a substitute technology of the FBR with the improvement of conversion ratio and high burn-up. (Suetake, M.)

  13. Long-term scenarios of power reactors and fuel cycle development and the role of reduced moderation water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Sato, Osamu; Tatematsu, Kenji; Tanaka, Yoji [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2000-06-01

    Reduced moderation spectrum reactor is one of water cooled type reactors in future, which is based on the advanced technology of conventional nuclear power plants. The reduced moderation water reactor (RMWR) has various advantages, such as effective utilization of uranium resources, high conversion ratio, high burn-up, long-term cycle operation, and multiple recycle of plutonium. The RMWR is expected to be a substitute of fast breeder reactor (FBR) of which the development encounters with some technical and financial difficulties, and discontinues in many countries. The role of the RMWR on long-term scenarios of power reactor and fuel cycle development in Japan is investigated from the point of view of uranium resource needed. The consumption of natural uranium needed up to the year 2200 is calculated on various assumptions for the following three cases: (1) no breeder reactor; plutonium-thermal cycle in conventional light water reactor, (2) introduction of the FBR, and (3) introduction of the RMWR. The amounts of natural uranium consumption depends largely on the conversion ratio and plutonium quantity needed of a reactor type. The RMWR has a possibility as a substitute technology of the FBR with the improvement of conversion ratio and high burn-up. (Suetake, M.)

  14. Development of remote decontamination technologies improving internal environment of reactor buildings at Fukushima Daiichi Nuclear Power Station

    International Nuclear Information System (INIS)

    Hotta, Koji; Hayashi, Hirotada; Sakai, Hitoshi

    2016-01-01

    The reactor buildings at the Fukushima Daiichi Nuclear Power Station of Tokyo Electric Power Co., Inc., which was seriously damaged by the Great East Japan Earthquake of March 11, 2011, have been highly contaminated by radioactive materials. To safely and efficiently advance the processes related to the forthcoming decommissioning of the reactors, it is necessary to improve the hazardous environment inside the reactor buildings. During the more than four years that have elapsed since the Great East Japan Earthquake, Toshiba has been implementing various measures to reduce the ambient dose rates inside the reactor buildings through decontamination work and participation in a national project for the development of remote decontamination technologies for reactor buildings. A variety of vehicles and technologies to support decontamination work have been developed through these activities, and are significantly contributing to improvement of the environment inside the reactor buildings. (author)

  15. Assessment of the thorium fuel cycle in power reactors

    International Nuclear Information System (INIS)

    Kasten, P.R.; Homan, F.J.; Allen, E.J.

    1977-01-01

    A study was conducted at Oak Ridge National Laboratory to evaluate the role of thorium fuel cycles in power reactors. Three thermal reactor systems were considered: Light Water Reactors (LWRs); High-Temperature Gas-Cooled Reactors (HTGRs); and Heavy Water Reactors (HWRs) of the Canadian Deuterium Uranium Reactor (CANDU) type; most of the effort was on these systems. A summary comparing thorium and uranium fuel cycles in Fast Breeder Reactors (FBRs) was also compiled

  16. Self-powered detectors for power reactors: an overview

    International Nuclear Information System (INIS)

    Ma, J.

    2006-01-01

    In this paper, Self-Powered Detectors (SPDs) for applications in nuclear power reactors have been reviewed. Based on their responses to radiation, these detectors can be divided into delayed response Self-Powered Neutron Detector (SPND), prompt response SPND and Self-Powered Gamma Detector (SPGD). The operational principles of these detectors are presented and their distinctive characteristics are examined accordingly. The analytical models and Monte Carlo method to calculate the responses of these detectors to neutron flux and external gamma rays are reviewed. The paper has also considered some related signal processing techniques, such as detector calibrations and detector signal compensations. Furthermore, a couple of failure modes have also been analyzed. Finally, applications of SPD in nuclear power reactors are summarized. (author)

  17. Self-powered detectors for power reactors: an overview

    Energy Technology Data Exchange (ETDEWEB)

    Ma, J. [Univ. of Western Ontario, Dept. of Mechanical and Materials Engineering, London, Ontario (Canada)]. E-mail: jma64@uwo.ca

    2006-07-01

    In this paper, Self-Powered Detectors (SPDs) for applications in nuclear power reactors have been reviewed. Based on their responses to radiation, these detectors can be divided into delayed response Self-Powered Neutron Detector (SPND), prompt response SPND and Self-Powered Gamma Detector (SPGD). The operational principles of these detectors are presented and their distinctive characteristics are examined accordingly. The analytical models and Monte Carlo method to calculate the responses of these detectors to neutron flux and external gamma rays are reviewed. The paper has also considered some related signal processing techniques, such as detector calibrations and detector signal compensations. Furthermore, a couple of failure modes have also been analyzed. Finally, applications of SPD in nuclear power reactors are summarized. (author)

  18. Transients in reactors for power systems compensation

    Science.gov (United States)

    Abdul Hamid, Haziah

    This thesis describes new models and investigations into switching transient phenomena related to the shunt reactors and the Mechanically Switched Capacitor with Damping Network (MSCDN) operations used for reactive power control in the transmission system. Shunt reactors and MSCDN are similar in that they have reactors. A shunt reactor is connected parallel to the compensated lines to absorb the leading current, whereas the MSCDN is a version of a capacitor bank designed as a C-type filter for use in the harmonic-rich environment. In this work, models have been developed and transient overvoltages due to shunt reactor deenergisation were estimated analytically using MathCad, a mathematical program. Computer simulations used the ATP/EMTP program to reproduce both single-phase and three-phase shunt reactor switching at 275 kV operational substations. The effect of the reactor switching on the circuit breaker grading capacitor was also examined by considering various switching conditions.. The main original achievement of this thesis is the clarification of failure mechanisms occurring in the air-core filter reactor due to MSCDN switching operations. The simulation of the MSCDN energisation was conducted using the ATP/EMTP program in the presence of surge arresters. The outcome of this simulation shows that extremely fast transients were established across the air-core filter reactor. This identified transient event has led to the development of a detailed air-core reactor model, which accounts for the inter-turn RLC parameters as well as the stray capacitances-to-ground. These parameters are incorporated into the transient simulation circuit, from which the current and voltage distribution across the winding were derived using electric field and equivalent circuit modelling. Analysis of the results has revealed that there are substantial dielectric stresses imposed on the winding insulation that can be attributed to a combination of three factors. (i) First, the

  19. The Ford Nuclear Reactor demonstration project for the evaluation and analysis of low enrichment fuel

    International Nuclear Information System (INIS)

    Kerr, W.; King, J.S.; Lee, J.C.; Martin, W.R.; Wehe, D.K.

    1991-07-01

    The whole-core LEU fuel demonstration project at the University of Michigan was begun in 1979 as part of the Reduced Enrichment Research and Test Reactor (RERTR) Program at Argonne National Laboratory. An LEU fuel design was selected which would produce minimum perturbations in the neutronic, operations, and safety characteristics of the 2-MW Ford Nuclear Reactor (FNR). Initial criticality with a full LEU core on December 8, 1981, was followed by low- and full-power testing of the fresh LEU core, transitional operation with mixed HEU-LEU configurations, and establishment of full LEU equilibrium core operation. The transition from the HEU to the LEU configurations was achieved with negligible impact on experimental utilization and safe operation of the reactor. 78 refs., 74 figs., 84 tabs

  20. Long-term ETR/INTOR magnet testing in support of the demonstration fusion reactor

    International Nuclear Information System (INIS)

    Herring, J.S.; Shah, V.N.; Rouhani, S.Z.

    1983-01-01

    This study considers ways that the proposed Engineering Test Reactor (ETR), or the proposed International Tokamak Reactor (INTOR), can be used for magnet performance tests that would be useful for the design and operation of the Demonstration Tokamak Power Plant (DEMO). Such testing must not interfere with the main function of the ETR/INTOR as an integrated fusion reactor. A performance test plan for the ETR/INTOR magnets is proposed and appropriate tests on the magnets is proposed and appropriate tests on the magnets for each phase of the ETR/INTOR operation are described. The suggested tests would verify design requirements and monitor long-term changes due to radiation. This paper also summarizes the design and operational performance of existing superconducting magnets and identifies the known failures and their predominant causes

  1. Fission Surface Power Technology Demonstration Unit Test Results

    Science.gov (United States)

    Briggs, Maxwell H.; Gibson, Marc A.; Geng, Steven M.; Sanzi, James L.

    2016-01-01

    The Fission Surface Power (FSP) Technology Demonstration Unit (TDU) is a system-level demonstration of fission power technology intended for use on manned missions to Mars. The Baseline FSP systems consists of a 190 kWt UO2 fast-spectrum reactor cooled by a primary pumped liquid metal loop. This liquid metal loop transfers heat to two intermediate liquid metal loops designed to isolate fission products in the primary loop from the balance of plant. The intermediate liquid metal loops transfer heat to four Stirling Power Conversion Units (PCU), each of which produce 12 kWe (48 kW total) and reject waste heat to two pumped water loops, which transfer the waste heat to titanium-water heat pipe radiators. The FSP TDU simulates a single leg of the baseline FSP system using an electrically heater core simulator, a single liquid metal loop, a single PCU, and a pumped water loop which rejects the waste heat to a Facility Cooling System (FCS). When operated at the nominal operating conditions (modified for low liquid metal flow) during TDU testing the PCU produced 8.9 kW of power at an efficiency of 21.7 percent resulting in a net system power of 8.1 kW and a system level efficiency of 17.2 percent. The reduction in PCU power from levels seen during electrically heated testing is the result of insufficient heat transfer from the NaK heater head to the Stirling acceptor, which could not be tested at Sunpower prior to delivery to the NASA Glenn Research Center (GRC). The maximum PCU power of 10.4 kW was achieved at the maximum liquid metal temperature of 875 K, minimum water temperature of 350 K, 1.1 kg/s liquid metal flow, 0.39 kg/s water flow, and 15.0 mm amplitude at an efficiency of 23.3 percent. This resulted in a system net power of 9.7 kW and a system efficiency of 18.7 percent.

  2. The European Fusion Energy Research Programme towards the realization of a fusion demonstration reactor

    International Nuclear Information System (INIS)

    Gasparotto, M.; Laesser, R.

    2006-01-01

    Since its inception, the European Fusion Programme has been orientated towards the establishment of the knowledge base needed for the definition of a reactor to be used for power production. Its ultimate goal is then to demonstrate the scientific and the technological feasibility of fusion power while incorporating the assessment of the safety, environmental, social and economic features of this type of energy source. At present, the JET device, the largest tokamak in the world, and the other medium-sized experimental machines are contributing essentially to the basic scientific phase of this development path. Their successful operation greatly contributed to support the design basis of ITER, the next step in fusion, which will aim to demonstrate the scientific and technical feasibility of fusion power production by achieving extended D-T burning plasma operation. Following ITER, the conception and construction of the DEMO device is planned. DEMO will be a demonstration power plant which will be the first fusion device to generate a significant amount of electrical power from fusion. This paper describes the status of fusion research and the European strategy for achievement of the ultimate goal of construction of a prototype reactor. (author)

  3. Outline of examination guides of water-cooled research reactors in Japan

    International Nuclear Information System (INIS)

    Yoshino, F.; Kimura, R.

    1992-01-01

    The Nuclear Safety Commission of Japan published two examination guides of water-cooled research reactors on July 18, 1991; one is for safety design, and another is for safety evaluation. In these guides, careful consideration is taken into account on the basic safety characteristic features of research reactors in order to be reasonable regulative requirements. This paper describes the fundamental philosophy and outline of the guides. (author)

  4. Method of estimating the reactor power distribution

    International Nuclear Information System (INIS)

    Mitsuta, Toru; Fukuzaki, Takaharu; Doi, Kazuyori; Kiguchi, Takashi.

    1984-01-01

    Purpose: To improve the calculation accuracy for the power distribution thereby improve the reliability of power distribution monitor. Constitution: In detector containing strings disposed within a reactor core, movable type neutron flux monitors are provided in addition to position fixed type neutron monitors conventionally disposed so far. Upon periodical monitoring, a power distribution X1 is calculated from a physical reactor core model. Then, a higher power position X2 is detected by position detectors and value X2 is sent to a neutron flux monitor driving device to displace the movable type monitors to a higher power position in each of the strings. After displacement, the value X1 is amended by an amending device using measured values from the movable type and fixed type monitors and the amended value is sent to a reactor core monitor device. Upon failure of the fixed type monitors, the position is sent to the monitor driving device and the movable monitors are displaced to that position for measurement. (Sekiya, K.)

  5. Reactor power distribution pattern judging device

    International Nuclear Information System (INIS)

    Ikehara, Tadashi.

    1992-01-01

    The judging device of the present invention comprises a power distribution readout system for intaking a power value from a fuel segment, a neural network having an experience learning function for receiving a power distribution value as an input variant, mapping it into a desirable property and self-organizing the map, and a learning date base storing a plurality of learnt samples. The read power distribution is classified depending on the similarity thereof with any one of representative learnt power distribution, and the corresponding state of the reactor core is outputted as a result of the judgement. When an error is found in the classified judging operation, erroneous cases are additionally learnt by using the experience and learning function, thereby improving the accuracy of the reactor core characteristic estimation operation. Since the device is mainly based on the neural network having a self-learning function and a pattern classification and judging function, a judging device having a human's intuitive pattern recognition performance and a pattern experience and learning performance is obtainable, thereby enabling to judge the state of the reactor core accurately. (N.H.)

  6. Experimental development of power reactor intelligent control

    International Nuclear Information System (INIS)

    Edwards, R.M.; Garcia, H.E.; Lee, K.Y.

    1992-01-01

    The US nuclear utility industry initiated an ambitious program to modernize the control systems at a minimum of ten existing nuclear power plants by the year 2000. That program addresses urgent needs to replace obsolete instrumentation and analog controls with highly reliable state-of-the-art computer-based digital systems. Large increases in functionality that could theoretically be achieved in a distributed digital control system are not an initial priority in the industry program but could be logically considered in later phases. This paper discusses the initial development of an experimental sequence for developing, testing, and verifying intelligent fault-accommodating control for commercial nuclear power plant application. The sequence includes an ultra-safe university research reactor (TRIGA) and a passively safe experimental power plant (Experimental Breeder Reactor 2)

  7. Nonlinear analysis on power reactor dynamics

    International Nuclear Information System (INIS)

    Konno, H.; Hayashi, K.

    1997-01-01

    We have shown that the origin of intermittent oscillation observed in a BWR can be ascribed to the couplings among the spatial modes starting from a non-linear center manifold equation with a delay-time and a spatial diffusion. We can reduce the problem to the stochastic coupled van der Pol oscillators with non-linear coupling term. This non-linear coupling term plays an important role to break the symmetry of the system and the non-linear damping of the system. The phenomenological generalization of van der Pol oscillator coupled by the linear diffusion term is not appropriate for describing the nuclear power reactors. However, one must start from the coupled partial differential equations by taking into account the two energy group neutrons, the thermo-hydraulic equations including two-phase flow. In this case, the diffusion constant must be a complex number as is demonstrated in a previous paper. The results will be reported in the near future. (J.P.N.)

  8. The program of reactors and nuclear power plants; Programa de reactores y centrales nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Calabrese, Carlos R [Comision Nacional de Energia Atomica, General San Martin (Argentina). Centro Atomico Constituyentes

    2001-07-01

    Into de framework of the program of research reactors and nuclear power plants, the operating Argentine reactors are described. The uses of the research reactors in Argentina are summarized. The reactors installed by Argentina in other countries (Peru, Algeria, Egypt) are briefly described. The CAREM project for the design and construction of an innovator small power reactor (27 MWe) is also described in some detail. The next biennial research and development program for reactor is briefly outlined.

  9. Tandem mirror reactor power balance studies

    International Nuclear Information System (INIS)

    Gorker, G.E.; Perkins, L.J.

    1985-01-01

    A tandem mirror reactor (TMR) power plant balance model has been developed and is now being used as a computer aid for performing parametric studies. End-cell power injection into the plasma and the physics thermal Q are used to determine the fusion power. About 80% of the fusion power is transferred by high-energy neutrons to the blanket modules and structures. The other 20% of the fusion power in the high-energy alpha particles is used to heat the deuterium-tritium (D-T) plasma. Most of the plasma-ionized particles transfer their energy to the halo dumps and direct converters. The plant efficiency is calculated for three different system cycles: (1) the pressurized water/saturated steam cycle; (2) the superheated steam cycle; and (3) the more complex superheat/reheat cycle. There is a signficiant improvement in plant efficiency as the electrical power multiplication factor and steam cycle efficiency increases

  10. TRIGA research reactors with higher power density

    International Nuclear Information System (INIS)

    Whittemore, W.L.

    1994-01-01

    The recent trend in new or upgraded research reactors is to higher power densities (hence higher neutron flux levels) but not necessarily to higher power levels. The TRIGA LEU fuel with burnable poison is available in small diameter fuel rods capable of high power per rod (≅48 kW/rod) with acceptable peak fuel temperatures. The performance of a 10-MW research reactor with a compact core of hexagonal TRIGA fuel clusters has been calculated in detail. With its light water coolant, beryllium and D 2 O reflector regions, this reactor can provide in-core experiments with thermal fluxes in excess of 3 x 10 14 n/cm 2 ·s and fast fluxes (>0.1 MeV) of 2 x 10 14 n/cm 2 ·s. The core centerline thermal neutron flux in the D 2 O reflector is about 2 x 10 14 n/cm 2 ·s and the average core power density is about 230 kW/liter. Using other TRIGA fuel developed for 25-MW test reactors but arranged in hexagonal arrays, power densities in excess of 300 kW/liter are readily available. A core with TRIGA fuel operating at 15-MW and generating such a power density is capable of producing thermal neutron fluxes in a D 2 O reflector of 3 x 10 14 n/cm 2 ·s. A beryllium-filled central region of the core can further enhance the core leakage and hence the neutron flux in the reflector. (author)

  11. In core system mapping reactor power distribution

    International Nuclear Information System (INIS)

    Yoriyaz, H.; Moreira, J.M.L.

    1989-01-01

    Based on the signals of SPND'S (Self Powered Neutron Detectors) distributed inside of a core, the spatial power distribution is obtained using the MAP program, developed in this work. The methodology applied in MAP program uses a least mean square technique to calculate expansion coefficients that depend on the SPND'S signals. The final power or neutron flux distribution is obtained by a combination of certains functions or expansion modes that are provided from diffusion calculation with the CITATION code. The MAP program is written in PASCAL language and will be used in IEA-R1 reactor for assisting its operation. (author) [pt

  12. Low power modular power generating reactors or Small Modular Reactors (SMR)

    International Nuclear Information System (INIS)

    Chenais, Jacques

    2016-01-01

    Electronuclear reactors were small reactors at the beginning, and then tend to be always bigger and more powerful, but since some recent times, several countries specialized in reactor design and fabrication (USA, Russia, China, and South Korea) have been developing Small Modular Reactors (SMR) of less than 300 MW. As France has already produced feasibility studies and is about to launch a SMR development programme, the author comments some specific aspects of this new architecture of reactors, characterises the targeted markets, gives an overview of the various more or less advanced existing concepts: a floating barge in Russia, the SMART 100 MW project in South Korea, several concepts in the USA (the mPower 125 MW, the NuScale 45 MW, the Westinghouse 225 MW, and the HI-SMUR 160 MW projects), the ACP 100 MW in China, the CAREM 27 MW in Argentina. French projects developed by the CEA, EDF, Areva and DCNS are then presented

  13. Basic training of nuclear power reactor personnel

    International Nuclear Information System (INIS)

    Palabrica, R.J.

    1981-01-01

    The basic training of nuclear power reactor personnel should be given very close attention since it constitutes the foundation of their knowledge of nuclear technology. Emphasis should be given on the thorough understanding of basic nuclear concepts in order to have reasonable assurance of successful assimilation by those personnel of more specialized and advanced concepts to which they will be later exposed. Basic training will also provide a means for screening to ensure that those will be sent for further spezialized training will perform well. Finally, it is during the basic training phase when nuclear reactor operators will start to acquire and develop attitudes regarding reactor operation and it is important that these be properly founded. (orig.)

  14. Safety Analysis for Power Reactor Protection System

    International Nuclear Information System (INIS)

    Eisawy, E.A.; Sallam, H.

    2012-01-01

    The main function of a Reactor Protection System (RPS) is to safely shutdown the reactor and prevents the release of radioactive materials. The purpose of this paper is to present a technique and its application for used in the analysis of safety system of the Nuclear Power Plant (NPP). A more advanced technique has been presented to accurately study such problems as the plant availability assessments and Technical Specifications evaluations that are becoming increasingly important. The paper provides the Markov model for the Reactor Protection System of the NPP and presents results of model evaluations for two testing policies in technical specifications. The quantification of the Markov model provides the probability values that the system will occupy each of the possible states as a function of time.

  15. Waste inventory, waste characteristics and waste repositories in Japan

    International Nuclear Information System (INIS)

    Shimooka, K.

    1997-01-01

    There are two types of repositories for the low level radioactive wastes in Japan. One is a trench type repository only for concrete debris generated from the dismantling of the research reactor. According to the safety assurance system, Japan Atomic Energy Research Institute (JAERI) has disposed of the concrete debris arose from the dismantling of the Japan Power Demonstration Reactor (JPDR). The other type is the concreted pit with engineered barriers. Rokkasho Low Level Radioactive Waste Disposal Center has this type of repository mainly for the power plant wastes. Japan Nuclear Fuel Ltd. (JNFL) established by electric power companies is the operator of the LLW disposal project. JNFL began the storage operation in 1992 and buried approximately 60,000 drums there. Two hundred thousand drums of uniformly solidified, waste may be buried ultimately. 4 refs, 3 tabs

  16. Safeguarding on-power fuelled reactors - instrumentation and techniques

    International Nuclear Information System (INIS)

    Waligura, A.; Konnov, Y.; Smith, R.M.; Head, D.A.

    1977-01-01

    Instrumentation and techniques applicable to safeguarding reactors that are fuelled on-power, particularly the CANDU type, have been developed. A demonstration is being carried out at the Douglas Point Nuclear Generating Station in Canada. Irradiated nuclear materials in certain areas - the reactor and spent fuel storage bays - are monitored using photographic and television cameras, and seals. Item accounting is applied by counting spent-fuel bundles during transfer from the reactor to the storage bay and by placing these spent-fuel bundles in a sealed enclosure. Provision is made for inspection and verification of the bundles before sealing. The reactor's power history is recorded by a track-etch power monitor. Redundancy is provided so that the failure of any single piece of equipment does not invalidate the entire safeguards system. Several safeguards instruments and devices have beeen developed and evaluated. These include a super-8 mm surveillance camera system, a television surveillance system, a spent-fuel bundle counter, a device to detect dummy fuel bundles, a cover for enclosing a stack of spent-fuel bundles, and a seal suitable for underwater installation and ultrasonic interrogation. The information provided by these different instruments should increase the effectiveness of Agency safeguards and, when used in combination with other measures, will facilitate inspection at reactor sites

  17. The core design of ALFRED, a demonstrator for the European lead-cooled reactors

    International Nuclear Information System (INIS)

    Grasso, G.; Petrovich, C.; Mattioli, D.; Artioli, C.; Sciora, P.; Gugiu, D.; Bandini, G.; Bubelis, E.; Mikityuk, K.

    2014-01-01

    Highlights: • The design for the lead fast reactor is conceived in a comprehensive approach. • Neutronic, thermal-hydraulic, and transient analyses show promising results. • The system is designed to withstand even design extension conditions accidents. • Activation products in lead, including polonium, are evaluated. - Abstract: The European Union has recently co-funded the LEADER (Lead-cooled European Advanced DEmonstration Reactor) project, in the frame of which the preliminary designs of an industrial size lead-cooled reactor (1500 MW th ) and of its demonstrator reactor (300 MW th ) were developed. The latter is called ALFRED (Advanced Lead-cooled Fast Reactor European Demonstrator) and its core, as designed and characterized in the project, is presented here. The core parameters have been fixed in a comprehensive approach taking into account the main technological constraints and goals of the system from the very beginning: the limiting temperature of the clad and of the fuel, the Pu enrichment, the achievement of a burn-up of 100 GWd/t, the respect of the integrity of the system even in design extension conditions (DEC). After the general core design has been fixed, it has been characterized from the neutronic point of view by two independent codes (MCNPX and ERANOS), whose results are compared. The power deposition and the reactivity coefficient calculations have been used respectively as input for the thermal-hydraulic analysis (TRACE, CFD and ANTEO codes) and for some preliminary transient calculations (RELAP, CATHARE and SIM-LFR codes). The results of the lead activation analysis are also presented (FISPACT code). Some issues of the core design are to be reviewed and improved, uncertainties are still to be evaluated, but the verifications performed so far confirm the promising safety features of the lead-cooled fast reactors

  18. The core design of ALFRED, a demonstrator for the European lead-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Grasso, G., E-mail: giacomo.grasso@enea.it [ENEA (Italian National Agency for New Technologies, Energy and Sustainable Economic Development), via Martiri di Monte Sole, 4, 40129 Bologna (Italy); Petrovich, C., E-mail: carlo.petrovich@enea.it [ENEA (Italian National Agency for New Technologies, Energy and Sustainable Economic Development), via Martiri di Monte Sole, 4, 40129 Bologna (Italy); Mattioli, D., E-mail: davide.mattioli@enea.it [ENEA (Italian National Agency for New Technologies, Energy and Sustainable Economic Development), via Martiri di Monte Sole, 4, 40129 Bologna (Italy); Artioli, C., E-mail: carlo.artioli@enea.it [ENEA (Italian National Agency for New Technologies, Energy and Sustainable Economic Development), via Martiri di Monte Sole, 4, 40129 Bologna (Italy); Sciora, P., E-mail: pierre.sciora@cea.fr [CEA (Alternative Energies and Atomic Energy Commission), DEN, DER, 13108 St Paul lez Durance (France); Gugiu, D., E-mail: daniela.gugiu@nuclear.ro [RATEN-ICN (Institute for Nuclear Research), Cod 115400 Mioveni, Str. Campului, 1, Jud. Arges (Romania); Bandini, G., E-mail: giacomino.bandini@enea.it [ENEA (Italian National Agency for New Technologies, Energy and Sustainable Economic Development), via Martiri di Monte Sole, 4, 40129 Bologna (Italy); Bubelis, E., E-mail: evaldas.bubelis@kit.edu [KIT (Karlsruhe Institute of Technology), Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Mikityuk, K., E-mail: konstantin.mikityuk@psi.ch [PSI (Paul Scherrer Institute), OHSA/D11, 5232 Villigen PSI (Switzerland)

    2014-10-15

    Highlights: • The design for the lead fast reactor is conceived in a comprehensive approach. • Neutronic, thermal-hydraulic, and transient analyses show promising results. • The system is designed to withstand even design extension conditions accidents. • Activation products in lead, including polonium, are evaluated. - Abstract: The European Union has recently co-funded the LEADER (Lead-cooled European Advanced DEmonstration Reactor) project, in the frame of which the preliminary designs of an industrial size lead-cooled reactor (1500 MW{sub th}) and of its demonstrator reactor (300 MW{sub th}) were developed. The latter is called ALFRED (Advanced Lead-cooled Fast Reactor European Demonstrator) and its core, as designed and characterized in the project, is presented here. The core parameters have been fixed in a comprehensive approach taking into account the main technological constraints and goals of the system from the very beginning: the limiting temperature of the clad and of the fuel, the Pu enrichment, the achievement of a burn-up of 100 GWd/t, the respect of the integrity of the system even in design extension conditions (DEC). After the general core design has been fixed, it has been characterized from the neutronic point of view by two independent codes (MCNPX and ERANOS), whose results are compared. The power deposition and the reactivity coefficient calculations have been used respectively as input for the thermal-hydraulic analysis (TRACE, CFD and ANTEO codes) and for some preliminary transient calculations (RELAP, CATHARE and SIM-LFR codes). The results of the lead activation analysis are also presented (FISPACT code). Some issues of the core design are to be reviewed and improved, uncertainties are still to be evaluated, but the verifications performed so far confirm the promising safety features of the lead-cooled fast reactors.

  19. Current Status and Future Outlook of Nuclear Power Generation in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, Yasuro; Yoshii, Ryosuke

    2007-07-01

    For Japan, a country poor in natural resources, in light of the tough energy situation in recent times, a National Energy Strategy with energy security at its core was established in May 2006. The key point of the Strategy is nuclear power generation, and the aim is to ensure that nuclear power generation continues to account for 30 to 40 percent or more of total electricity generated even after 2030. The first step to achieving this goal is to make maximum use of existing plants (55 plants, 49580MWe), and the aim is to achieve a 60-year service life by making improvements to plant operation and maintenance, such as extending current monitoring and maintenance of plant condition, and the implementation of plant aging management. In Japan, plant construction has been continuous since the 1970s. The current new plant construction plan (13 plants, 17230MWe) is to be achieved with a concerted, cohesive national effort. In addition, in order to complete the nuclear fuel cycle, a reprocessing plant is being constructed strictly for peaceful use, and construction of a site for disposing of high-level radioactive waste is also proceeding. Development of the next generation light water reactors and fast breeder reactor cycle is also underway. (auth)

  20. Maintenance management of nuclear power reactors at the stage of research and development

    International Nuclear Information System (INIS)

    Takaya, Shigeru; Chikazawa, Yoshitaka; Kubo, Shigenobu; Hayashida, Kiichi; Tagawa, Akihiro; Yamashita, Atsushi

    2016-07-01

    A maintenance management required to nuclear power reactors at the R and D stage was discussed in this report. It is the most important to ensure safety of nuclear power plants by taking account of characteristics of nuclear power reactors at the R and D stage. In addition, it is needed to establish a system of maintenance management technologies suitable for reactor types. In this report, objectives of maintenance management of nuclear power reactors at the R and D stage were clarified. Next, requirements and consideration for maintenance management of nuclear power reactors at the R and D stage were discussed according to the objectives. 'Code for Maintenance at Nuclear Power Plants' and 'Guide for Maintenance at Nuclear Power Plants' published by the Japan Electric Association were refereed in the discussion. Then, a draft of codes for maintenance management of nuclear power plants at the R and D stage was newly proposed. Finally, an example that the draft codes were applied to components containing sodium, typical components of sodium-cooled fast reactor, was presented. (author)

  1. Source driven breeding thermal power reactors

    International Nuclear Information System (INIS)

    Greenspan, E.; Ben-Gurion Univ. of the Negev, Beersheba

    1978-03-01

    The feasibility of fusion devices operating in the semi-catalyzed deuterium (SCD) mode and of high energy proton accelerators to provide the neutron sources for driving subcritical breeding light water power reactors is assessed. The assessment is done by studying the energy balance of the resulting source driven light water reactors (SDLWR) and comparing it with the energy balance of the reference light water hybrid reactors (LWHR) driven by a D-T neutron source (DT-LWHR). The conditions the non-DT neutron sources should satisfy in order to make the SDLWR viable power reactors are identified. It is found that in order for a SCD-LWHR to have the same overall efficiency as a DT-LWHR, the fusion energy gain of the SCD device should be at least one half that the DT device. The efficienct of ADLWRs using uranium targets is comparable with that of DT-LWHRs having a fusion energy gain of unity. Advantages and disadvantages of the DT-LWHR, SCD-LWHR and ADLWR are discussed. (aurthor)

  2. High Performance Auxiliary Power Unit Technology Demonstrator.

    Science.gov (United States)

    1980-12-01

    aft bearings 1.13 P3 - Power producer CDP 1.14 DPHE - Lube pressure drop at heat exchanger 1.15 POFP - Load airflow orifice pressure 1.16 DPOFP - Load...P𔃽I -PSI G PEBL -PSIG P2 -PS.IG DPHE -PID POFP -F Iu 0. 022±_ 77. 3478 6o5. 6 4±4 ±8L-. 4852 19. 51-17.4 DPOFP -PSID Ni -,. N2-i -RPM NSATM -FPM...28. 0250 83. 3505 29. 861 1:9. 7680 PGi -PSIG PEBL -PSIG P3 -PSIG DPHE -PSID POFP -PSIG 0. 0100 77. 9199 72.4862 17. 25 ±19. 4122 1= DPOFP -PSID NI

  3. International nuclear reactor hazard study. Design and operational features, and hazards of commercial nuclear power reactors in the world. Vol. 1

    International Nuclear Information System (INIS)

    Anderson, R.; Benecke, J.; Donderer, R.; Etemad, S.; Hirsch, H.; Kjellstroem, B.; Large, J.; Scheer, J.; Schneider, M.; Schumacher, H.; Schumacher, O.; Scott, M.; Takagi, J.; Thompson, G.; Torrie, R.; Ziggel, H.

    1986-09-01

    Project management and scientific coordination was performed by Gruppe Oekologie Hannover. The experts from Canada, the Federal Republic of Germany, France, Great Britain, Sweden, USA, and Japan represented an annulated experience of work on nuclear safety. The panel members analyzed the reactor types of their respective countries. The draft reports on the individual reactor types were discussed. Also included are brief descriptions of the different reactor types with simple drawing, evolution of reactor types and a brief report on nuclear power plants in the world. (DG)

  4. A series of lectures on operational physics of power reactors

    International Nuclear Information System (INIS)

    Mohanakrishnan, P.; Rastogi, B.P.

    1982-01-01

    This report discusses certain aspects of operational physics of power reactors. These form a lecture series at the Winter College on Nuclear Physics and Reactors, Jan. - March 1980, conducted at the International Centre for Theoretical Physics, Trieste, Italy. The topics covered are (a) the reactor physics aspects of fuel burnup (b) theoretical methods applied for burnup prediction in power reactors (c) interpretation of neutron detector readings in terms of adjacent fuel assembly powers (d) refuelling schemes used in power reactors. The reactor types chosen for the discussion are BWR, PWR and PHWR. (author)

  5. Source driven breeding thermal power reactors

    International Nuclear Information System (INIS)

    Greenspan, E.; Schneider, A.; Misulovin, A.; Gilai, D.; Levin, P.; Ben-Gurion Univ. of the Negev, Beersheba

    1978-03-01

    Improvements in the performance of fission power reactors made possible by designing them subcritical driven by D-T neutron sources are investigated. Light-water thermal systems are found to be most promising, neutronically and energetically, for the source driven mode of operation. The range of performance characteristics expected from breeding Light Water Hybrid Reactors (LWHR) is defined. Several promising types of LWHR blankets are identified. Options opened for the nuclear energy strategy by four types of the LWHRs are examined, and the potential contribution of these LWHRs to the nuclear energy economy are discussed. The power systems based on these LWHRs are found to enable a high utilization of the energy content of the uranium resources in all forms available - including depleted uranium and spent fuel from LWRs, while being free from the need for uranium enrichment and plutonium separation capabilities. (author)

  6. Water electrolysis plants for hydrogen and oxygen production. Shipped to Tsuruga Power Station Unit No.1, and Tokai No.2 power station, the Japan Atomic Power Co

    International Nuclear Information System (INIS)

    Ueno, Syuichi; Sato, Takao; Ishikawa, Nobuhide

    1997-01-01

    Ebara's water electrolysis plants have been shipped to Tsuruga Power Station Unit No.1, (H 2 generation rate: 11 Nm 3 /h), and Tokai No.2 Power Station (H 2 generation rate: 36 Nm 3 /h), Japan Atomic Power Co. An outcome of a business agreement between Nissho Iwai Corporation and Norsk Hydro Electrolysers (Norway), this was the first time that such water electrolysis plants were equipped in Japanese boiling water reactor power stations. Each plant included an electrolyser (for generating hydrogen and oxygen), an electric power supply, a gas compression system, a dehumidifier system, an instrumentation and control system, and an auxiliary system. The plant has been operating almost continuously, with excellent feedback, since March 1997. (author)

  7. Tritium supply assessment for ITER and DEMOnstration power plant

    International Nuclear Information System (INIS)

    Ni, Muyi; Wang, Yongliang; Yuan, Baoxin; Jiang, Jieqiong; Wu, Yican

    2013-01-01

    Highlights: • The tritium production rate in CANDU reactor was simulated and estimated. • Possible routes, including APT, CLWR and tritium production schemes of ADS, were evaluated in feasibility and economy. • The possible tritium consumption of ITER and initial supply for DEMO was assessed. • Result of supply and demand showed that after ITER retired in 2038, the tritium production in CANDU reactor might not be enough for a FDS-II scale DEMO reactor startup if without additional tritium resource. -- Abstract: The International Thermonuclear Experimental Reactor (ITER) and next generation DEMOnstration fusion reactor need amounts of tritium for test/initial startup and will consume kilograms tritium for operation per year. The available supply of tritium for fusion reactor is man-made sources. Now most of commercial tritium resource is extracted from moderator and coolant of CANada Deuterium Uranium (CANDU) type Heavy Water Reactor (HWR), in the Ontario Hydro Darlington facility of Canada and Wolsong facility of Korea. In this study, the tritium production rate in CANDU reactor was simulated and estimated. And other possible routes, including Accelerator Production of Tritium (APT), tritium production in Commercial Light Water Reactor (CLWR) and Accelerator Driven Subcritical system (ADS), were also evaluated in feasibility and economy. Based on the tritium requirement investigated according to ITER test schedule and startup inventory required for a FDS-II-scale DEMO calculated by TAS1.0, the assessment results showed that after ITER retired in 2038, the tritium inventory of CANDU reactor could not afford DEMO reactor startup without extra resource

  8. Tritium supply assessment for ITER and DEMOnstration power plant

    Energy Technology Data Exchange (ETDEWEB)

    Ni, Muyi, E-mail: muyi.ni@fds.org.cn; Wang, Yongliang; Yuan, Baoxin; Jiang, Jieqiong; Wu, Yican

    2013-10-15

    Highlights: • The tritium production rate in CANDU reactor was simulated and estimated. • Possible routes, including APT, CLWR and tritium production schemes of ADS, were evaluated in feasibility and economy. • The possible tritium consumption of ITER and initial supply for DEMO was assessed. • Result of supply and demand showed that after ITER retired in 2038, the tritium production in CANDU reactor might not be enough for a FDS-II scale DEMO reactor startup if without additional tritium resource. -- Abstract: The International Thermonuclear Experimental Reactor (ITER) and next generation DEMOnstration fusion reactor need amounts of tritium for test/initial startup and will consume kilograms tritium for operation per year. The available supply of tritium for fusion reactor is man-made sources. Now most of commercial tritium resource is extracted from moderator and coolant of CANada Deuterium Uranium (CANDU) type Heavy Water Reactor (HWR), in the Ontario Hydro Darlington facility of Canada and Wolsong facility of Korea. In this study, the tritium production rate in CANDU reactor was simulated and estimated. And other possible routes, including Accelerator Production of Tritium (APT), tritium production in Commercial Light Water Reactor (CLWR) and Accelerator Driven Subcritical system (ADS), were also evaluated in feasibility and economy. Based on the tritium requirement investigated according to ITER test schedule and startup inventory required for a FDS-II-scale DEMO calculated by TAS1.0, the assessment results showed that after ITER retired in 2038, the tritium inventory of CANDU reactor could not afford DEMO reactor startup without extra resource.

  9. Hybrid reactors: recent progress of a demonstration pilot; Reacteurs hybrides: avancees recentes pour un demonstrateur

    Energy Technology Data Exchange (ETDEWEB)

    Billebaud, Annick [Laboratoire de Physique Subatomique et de Cosmologie IN2P3-CNRS/UJF/INPG, 53 av. des Martyrs, 38026 Grenoble Cedex (France)

    2006-12-15

    Accelerator driven sub-critical reactors are subject of many research programmes since more than ten years, with the aim of testing the feasibility of the concept as well as their efficiency as a transmutation tool. Several key points like the accelerator, the spallation target, or neutronics in a subcritical medium were investigated extensively these last years, allowing for technological choices and the design of a low power European demonstration ADS (a few tens of MWth). Programmes dedicated to subcritical reactor piloting proposed a monitoring procedure to be validated in forthcoming experiments. Accelerator R and D provided the design of a LINAC for an ADS and research work on accelerator reliability is going on. A spallation target was operated at PSI and the design of a windowless target is in progress. All this research work converges to the design of a European demonstration ADS, the ETD/XT-ADS, which could be the Belgian MYRRHA project. (author)

  10. Power conditioning for space nuclear reactor systems

    Science.gov (United States)

    Berman, Baruch

    1987-01-01

    This paper addresses the power conditioning subsystem for both Stirling and Brayton conversion of space nuclear reactor systems. Included are the requirements summary, trade results related to subsystem implementation, subsystem description, voltage level versus weight, efficiency and operational integrity, components selection, and shielding considerations. The discussion is supported by pertinent circuit and block diagrams. Summary conclusions and recommendations derived from the above studies are included.

  11. Compact approach to fusion power reactors

    International Nuclear Information System (INIS)

    Hagenson, R.L.; Krakowski, R.A.; Bathke, C.G.; Miller, R.L.

    1984-01-01

    The potential of the Reversed-Field Pinch (RFP) for development into an efficient, compact, copper-coil fusion reactor has been quantified by comprehensive parametric tradeoff studies. These compact systems promise to be competitive in size, power density, and cost to alternative energy sources. Conceptual engineering designs that largely substantiate these promising results have since been completed. This 1000-MWe(net) design is described along with a detailed rationale and physics/technology assessment for the compact approach to fusion

  12. Small and medium power reactors 1985

    International Nuclear Information System (INIS)

    1986-05-01

    This report is intended for designers and planners concerned with Small and Medium Power Reactors. It provides a record of the presentations during the meetings held on this subject at the Agency's General Conference in September 1985. This information should be useful as it indicates the principal findings and main conclusions and recommendations resulting from these meetings. A separate abstract was prepared for each of the 10 presentations in this report

  13. Leaching of nuclear power reactor wastes forms

    International Nuclear Information System (INIS)

    Endo, L.S.; Villalobos, J.P.; Miyamoto, H.

    1986-01-01

    The leaching tests for power reactor wastes carried out at IPEN/CNEN-SP are described. These waste forms consist mainly of spent resins and boric acid concentrates solidified in ordinary Portland cement. All tests were conducted according to the ISO and IAEA recommendations. 3 years leaching results are reported, determining cesium and strontium diffusivity coefficients for boric acid waste form and ion-exchange resins. (Author) [pt

  14. Current situation and future of research reactor in Japan

    International Nuclear Information System (INIS)

    Metoki, Naoto

    2011-01-01

    A new style of neutron facilities coexisting high intensity pulse neutron source (J-PARC/MLF) and steady state neutron from research reactor (JRR-3) is proposed. The coexistence of the pulse and reactor source is the world trend which is inevitable for the next generation neutron science. A sophisticated management and comprehensive user program are important for Japanese neutron community, involving high performance instruments with pulse neutron and steady neutron source with widely spread users. JAEA is the most responsible for the future of neutron science, because of the commitment for both type of neutron source, J-PARC/MLF and JRR-3. (author)

  15. Preliminary Demonstration Reactor Point Design for the Fluoride Salt-Cooled High-Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Qualls, A. L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Betzler, Benjamin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Carbajo, Juan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Greenwood, Michael Scott [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hale, Richard Edward [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Harrison, Thomas J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrell, Jerry W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-12-01

    Development of the Fluoride Salt-Cooled High-Temperature Reactor (FHR) Demonstration Reactor (DR) is a necessary intermediate step to enable commercial FHR deployment through disruptive and rapid technology development and demonstration. The FHR DR will utilize known, mature technology to close remaining gaps to commercial viability. Lower risk technologies are included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated within an acceptable budget and schedule. These technologies include tristructural-isotropic (TRISO) particle fuel, replaceable core structural material, the use of that same material for the primary and intermediate loops, and tube-and-shell heat exchangers. This report provides an update on the development of the FHR DR. At this writing, the core neutronics and thermal hydraulics have been developed and analyzed. The mechanical design details are still under development and are described to their current level of fidelity. It is anticipated that the FHR DR can be operational within 10 years because of the use of low-risk, near-term technology options.

  16. Preliminary Demonstration Reactor Point Design for the Fluoride Salt-Cooled High-Temperature Reactor

    International Nuclear Information System (INIS)

    Qualls, A. L.; Betzler, Benjamin R.; Brown, Nicholas R.; Carbajo, Juan; Greenwood, Michael Scott; Hale, Richard Edward; Harrison, Thomas J.; Powers, Jeffrey J.; Robb, Kevin R.; Terrell, Jerry W.

    2015-01-01

    Development of the Fluoride Salt-Cooled High-Temperature Reactor (FHR) Demonstration Reactor (DR) is a necessary intermediate step to enable commercial FHR deployment through disruptive and rapid technology development and demonstration. The FHR DR will utilize known, mature technology to close remaining gaps to commercial viability. Lower risk technologies are included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated within an acceptable budget and schedule. These technologies include tristructural-isotropic (TRISO) particle fuel, replaceable core structural material, the use of that same material for the primary and intermediate loops, and tube-and-shell heat exchangers. This report provides an update on the development of the FHR DR. At this writing, the core neutronics and thermal hydraulics have been developed and analyzed. The mechanical design details are still under development and are described to their current level of fidelity. It is anticipated that the FHR DR can be operational within 10 years because of the use of low-risk, near-term technology options.

  17. Evolution of on-power fuelling machines on Canadian natural uranium power reactors

    International Nuclear Information System (INIS)

    Isaac, P.

    1984-10-01

    The evolution of the on-power fuel changing process and fuelling machines on CANDU heavy-water pressure tube power reactors from the first nuclear power demonstration plant, 22 MWe NPD, to the latest plants now in design and development is described. The high availability of CANDU's is largely dependent on on-power fuelling. The on-power fuelling performance record of the 16 operating CANDU reactors, covering a 22 year period since the first plant became operational, is given. This shows that on-power fuel changing with light (unshielded), highly mobile and readily maintainable fuelling machines has been a success. The fuelling machines have contributed very little to the incapabilities of the plants and have been a key factor in placing CANDUs in the top ten list of world performance. Although fuel handling technology has reached a degree of maturity, refinements are continuing. A new single-ended fuel changing concept for horizontal reactors under development is described. This has the potential for reducing capital and operating costs for small reactors and increasing the fuelling capability of possible large reactors of the future

  18. Design approaches for enhancing the engineering feasibility of tokamak power reactors

    International Nuclear Information System (INIS)

    Shannon, T.E.; Steiner, D.

    1977-01-01

    The design approach developed in the ORNL Fusion Power Demonstration Study is reviewed. The design concepts having greatest impact on reactor feasibility by the application of current or near term technology are described briefly. These are: blanket structural material, blanket coolant, power conversion system, and pulsed electrical system. Concepts relative to the approach taken to simplify the overall reactor design are listed

  19. Response of pressurized water reactor (PWR) to network power generation demands

    International Nuclear Information System (INIS)

    Schreiner, L.A.

    1991-01-01

    The flexibility of the PWR type reactor in terms of response to the variations of the network power demands, is demonstrated. The factors that affect the transitory flexibility and some design prospects that allow the reactor fits the requirements of the network power demands, are also discussed. (M.J.A.)

  20. Critical Power Response to Power Oscillations in Boiling Water Reactors

    International Nuclear Information System (INIS)

    Farawila, Yousef M.; Pruitt, Douglas W.

    2003-01-01

    The response of the critical power ratio to boiling water reactor (BWR) power oscillations is essential to the methods and practice of mitigating the effects of unstable density waves. Previous methods for calculating generic critical power response utilized direct time-domain simulations of unstable reactors. In this paper, advances in understanding the nature of the BWR oscillations and critical power phenomena are combined to develop a new method for calculating the critical power response. As the constraint of the reactor state - being at or slightly beyond the instability threshold - is removed, the new method allows the calculation of sensitivities to different operation and design parameters separately, and thus allows tighter safety margins to be used. The sensitivity to flow rate and the resulting oscillation frequency change are given special attention to evaluate the extension of the oscillation 'detect-and-suppress' methods to internal pump plants where the flow rate at natural circulation and oscillation frequency are much lower than jet pump plants

  1. Analysis of accidents and troubles of nuclear power plants in Japan

    International Nuclear Information System (INIS)

    Kobayashi, Kunio

    1980-01-01

    In Japan, electric power companies are obliged to report the accidents and troubles occurred in nuclear power stations to the MITI according to the relevant laws, and 166 cases in total have been reported as of the end of March, 1980. These accidents and troubles are all trivial, and do not cause problems from the viewpoint of the safety nuclear power stations. Regarding respective accidents and troubles, the causes have been sought thoroughly, and the sufficient countermeasures have been taken on all occasions. But in order to improve the reliability of nuclear power stations further, it is important to treat the accidents and troubles occurred so far statistically and grasp the general trend. Thereupon, 152 accidents and troubles occurred till September, 1979, were analyzed quantitatively, and the results are reported in this paper. From the results, the prospect hereafter is discussed. The number of the reported cases of accidents and troubles in each nuclear power plant in operation every year is tabulated. The accidents and troubles were relatively frequent in the initial two or three years of operation of respective new reactor types, but decreased thereafter. The systems to which troubled equipments belong and the troubled equipments are shown. Most troubles have occurred in reactor cooling systems and valves. The situations and causes of troubles, the operational conditions at the time of the accidents and troubles and the effects and others are reported. (Kako, I.)

  2. Review of tokamak power reactor and blanket designs in the United States

    International Nuclear Information System (INIS)

    Baker, C.; Brooks, J.; Ehst, D.; Gohar, Y.; Smith, D.; Sze, D.

    1986-01-01

    The last major conceptual design study of a tokamak power reactor in the United States was STARFIRE which was carried out in 1979-1980. Since that time US studies have concentrated on engineering test reactors, demonstration reactors, parametric systems studies, scoping studies, and studies of selected critical issues such as pulsed vs. steady-state operation and blanket requirements. During this period, there have been many advancements in tokamak physics and reactor technology, and there has also been a recognition that it is desirable to improve the tokamak concept as a commercial power reactor candidate. During 1984-1985 several organizations participated in the Tokamak Power Systems Study (TPSS) with the objective of developing ideas for improving the tokamak as a power reactor. Also, the US completed a comprehensive Blanket Comparison and Selection Study which formed the basis for further studies on improved blankets for fusion reactors

  3. Water chemistry experience of nuclear power plants in Japan

    International Nuclear Information System (INIS)

    Ishigure, Kenkichi; Abe, Kenji; Nakajima, Nobuo; Nagao, Hiroyuki; Uchida, Shunsuke.

    1989-01-01

    Japanese LWRs have experienced several troubles caused by corrosions of structural materials in the past ca. 20 years of their operational history, among which are increase in the occupational radiation exposures, intergranular stress corrosion cracking (IGSCC) of stainless steel piping in BWR, and steam generator corrosion problems in PWR. These problems arised partly from the improper operation of water chemistry control of reactor coolant systems. Consequently, it has been realized that water chemistry control is one of the most important factors to attain high availability and reliability of LWR, and extensive researches and developments have been conducted in Japan to achieve the optimum water chemistry control, which include the basic laboratory experiments, analyses of plant operational data, loop tests in operating plants and computer code developments. As a result of the continuing efforts, the Japanese LWR plants have currently attained a very high performance in their operation with high availability and low occupational radiation exposures. A brief review is given here on the R and D of water chemistry in Japan. (author)

  4. Proposal on experience learning of a nuclear reactor for children in future. A basic concept on a nuclear reactor facility for demonstration and education

    International Nuclear Information System (INIS)

    Murata, Takashi; Yoshiki, Nobuya; Kinehara, Yoshiki; Nakagawa, Haruo

    2001-01-01

    The Science Council of Japan indicates in a proposal on R and D on nuclear energy forward the 21st Century that it is important to expand the educational object on nuclear energy from colleges and gradual schools to elementary, middle high schools. And, the Committee of Japan Nuclear Energy Industries also proposed that as an effort forward security of reliability and popularization of knowledge, completeness of learning chance on energy and nuclear energy in education such as usage of general learning time, concept on establishment of educational reactor for demonstration and experience, is essential. Here was described on a concept on establishment of nuclear reactor for demonstration and experience at objectives of common national peoples, which was based on results of searches and investigations carried out by authors and aimed to supply to a field to grow up a literary adequately and widely capable of judging various information on the peoples by focusing to effectiveness of empirical learning as a method of promoting corrective understanding of common citizens on high class technical system and by establishment of the reactor aiming at general education on nuclear energy at a place easily accessible by common citizens, such as large city. (G.K.)

  5. Proposal on experience learning of a nuclear reactor for children in future. A basic concept on a nuclear reactor facility for demonstration and education

    Energy Technology Data Exchange (ETDEWEB)

    Murata, Takashi [Kyoto Univ., Graduate School of Energy Science, Kyoto (Japan); Yoshiki, Nobuya; Kinehara, Yoshiki; Nakagawa, Haruo

    2001-12-01

    The Science Council of Japan indicates in a proposal on R and D on nuclear energy forward the 21st Century that it is important to expand the educational object on nuclear energy from colleges and gradual schools to elementary, middle high schools. And, the Committee of Japan Nuclear Energy Industries also proposed that as an effort forward security of reliability and popularization of knowledge, completeness of learning chance on energy and nuclear energy in education such as usage of general learning time, concept on establishment of educational reactor for demonstration and experience, is essential. Here was described on a concept on establishment of nuclear reactor for demonstration and experience at objectives of common national peoples, which was based on results of searches and investigations carried out by authors and aimed to supply to a field to grow up a literary adequately and widely capable of judging various information on the peoples by focusing to effectiveness of empirical learning as a method of promoting corrective understanding of common citizens on high class technical system and by establishment of the reactor aiming at general education on nuclear energy at a place easily accessible by common citizens, such as large city. (G.K.)

  6. Nuclear power reactors and hydrogen storage systems

    International Nuclear Information System (INIS)

    Ibrahim Aly Mahmoud El Osery.

    1980-01-01

    Among conclusions and results come by, a nuclear-electric-hydrogen integrated power system was suggested as a way to prevent the energy crisis. It was shown that the hydrogen power system using nuclear power as a leading energy resource would hold an advantage in the current international situation as well as for the long-term future. Results reported provide designers of integrated nuclear-electric-hydrogen systems with computation models and routines which will allow them to explore the optimal solution in coupling power reactors to hydrogen producing systems, taking into account the specific characters of hydrogen storage systems. The models were meant for average computers of a type easily available in developing countries. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-01

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

  8. Fluoride Salt-Cooled High-Temperature Demonstration Reactor Point Design

    Energy Technology Data Exchange (ETDEWEB)

    Qualls, A. L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Betzler, Benjamin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Carbajo, Juan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hale, Richard Edward [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Harrison, Thomas J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrell, Jerry W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wysocki, Aaron J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-02-01

    concept, and it will demonstrate key operational features of that design. The FHR DR will be closely scaled to the SmAHTR concept in power and flows, so any technologies demonstrated will be directly applicable to a reactor concept of that size. The FHR DR is not a commercial prototype design, but rather a DR that serves a cost and risk mitigation function for a later commercial prototype. It is expected to have a limited operational lifetime compared to a commercial plant. It is designed to be a low-cost reactor compared to more mature advanced prototype DRs. A primary reason to build the FHR DR is to learn about salt reactor technologies and demonstrate solutions to remaining technical gaps.

  9. Role of nuclear power in energy policy of Japan

    International Nuclear Information System (INIS)

    Ikuta, Toyoaki

    1984-01-01

    About 62 % of the energy supply in Japan depends on petroleum, and about 65 % of oil import comes from Persian Gulf. It is very important to ensure the supply of oil for the energy policy of Japan. For the purpose, the conversation between oil producing countries and oil consuming countries is necessary, but all the conversation carried out so far failed. The oil consumption in the world continued to decline, and the situation of oil market changed. The future situation of oil market largely depends on the development of substitute energy resources for oil. In order to ensure the supply of oil, the buyer's market must continue, therefore, effort must be continued to expand substitute energy and to promote energy saving. As the energy policy hereafter, various energy resources should be most effectively used in combination. In this compound energy age, the importance of nuclear power increases. The stable supply and economical efficiency of energy must be taken in consideration with the same weight. The only method to reduce the dependence on import is nuclear power, and this feature should be evaluated high. Nuclear power generation must be expanded hereafter. (Kako, I.)

  10. Data list of nuclear power plants in Japan

    International Nuclear Information System (INIS)

    Izumi, Fumio; Morishima, Atsuyoshi; Suzuki, Motoe; Harayama, Yasuo

    1980-07-01

    this report has collected and compiled the data concerning performances, equipments and installations of the nuclear power plants constructed in Japan by December 1979. The data have been modified according to the changes produced after publication of 1978 edition (JAERI-M 8083), and extended to cover the new plants developed thereafter. All data have been processed and tabulated with a computer program FREP for the exclusive use of data processing. While this plant data list has been edited annually, there are increasing use of this in foreign countries; hence, a commentary in English on the usage has been presented in the Appendix. (author)

  11. Technical Bases to Consider for Performance and Demonstration Testing of Space Fission Reactors

    International Nuclear Information System (INIS)

    Hixson, Laurie L.; Houts, Michael G.; Clement, Steven D.

    2004-01-01

    Performance and demonstration testing are critical to the success of a space fission reactor program. However, the type and extent to which testing of space reactors should be performed has been a point of discussion within the industry for many years. With regard to full power ground nuclear tests, questions such as 'Do the benefits outweigh the risks? Are there equivalent alternatives? Can a test facility be constructed (or modified) in a reasonable amount of time? Will the test article accurately represent the flight system? Are the costs too restrictive?' have been debated for decades. There are obvious benefits of full power ground nuclear testing such as obtaining systems integrated reliability data on a full-scale, complete end-to-end system. But these benefits come at some programmatic risk. In addition, this type of testing does not address safety related issues. This paper will discuss and assess these and other technical considerations essential in deciding which type of performance and demonstration testing to conduct on space fission reactor systems. (authors)

  12. Power reactor noise studies and applications

    Energy Technology Data Exchange (ETDEWEB)

    Arzhanov, V

    2002-03-01

    The present thesis deals with the neutron noise arising in power reactor systems. Generally, it can be divided into two major parts: first, neutron noise diagnostics, or more specifically, novel methods and algorithms to monitor nuclear industrial reactors; and second, contributions to neutron noise theory as applied to power reactor systems. Neutron noise diagnostics is presented by two topics. The first one is a theoretical study on the possibility to use a newly proposed current-flux (C/F) detector in Pressurised Water Reactors (PWR) for the localisation of anomalies. The second topic concerns various methods to detect guide tube impacting in Boiling Water Reactors (BWR). The significance of these problems comes from the operational experience. The thesis describes a novel method to localise vibrating control rods in a PWR by using only one C/F detector. Another novel method, based on wavelet analysis, is put forward to detect impacting guide tubes in a BWR. Neutron noise theory is developed for both Accelerator Driven Systems (ADS) and traditional reactors. By design the accelerator-driven systems would operate in a subcritical mode with a strong external source. This calls for a revision of many concepts and methods that have been developed for traditional reactors and also it poses a number of new problems. As for the latter, the thesis investigates the space-dependent neutron noise caused by a fluctuating source. It is shown that the frequency-dependent spatial behaviour exhibits some new properties that are different from those known in traditional critical systems. On the other hand, various reactor physics approximations (point kinetic, adiabatic etc.) have not been defined yet for the subcritical systems. In this respect the thesis presents a systematic formulation of the above mentioned approximations as well as investigations of their properties. Another important problem in neutron noise theory is the treatment of moving boundaries. In this case one

  13. Power reactor noise studies and applications

    International Nuclear Information System (INIS)

    Arzhanov, V.

    2002-03-01

    The present thesis deals with the neutron noise arising in power reactor systems. Generally, it can be divided into two major parts: first, neutron noise diagnostics, or more specifically, novel methods and algorithms to monitor nuclear industrial reactors; and second, contributions to neutron noise theory as applied to power reactor systems. Neutron noise diagnostics is presented by two topics. The first one is a theoretical study on the possibility to use a newly proposed current-flux (C/F) detector in Pressurised Water Reactors (PWR) for the localisation of anomalies. The second topic concerns various methods to detect guide tube impacting in Boiling Water Reactors (BWR). The significance of these problems comes from the operational experience. The thesis describes a novel method to localise vibrating control rods in a PWR by using only one C/F detector. Another novel method, based on wavelet analysis, is put forward to detect impacting guide tubes in a BWR. Neutron noise theory is developed for both Accelerator Driven Systems (ADS) and traditional reactors. By design the accelerator-driven systems would operate in a subcritical mode with a strong external source. This calls for a revision of many concepts and methods that have been developed for traditional reactors and also it poses a number of new problems. As for the latter, the thesis investigates the space-dependent neutron noise caused by a fluctuating source. It is shown that the frequency-dependent spatial behaviour exhibits some new properties that are different from those known in traditional critical systems. On the other hand, various reactor physics approximations (point kinetic, adiabatic etc.) have not been defined yet for the subcritical systems. In this respect the thesis presents a systematic formulation of the above mentioned approximations as well as investigations of their properties. Another important problem in neutron noise theory is the treatment of moving boundaries. In this case one

  14. The effect of the Fukushima nuclear accident on stock prices of electric power utilities in Japan

    International Nuclear Information System (INIS)

    Kawashima, Shingo; Takeda, Fumiko

    2012-01-01

    The purpose of this study is to investigate the effect of the accident at the Fukushima Daiichi nuclear power station, which is owned by Tokyo Electric Power Co. (TEPCO), on the stock prices of the other electric power utilities in Japan. Because the other utilities were not directly damaged by the Fukushima nuclear accident, their stock price responses should reflect the change in investor perceptions on risk and return associated with nuclear power generation. Our first finding is that the stock prices of utilities that own nuclear power plants declined more sharply after the accident than did the stock prices of other electric power utilities. In contrast, investors did not seem to care about the risk that may arise from the use of the same type of nuclear power reactors as those at the Fukushima Daiichi station. We also observe an increase of both systematic and total risks in the post-Fukushima period, indicating that negative market reactions are not merely caused by one-time losses but by structural changes in society and regulation that could increase the costs of operating a nuclear power plant.

  15. Boiler systems for nuclear powered reactors

    International Nuclear Information System (INIS)

    Cook, R.K.; George, B.V.

    1979-01-01

    A power generating plant which comprises a heat source, at least one main steam turbine and at least one main boiler heated by heat from the heat source and providing the steam to drive the turbine, comprises additionally at least one further steam turbine, smaller than the main turbine, and at least one further boiler, of lower capacity than the main boiler, and heated from the same heat source and providing steam for the further turbine. Particularly advantageous in nuclear power stations, where the heat source is a nuclear reactor, the invention enables peak loads, above the normal continuous rating of the main generators driven by the main turbines, to be met by the further turbine(s) and one or more further generators driven thereby. This enables the main turbines to be freed from the thermal stresses of rapid load changes, which stresses are more easily accommodated by the smaller and thus more tolerant further turbine(s). Thus auxiliary diesel-driven or other independent power plant may be made partly or wholly unnecessary. Further, low-load running which would be inefficient if achieved by means of the main turbine(s), can be more efficiently effected by shutting them down and using the smaller further turbine(s) instead. These latter may also be used to provide independent power for servicing the generating plant during normal operation or during emergency or other shutdown, and in this latter case may also serve as a heat sink for the shutdown reactor

  16. Reactor power cutback system test experience at YGN 4

    International Nuclear Information System (INIS)

    Chi, Sung Goo; Kim, Se Chang; Seo, Jong Tae; Eom, Young Meen; Wook, Jeong Dae; Choi, Young Boo

    1995-01-01

    YGN 3 and 4 are the nuclear power plants having System 80 characteristics with a rated thermal output of 2815 MWth and a nominal net electrical output of 1040 MWe. YGN 3 achieved commercial operation on March 31, 1995 and YGN 4 completed Power Ascension Test (PAT) at 20%, 50%, 80% and 100% power by September 23, 1995. YGN 3 and 4 design incorporates the Reactor POwer Cutback System (RPCS) which reduces plant trips caused by Loss of Load (LOL)/ Turbine Trip and Loss of One Main Feedwater Pump (LOMFWP). The key design objective of the RPCS is to improve overall plant availability and performance, while minimizing challenges to the plant safety systems. The RPCS is designed to rapidly reduce reactor power by dropping preselected Control Element Assemblies (CEAs) while other NSSS control systems maintain process parameters within acceptable ranges. Extensive RPCS related tests performed during the initial startup of YGN 4 demonstrated that the RPCS can maintain the reactor on-line without opening primary or secondary safety valves and without actuating the Engineered Safety Features Actuation System (ESFAS). It is expected that use of the RPCS at YGN will increase the overall availability of the units and reduce the number of challenges to plant safety systems

  17. Safeguarding on-power fuelled reactors - instrumentation and techniques

    International Nuclear Information System (INIS)

    Waligura, A.; Konnov, Y.; Smith, R.M.; Head, D.A.

    1977-05-01

    Instrumentation and techniques applicable to safeguarding reactors that are fuelled on-power, particularly the CANDU type, have been developed. A demonstration is being carried out at the Douglas Point Nuclear Generating Station in Canada. Irradiated nuclear materials in certain areas - the reactor and spent fuel storage bays - are monitored using photographic and television cameras, and seals. Item accounting is applied by counting spent-fuel bundles during transfer from the reactor to the storage bay and by placing these spent-fuel bundles in a sealed enclosure. Provision is made for inspection and verification of the bundles before sealing. The reactor's power history is recorded by a Track-Etch power monitor. Redundancy is provided so that the failure of any single piece of equipment does not invalidate the entire safeguards system. Several safeguards instruments and devices have been developed and evaluated. These include a super-8-mm surveillance camera system, a television surveillance system, a spent-fuel bundle counter, a device to detect dummy fuel bundles, a cover for enclosing a stack of spent-fuel bundles, and a seal suitable for underwater installation and ultrasonic interrogation. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-01

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

  19. Summary of advanced LMR [Liquid Metal Reactor] evaluations: PRISM [Power Reactor Inherently Safe Module] and SAFR [Sodium Advanced Fast Reactor

    International Nuclear Information System (INIS)

    Van Tuyle, G.J.; Slovik, G.C.; Chan, B.C.; Kennett, R.J.; Cheng, H.S.; Kroeger, P.G.

    1989-10-01

    In support of the US Nuclear Regulatory Commission (NRC), Brookhaven National Laboratory (BNL) has performed independent analyses of two advanced Liquid Metal Reactor (LMR) concepts. The designs, sponsored by the US Department of Energy (DOE), the Power Reactor Inherently Safe Module (PRISM) [Berglund, 1987] and the Sodium Advanced Fast Reactor (SAFR) [Baumeister, 1987], were developed primarily by General Electric (GE) and Rockwell International (RI), respectively. Technical support was provided to DOE, RI, and GE, by the Argonne National Laboratory (ANL), particularly with respect to the characteristics of the metal fuels. There are several examples in both PRISM and SAFR where inherent or passive systems provide for a safe response to off-normal conditions. This is in contrast to the engineered safety systems utilized on current US Light Water Reactor (LWR) designs. One important design inherency in the LMRs is the ''inherent shutdown'', which refers to the tendency of the reactor to transition to a much lower power level whenever temperatures rise significantly. This type of behavior was demonstrated in a series of unscrammed tests at EBR-II [NED, 1986]. The second key design feature is the passive air cooling of the vessel to remove decay heat. These systems, designated RVACS in PRISM and RACS in SAFR, always operate and are believed to be able to prevent core damage in the event that no other means of heat removal is available. 27 refs., 78 figs., 3 tabs

  20. Reactor power control device in BWR power plant

    International Nuclear Information System (INIS)

    Kurosawa, Tsuneo.

    1997-01-01

    The present invention provides a device for controlling reactor power based on a start-up/shut down program in a BWR type reactor, as well as for detecting deviation, if occurs, of the power from the start-up/shut down program, to control a recycling flow rate control system or control rod drive mechanisms. Namely, a power instruction section successively executes the start-up/shut down program and controls the coolant recycling system and the control rod driving mechanisms to control the power. A current state monitoring and calculation section receives a process amount, calculates parameters showing the plant state, compares/monitors them with predetermined values, detecting the deviation, if occurs, of the plant state from the start-up/shut down program, and prevents output of a power increase control signal which leads to power increase. A forecasting and monitoring/calculation section forecasts and calculates the plant state when not yet executed steps of the start-up/shut down program are performed, stops the execution of the start-up/shut down program in the next step in a case of forecasting that the results of the calculation will deviate from the start-up/shut down program. (I.S.)

  1. Reactor power control systems in nuclear power plants

    International Nuclear Information System (INIS)

    Nakajima, Kazuo.

    1980-01-01

    Purpose: To enable power control by automatic control rod operation based on the calculated amounts of operation for the control rods determined depending on a power set value from reactor operators or on power variation amounts from other devices. Constitution: When an operator designates an automatic selection by way of a control rod operation panel, automatic signals are applied to a manual-automatic switching circuit and the mode judging circuit of a rod pattern control device. Then, mode signals such as for single operation, load setting, load following and the like produced by the operator are judged in a circuit, wherein a control rod pattern operation circuit calculates the designation for the control rods and the operation amounts for the control rods depending on the designated modes and automatic control is conducted for the control rods by a rod position control circuit, a rod drive control device and the like connected at a rod position monitor device. The reactor power is thus controlled automatically to reduce the operator's labours. The automatic power control can also be conducted in the same manner by the amount of power variations applied to the device from the external device. (Yoshino, Y.)

  2. Location and public acceptance of nuclear power plants in Japan

    International Nuclear Information System (INIS)

    Takahashi, Hiroshi

    1978-01-01

    Analysis of problems concerning the sites of nuclear power plants and policy of the government to develop nuclear power plants are presented. At present, national consensus about atomic energy is not yet sufficient in Japan. Accordingly, it is hard to get proper location for nuclear power plants, and more effort is required. Reasons of the hindrance of atomic energy development are not same, but they are based on lack of understanding, social and local situations, and interests accompanying atomic energy development. Also, there is effects from the activities of opposition groups. The most important factor is lack of communication between those concerned with the development and residents around prospective sites. The government has investigated how to promote the atomic energy development, taking into account the present status of public acceptance. The system to promote the development of sites for nuclear power plants has been established. Political efforts for improving the welfare of residents have been made, and three laws for the purpose were approved. According to these laws, subsidiary money is paid to cities, towns and villages where power plants are located. Speeding up and smoothing of legal procedures concerning the location for power plants are also studied. (Kato, T.)

  3. System for forecasting a reactor power distribution

    International Nuclear Information System (INIS)

    Motoda, Hiroshi; Nishizawa, Yasuo.

    1976-01-01

    Purpose: To dispense with frequent running of detector in a BWR type reactor and permit calculation of the prevailing value and forecast value of power distribution in a specified region in an on-line basis. Constitution: The prevailing power distribution P sub(OZ) (where Z indicates a position in the axial direction) at a given position is estimated by prevailing power distribution estimating means, and the average prevailing power distribution Q sub(OZ) in the core is estimated while making correction of a primary neutron distribution model by core average characteristic measuring means. Then, the estimated core average power distribution Q sub(Z) after alteration of the core flow rate or alteration of Xe concentration is estimated by core average power distribution estimating means. At this time, a forecast power distribution P sub(Z) in a specified region after alteration of the flow rate or alteration of the Xe concentration is calculated on the basis of a relation P sub(Z) = (Q sub(Z)/Q sub(OZ)) by using P sub(OZ), Q sub(OZ) and Q sub(Z). The above calculations are carried out in a short period of time by using a process computer. (Ikeda, J.)

  4. Safety Analysis Of Actinide Recycled Fast Power Reactor

    International Nuclear Information System (INIS)

    Taufik, Mohammad

    2001-01-01

    Simulation for safety analysis of actinide recycled fast power reactor has been performed. The objective is to know reactor response about ULOF and ULOF and UTOP simultaneous accident. From parameter result such reactivity feedback, power, temperature, and cooled flow rate can conclusion that reactor have inherent safety system, which can back to new Equilibrium State

  5. Power distribution monitor for nuclear reactor

    International Nuclear Information System (INIS)

    Nishizawa, Yasuo; Kiguchi, Takashi.

    1974-01-01

    Object: To compare the measured local power region monitor (LPRM) index with the result of a primary calculation to correct the threshold condition for the primary calculation thereby to rapidly grasp and monitor the existing power distribution. Structure: The index of an LPRM disposed in a nuclear reactor is processed in a data processor to remove therefrom a noise, and transmitted to a threshold condition processor to be stored therein. The LPRM index measured by the threshold condition processor is compared with the calculated LPRM value transmitted from the primary processor, whereby the threshold condition is corrected and transmitted to the primary processor. After the completion of calculation, the traversing incore probe (TIP) indexing value is converted to a thermal output distribution or a linear output density distribution and transmitted to an output indicator or an output typewriter. The operator may monitor the existing power distribution by monitoring the output indicator. (Kamimura, M.)

  6. Power supply trip control for nuclear reactor

    International Nuclear Information System (INIS)

    Hager, R.E.; Gutman, Jerzy.

    1987-01-01

    A control system for a trip coil in a switchgear mechanism controls the supply of electrical power to a process control device and ensures de-energization of the trip coil shortly after the trip coil is energized. The trip coil is energized not by an independent dc source as in prior art, but from rectified power from a step down transformer supplied from the switchgear output side. The transformer feeds a rectifier which is connected to the trip coil via a trip activation device. The output of the rectifier can be monitored using an optical converter to determine the ability of the control system to activate the trip coil and the condition of the power supplied to the process control device. The control device may be a rod positioner in a pressurised water nuclear reactor. (author)

  7. Advanced power reactors with improved safety characteristics

    International Nuclear Information System (INIS)

    Birkhofer, A.

    1994-01-01

    The primary objective of nuclear safety is the protection of individuals, society and environment against radiological hazards from accidental releases of radioactive materials contained in nuclear reactors. Hereto, these materials are enclosed by several successive barriers and the barriers protected against mishaps and accidents by a multi-level system of safety precautions. The evolution of reactor technology continuously improves this concept and its implementation. At a world-wide scale, several advanced reactor concepts are currently being considered, some of them already at a design stage. Essential safety objectives include both further strengthening the prevention of accidents and improving the containment of fission products should an accident occur. The proposed solutions differ considerably with regard to technical principles, plant size and time scales considered for industrial application. Two typical approaches can be distinguished: The first approach basically aims at an evolution of power reactors currently in use, taking into account the findings from safety research and from operation of current plants. This approach makes maximum use of proven technology and operating experience but may nevertheless include new safety features. The corresponding designs are often termed 'large evolutionary'. The second approach consists in more fundamental changes compared to present designs, often with strong emphasis on specific passive features protecting the fuel and fuel cladding barriers. Owing to the nature and capability of those passive features such 'innovative designs' are mostly smaller in power output. The paper describes the basic objectives of such developments and illustrates important technical concepts focusing on next generation plants, i.e. designs to be available for industrial application until the end of this decade. 1 tab. (author)

  8. Nuclear safety as applied to space power reactor systems

    International Nuclear Information System (INIS)

    Cummings, G.E.

    1987-01-01

    Current space nuclear power reactor safety issues are discussed with respect to the unique characteristics of these reactors. An approach to achieving adequate safety and a perception of safety is outlined. This approach calls for a carefully conceived safety program which makes uses of lessons learned from previous terrestrial power reactor development programs. This approach includes use of risk analyses, passive safety design features, and analyses/experiments to understand and control off-design conditions. The point is made that some recent accidents concerning terrestrial power reactors do not imply that space power reactors cannot be operated safety

  9. Cobalt-60 production in CANDU power reactors

    International Nuclear Information System (INIS)

    Slack, J.; Norton, J.L.; Malkoske, G.R.

    2003-01-01

    therapy machines. Today the majority of the cancer therapy cobalt-60 sources used in the world are manufactured using material from the NRU reactor in Chalk River. The same technology that was used for producing cobalt-60 in a research reactor was then adapted and transferred for use in a CANDU power reactor. In the early 1970s, in co-operation with Ontario Power Generation (formerly Ontario Hydro), bulk cobalt-60 production was initiated in the four Pickering A CANDU reactors located east of Toronto. This was the first full scale production of millions of curies of cobalt-60 per year. As the demand and acceptance of sterilization of medical products grew, MDS Nordion expanded its bulk supply by installing the proprietary Canadian technology in additional CANDUs. Over the years MDS Nordion has partnered with CANDU reactor owners to produce cobalt-60 at various sites. CANDU reactors that have, or are still producing cobalt-60, include Pickering A, Pickering B, Gentilly 2, Embalse in Argentina, and Bruce B. In conclusion, the technology for cobalt-60 production in CANDU reactors, designed and developed by MDS Nordion and Atomic Energy of Canada, has been safely, economically and successfully employed in CANDU reactors with over 195 reactor years of production. Today over forty percent of the world's disposable medical supplies are made safer through sterilization using cobalt-60 sources from MDS Nordion. Over the past 40 years, MDS Nordion with its CANDU reactor owner partners, has safely and reliably shipped more than 500 million curies of cobalt-60 sources to customers around the world. MDS Nordion is presently adding three more CANDU power reactors to its supply chain. These three additional cobalt producing CANDU's will help supplement the ability of the health care industry to provide safe, sterile, medical disposable products to people around the world. As new applications for cobalt-60 are identified, and the demand for bulk cobalt-60 increases, MDS Nordion and AECL

  10. Nuclear power reactors in the world. April 1990 ed.

    International Nuclear Information System (INIS)

    1990-01-01

    This is the tenth edition of Reference Data Series No. 2, Nuclear Power Reactors in the World, which is published once per year, to present the most recent reactor data available to the Agency. It contains the following summarized information: General information as of the end of 1989 on power reactors operating or under construction, and shut down; Performance data on reactors operating in the Agency's Member States, as reported to the IAEA. The information is collected by the Agency by circulating questionnaires to the Member States through the designated national correspondents. The replies are used to maintain computerized files on general and design data of, and operating experience with, power reactors. The Agency's power reactor information system (PRIS) comprising the above files provides all the information and data previously published in the Agency's Power Reactors in Member States and currently published in the Agency's Operating Experience with Nuclear Power Stations in Member States

  11. Nuclear power reactors in the world. Apr 1991 ed.

    International Nuclear Information System (INIS)

    1991-01-01

    This is the eleventh edition of Reference Data Series No. 2, Nuclear Power Reactors in the World, which is published once per year, to present the most recent reactor data available to the Agency. It contains the following summarized information: General information as of the end of 1990, on power reactors operating or under construction, and shut down; performance data on reactors operating in the Agency's Member States, as reported to the IAEA. This information is collected by the Agency by circulating questionnaires to the Member States through the designated national correspondents. The replies are used to maintain computerized files on general and design data of, and operating experience with, power reactors. The Agency's Power Reactor Information System (PRIS) comprising the above files provides all the information and data previously published in the Agency's Power Reactors in Member States and currently published in the Agency's Operating Experience with Nuclear Power Stations in Member States. 5 figs, 19 tabs

  12. Hydrogen fueling stations in Japan hydrogen and fuel cell demonstration project

    International Nuclear Information System (INIS)

    Koseki, K.; Tomuro, J.; Sato, H.; Maruyama, S.

    2004-01-01

    A new national demonstration project of fuel cell vehicles, which is called Japan Hydrogen and Fuel Cell Demonstration Project (JHFC Project), has started in FY2002 on a four-year plan. In this new project, ten hydrogen fueling stations have been constructed in Tokyo and Kanagawa area in FY2002-2003. The ten stations adopt the following different types of fuel and fueling methods: LPG reforming, methanol reforming, naphtha reforming, desulfurized-gasoline reforming, kerosene reforming, natural gas reforming, water electrolysis, liquid hydrogen, by-product hydrogen, and commercially available cylinder hydrogen. Approximately fifty fuel cell passenger cars and a fuel cell bus are running on public roads using these stations. In addition, two hydrogen stations will be constructed in FY2004 in Aichi prefecture where The 2005 World Exposition (EXPO 2005) will be held. The stations will service eight fuel cell buses used as pick-up buses for visitors. We, Engineering Advancement Association of Japan (ENAA), are commissioned to construct and operate a total of twelve stations by Ministry of Economy Trade and Industry (METI). We are executing to demonstrate or identify the energy-saving effect, reduction of the environmental footprint, and issues for facilitating the acceptance of hydrogen stations on the basis of the data obtained from the operation of the stations. (author)

  13. Fort St. Vrain reactor performance and operation to full power

    International Nuclear Information System (INIS)

    Simon, W.A.; Bramblett, G.C.

    1982-01-01

    The Fort St. Vrain Nuclear Generating Station, powered by a high-temperature gas-cooled reactor (HTGR), has now been tested to full thermal power. Testing was conducted for the dual purposes of demonstrating component and system capability as a part of the rise-to-power program and determining core fluctuation/redistribution behavior under full power conditions. Both objectives were met. Full power performance of all major components and the achievement of nearly all design objectives has been verified. In addition, the tests showed that the fluctuation phenomenon has been corrected. Core region outlet temperature redistributions have been characterized, related to a physical mechanism, and shown to be inconsequential for overall plant operation

  14. Nuclear safety as applied to space power reactor systems

    International Nuclear Information System (INIS)

    Cummings, G.E.

    1987-01-01

    To develop a strategy for incorporating and demonstrating safety, it is necessary to enumerate the unique aspects of space power reactor systems from a safety standpoint. These features must be differentiated from terrestrial nuclear power plants so that our experience can be applied properly. Some ideas can then be developed on how safe designs can be achieved so that they are safe and perceived to be safe by the public. These ideas include operating only after achieving a stable orbit, developing an inherently safe design, ''designing'' in safety from the start and managing the system development (design) so that it is perceived safe. These and other ideas are explored further in this paper

  15. A Review of the Fast Reactor Programme in Japan

    International Nuclear Information System (INIS)

    Matsuno, Y.

    1988-01-01

    This report covers the activities of the experimental fast reactor JOYO from April 1986 to December 1987. Five duty cycles operations from the 10th to 14th were done. Starting from the 13th operation, the duration per cycle was extended to 55 days at 13th cycle and 60 days at 14th, in order to elongate the fuel maximum burn-up from 50,000 MWd/t to 75,000 MWd/t. By the end of 15th cycle (mid-May of '88), the objective 70 days duration is attained. The 6th annual inspection was finished in September 1987. Work for maintenance or modification of the plant has been done on such system as Cecium trap in order to enable the loop to trap the Cecium which will be released from failed fuel during the reactor operation, the cold trap (replacement) of the primary coolant system in order to get higher trapping efficiency and the 2nd set-up (reloading) of on-line irradiation test assembly (INTA-S)

  16. Rise-to-power test in High Temperature Engineering Test Reactor. Test progress and summary of test results up to 30 MW of reactor thermal power

    International Nuclear Information System (INIS)

    Nakagawa, Shigeaki; Fujimoto, Nozomu; Shimakawa, Satoshi

    2002-08-01

    The High Temperature Engineering Test Reactor (HTTR) is a graphite moderated and gas cooled reactor with the thermal power of 30 MW and the reactor outlet coolant temperature of 850degC/950degC. Rise-to-power test in the HTTR was performed from April 23rd to June 6th in 2000 as phase 1 test up to 10 MW in the rated operation mode, from January 29th to March 1st in 2001 as phase 2 test up to 20 MW in the rated operation mode and from April 14th to June 8th in 2001 as phase 3 test up to 20 MW in the high temperature test the mechanism of the reactor outlet coolant temperature becomes 850degC at 30 MW in the rated operation mode and 950degC in the high temperature test operation mode. Phase 4 rise-to-power test to achieve the thermal reactor power of 30 MW started on October 23rd in 2001. On December 7th in 2001 it was confirmed that the thermal reactor power and the reactor outlet coolant temperature reached to 30 MW and 850degC respectively in the single loaded operation mode in which only the primary pressurized water cooler is operating. Phase 4 test was performed until March 6th in 2002. JAERI (Japan Atomic Energy Research Institute) obtained the certificate of the pre-operation test from MEXT (Ministry of Education Culture Sports Science and Technology) after all the pre-operation tests by MEXT were passed successfully with the reactor transient test at an abnormal event as a final pre-operation test. From the test results of the rise-up-power test up to 30 MW in the rated operation mode, performance of the reactor and cooling system were confirmed, and it was also confirmed that an operation of reactor facility can be performed safely. Some problems to be solved were found through the tests. By solving them, the reactor operation with the reactor outlet coolant temperature of 950degC will be achievable. (author)

  17. U.S. Nuclear Power Reactor Plant Status

    Data.gov (United States)

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

  18. Decommissioning of Swedish nuclear power reactors. Technology and costs

    International Nuclear Information System (INIS)

    1994-06-01

    The main topics discussed are planning, technology and costs of decommissioning nuclear power reactors. Oskarshamn-3 (BWR) and Ringhals-4 (PWR) have been used as reference reactors. 29 refs, figs, tabs

  19. power system reliability in supplying nuclear reactors

    International Nuclear Information System (INIS)

    Gad, M.M.M.

    2007-01-01

    this thesis presents a simple technique for deducing minimal cut set (MCS) from the defined minimal path set (MPS) of generic distribution system and this technique have been used to evaluate the basic reliability indices of Egypt's second research reactor (ETRR-2) electrical distribution network. the alternative system configurations are then studied to evaluate their impact on service reliability. the proposed MCS approach considers both sustained and temporary outage. the temporary outage constitutes an important parameter in characterizing the system reliability indices for critical load point in distribution system. it is also consider the power quality impact on the reliability indices

  20. Leaching of nuclear power reactor waste forms

    International Nuclear Information System (INIS)

    Endo, L.S.; Villalobos, J.P.; Miyamoto, H.

    1987-01-01

    The leaching tests for immobilized power reactor wastes carried out at IPEN are described. These wastes forms consist mainly of spent resins and boric acid concentrates solidified in ordinary Portland cement. All tests were conducted according to the ISO and IAEA recommendations. Three years leaching results are reported. The cesium diffuvity coefficients determined out of these results are about 1 x 10 -8 cm 2 /s for boric acid waste form and 9 x 10 -9 cm 2 /s for ion-exchange resin waste. Strontium diffusivity coefficients found are about 3 x 10 -11 cm 2 /s and 9 x 10 -11 cm 2 /s respectively. (Author) [pt

  1. Japan: Sendai, first reactor to restart. Sendai restart: how does it work? Japan: restart will be 'progressive'. 2015: which role for nuclear energy in Japan?

    International Nuclear Information System (INIS)

    Le Ngoc, Boris; Jouette, Isabelle

    2015-01-01

    A set of articles addresses the restart of nuclear plants in Japan. The first one presents the Sendai nuclear plant, evokes the commitment of the Japanese nuclear safety authority (the NRA) at each step of the restart process, the agreement of local populations, the loading of the nuclear fuel, a successful crisis exercise, and the benefits expected from this restart. The second article addresses the restart process with its administrative aspects, the implication of local authorities, its technical aspects, and investments made to improve nuclear safety. The third article proposes an interview of the nuclear expert of the French embassy in Tokyo. He outlines that the restart of nuclear plants will be progressive, comments how Sendai restart has been commented in the Japanese press, evokes how this restart is part of the Japanese Prime Minister's policy, evokes the role and challenges of nuclear energy in Japan for the years to come, and the role France may play. The last article discusses the role of nuclear energy in Japan in 2015: importance of the old 3E policy (Energy, Environment, Economy) which is put into question again by the Fukushima accident, creation of a new nuclear safety authority as a first step before restarting nuclear reactors

  2. Heat Pipe Powered Stirling Conversion for the Demonstration Using Flattop Fission (DUFF) Test

    Science.gov (United States)

    Gibson, Marc A.; Briggs, Maxwell H.; Sanzi, James L.; Brace, Michael H.

    2013-01-01

    Design concepts for small Fission Power Systems (FPS) have shown that heat pipe cooled reactors provide a passive, redundant, and lower mass option to transfer heat from the fuel to the power conversion system, as opposed to pumped loop designs typically associated with larger FPS. Although many systems have been conceptually designed and a few making it to electrically heated testing, none have been coupled to a real nuclear reactor. A demonstration test named DUFF Demonstration Using Flattop Fission, was planned by the Los Alamos National Lab (LANL) to use an existing criticality experiment named Flattop to provide the nuclear heat source. A team from the NASA Glenn Research Center designed, built, and tested a heat pipe and power conversion system to couple to Flattop with the end goal of making electrical power. This paper will focus on the design and testing performed in preparation for the DUFF test.

  3. Introduction to nuclear power reactors and their health physics systems

    International Nuclear Information System (INIS)

    Brtis, J.S.

    1982-01-01

    This paper provides an introduction to: (1) the major systems of Boiling Water Reactors (BWR's) and Pressurized Water Reactors (PWR's), (2) the production and distribution of radiation sources in BWR's and PWR's, (3) the regulatory and functional requirements for nuclear power reactor design from a health physics standpoint, (4) the health physics systems provided to meet such requirements, and (5) a bibliography of documents germane to power reactor health physics design

  4. Development programs on decommissioning technology for reactors and fuel cycle facilities in Japan

    International Nuclear Information System (INIS)

    Fujiki, K.

    1992-01-01

    The Science and Technology Agency (STA) of Japan is promoting technology development for decommissioning of nuclear facilities by entrusting various research programs to concerned research organisations: JAERI, PNC and RANDEC, including first full scale reactor decommissioning of JPDR. According to the results of these programs, significant improvement on dismantling techniques, decontamination, measurement etc. has been achieved. Further development of advanced decommissioning technology has been started in order to achieve reduction of duration of decommissioning work and occupational exposures in consideration of the decommissioning of reactors and fuel cycle facilities. (author) 5 refs.; 7 figs.; 1 tab

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

    Energy Technology Data Exchange (ETDEWEB)

    Buden, D.

    1993-08-11

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

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

    International Nuclear Information System (INIS)

    Buden, D.

    1993-01-01

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

  7. Cascade: a high-efficiency ICF power reactor

    International Nuclear Information System (INIS)

    Pitts, J.H.

    1985-01-01

    Cascade attains a net power-plant efficiency of 49% and its cost is competitive with high-temperature gas-cooled reactor, pressurized-water reactor, and coal-fired power plants. The Cascade reactor and blanket are made of ceramic materials and activation is 6 times less than that of the MARS Tandem Mirror Reactor operating at comparable power. Hands-on maintenance of the heat exchangers is possible one day after shutdown. Essentially all tritium is recovered in the vacuum system, with the remainder recovered from the helium power conversion loop. Tritium leakage external to the vacuum system and power conversion loop is only 0.03 Ci/d

  8. Study on Reactor Performance of Online Power Monitoring in PUSPATI TRIGA Reactor (RTP)

    International Nuclear Information System (INIS)

    Zareen Khan Abdul Jalil Khan; Ridzuan Abdul Mutalib; Mohd Sabri Minhat

    2014-01-01

    The Reactor TRIGA PUSPATI (RTP) at Malaysia Nuclear Agency is a TRIGA Mark II type reactor and pool type cooled by natural circulation of light water. This paper describe on reactor performance of online power monitoring based on various parameter of reactor such as log power, linear power, period, Fuel and coolant temperature and reactivity parameter with using neutronic and other instrumentation system of reactor. Methodology of online power estimation and monitoring is to evaluate and analysis of reactor power which is important of reactor safety and control. Neutronic instrumentation system will use to estimate power measurement, differential of log and linear power and period during reactor operation .This study also focus on noise fluctuation from fission chamber during reactor operation .This work will present result of online power monitoring from RTP which indicated the safety parameter identification and initiate safety action on crossing the threshold set point trip. Conclude that optimization of online power monitoring will improved the reactor control and safety parameter of reactor during operation. (author)

  9. Financing the electric power utilities, especially the nuclear power in Japan

    International Nuclear Information System (INIS)

    Tajima, T.

    1975-04-01

    Electric power demands in Japan have shown a remarkable growth at an annual rate of 12% since 1965. Nine electric power companies have invested large amounts of money so far, amounting to over 1 trillion yen every year since 1972. A survey of the electric power supply system and an estimation of the electric power demands in 1980 and in 1985 are given. It is expected that the main portion of electric power in the future will gradually be generated by nuclear plants. Financial features of the electrical power utilities, the credit risk of the electric power utilities, and the raising of funds by electric power utilities are discussed. It is concluded that it will be necessary (1) to expand the capital market, (2) to enable the electric power companies to issue a sufficient amount of bonds, (3) to make the Government financing institutions, such as the Japan Development Bank, provide the electric power companies with larger funds on a long-term and low-interest rate basis, and (4) even to take such drastic steps as subsidizing interest on private loans to the electric power companies. (B.P.)

  10. Revision of the second basic plans of power reactor development in Power Reactor and Nuclear Fuel Development Corporation

    International Nuclear Information System (INIS)

    1978-01-01

    Revision of the second basic plans concerning power reactor development in PNC (Power Reactor and Nuclear Fuel Development Corporation) is presented. (1) Fast breeder reactors: As for the experimental fast breeder reactor, after reaching the criticality, the power is raised to 50 MW thermal output within fiscal 1978. The prototype fast breeder reactor is intended for the electric output of 200 MW -- 300 MW, using mixed plutonium/uranium oxide fuel. Along the above lines, research and development will be carried out on reactor physics, sodium technology, machinery and parts, nuclear fuel, etc. (2) Advanced thermal reactor: The prototype advanced thermal reactor, with initial fuel primarily of slightly enriched uranium and heavy water moderation and boiling water cooling, of 165 MW electric output, is brought to its normal operation by the end of fiscal 1978. Along the above lines, research and development will be carried out on reactor physics, machinery and parts, nuclear fuel, etc. (Mori, K

  11. 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)

  12. 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.

  13. 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

  14. ATMEA and medium power reactors. The ATMEA joint venture and the ATMEA1 medium power reactor

    International Nuclear Information System (INIS)

    Mathet, Eric; Castello, Gerard

    2012-01-01

    This Power Point presentation presents the ATMEA company (a joint venture of Areva and Mitsubishi), the main features of its medium power reactor (ATMEA1) and its building arrangement, indicates the general safety objectives. It outlines the features of its robust design which aim at protecting, cooling down and containing. It indicates the regulatory and safety frameworks, comments the review of the safety options by the ASN and the results of this assessment

  15. Neutron measurements at nuclear power reactors [55

    CERN Document Server

    Scherpelz, R I

    2002-01-01

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

  16. 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

  17. Utilization of the research reactors for the power reactor control instrumentation development

    International Nuclear Information System (INIS)

    Duchene, J.; Verdant, R.; Gilbert, J.

    1977-01-01

    Studies on characteristics and reliability of control instruments lead to testing with various radiations of various intensities and energy spectra. Osiris and Triton reactors present this great variety of radiations and a flexibility of use better than power reactors [fr

  18. UF6 breeder reactor power plants for electric power generation

    International Nuclear Information System (INIS)

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

    1976-01-01

    The reactor concept analyzed is a 233 UF 6 core surrounded by a molten salt (Li 7 F, BeF 2 , ThF 4 ) blanket. Nuclear survey calculations were carried out for both spherical and cylindrical geometries. A maximum breeding ratio of 1.22 was found. Thermodynamic cycle calculations were performed for a variety of Rankine cycles. Optimization of a Rankine cycle for a gas core breeder reactor employing an intermediate heat exchanger gave a maximum efficiency of 37 percent. A conceptual design is presented along with a system layout for a 1000 MW stationary power plant. The advantages of the GCBR are as follows: (1) high efficiency, (2) simplified on-line reprocessing, (3) inherent safety considerations, (4) high breeding ratio, (5) possibility of burning all or most of the long-lived nuclear waste actinides, and (6) possibility of extrapolating the technology to higher temperatures and MHD direct conversion

  19. 78 FR 64028 - Decommissioning of Nuclear Power Reactors

    Science.gov (United States)

    2013-10-25

    ... NUCLEAR REGULATORY COMMISSION [NRC-2012-0035] Decommissioning of Nuclear Power Reactors AGENCY... the NRC's regulations relating to the decommissioning process for nuclear power reactors. The revision... Commission (NRC) is issuing Revision 1 of regulatory guide (RG) 1.184 ``Decommissioning of Nuclear Power...

  20. 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.

  1. Tokamak power reactor ignition and time dependent fractional power operation

    International Nuclear Information System (INIS)

    Vold, E.L.; Mau, T.K.; Conn, R.W.

    1986-06-01

    A flexible time-dependent and zero-dimensional plasma burn code with radial profiles was developed and employed to study the fractional power operation and the thermal burn control options for an INTOR-sized tokamak reactor. The code includes alpha thermalization and a time-dependent transport loss which can be represented by any one of several currently popular scaling laws for energy confinement time. Ignition parameters were found to vary widely in density-temperature (n-T) space for the range of scaling laws examined. Critical ignition issues were found to include the extent of confinement time degradation by alpha heating, the ratio of ion to electron transport power loss, and effect of auxiliary heating on confinement. Feedback control of the auxiliary power and ion fuel sources are shown to provide thermal stability near the ignition curve

  2. Statistic method of research reactors maximum permissible power calculation

    International Nuclear Information System (INIS)

    Grosheva, N.A.; Kirsanov, G.A.; Konoplev, K.A.; Chmshkyan, D.V.

    1998-01-01

    The technique for calculating maximum permissible power of a research reactor at which the probability of the thermal-process accident does not exceed the specified value, is presented. The statistical method is used for the calculations. It is regarded that the determining function related to the reactor safety is the known function of the reactor power and many statistically independent values which list includes the reactor process parameters, geometrical characteristics of the reactor core and fuel elements, as well as random factors connected with the reactor specific features. Heat flux density or temperature is taken as a limiting factor. The program realization of the method discussed is briefly described. The results of calculating the PIK reactor margin coefficients for different probabilities of the thermal-process accident are considered as an example. It is shown that the probability of an accident with fuel element melting in hot zone is lower than 10 -8 1 per year for the reactor rated power [ru

  3. Power generation costs for alternate reactor fuel cycles

    International Nuclear Information System (INIS)

    Smolen, G.R.; Delene, J.G.

    1980-09-01

    The total electric generating costs at the power plant busbar are estimated for various nuclear reactor fuel cycles which may be considered for power generation in the future. The reactor systems include pressurized water reactors (PWR), heavy-water reactors (HWR), high-temperature gas cooled reactors (HTGR), liquid-metal fast breeder reactors (LMFBR), light-water pre-breeder and breeder reactors (LWPR, LWBR), and a fast mixed spectrum reactor (FMSR). Fuel cycles include once-through, uranium-only recycle, and full recycle of the uranium and plutonium in the spent fuel assemblies. The U 3 O 8 price for economic transition from once-through LWR fuel cycles to both PWR recycle and LMFBR systems is estimated. Electric power generation costs were determined both for a reference set of unit cost parameters and for a range of uncertainty in these parameters. In addition, cost sensitivity parameters are provided so that independent estimations can be made for alternate cost assumptions

  4. An overview of in-service inspection for nuclear power plants in Japan

    International Nuclear Information System (INIS)

    Takano, Eishiro

    1996-01-01

    In nuclear power stations, for maintaining the safety of machinery and equipment, it is stipulated to carry out periodic inspection. The contents of the periodic inspection and the inspections of important matters are explained. In-service inspection (ISI) is the nondestructive inspection for confirming the soundness of reactor pressure vessels, main pipings, valves and others. The stipulations on the scope and time of the periodic inspection and ISI, the course of introducing ISI into Japan, and the basic way of thinking on ISI are described. Pre-service inspection (PSI) is carried out for the purpose of collecting and recording the initial data for ISI. The range of the objects of nondestructive inspection and leak test and the range of the objects of leak test only are shown. As to nondestructive inspection, the parts to be inspected, the frequency of inspection and the methods of inspection, and the method of leak test are described. As the present state of inspection technology, the automatic ultrasonic flow detectors for reactor vessels, piping welded parts and reactor vessel stud bolts, and the eddy current flaw detector for steam generator tubes are explained. (K.I.)

  5. Design of PCCV for Tsuruga No.2 plant of Japan Atomic Power Co

    International Nuclear Information System (INIS)

    Toda, Shiro; Yamada, Yasushi; Kuno, Toshio.

    1987-01-01

    The reactor containment vessel for Tsuruga No.2 plant is a prestressed concrete structure, the design of which was carried out in USA for three years from August, 1977. The research on PCCVs has been carried out for more than ten years in Japan, but this is the first case adopting it for an actual plant, so the design was advanced by the close cooperation between both Japanese and USA sides so as to conform to Japanese laws and tradition, and the check and review of the design results were jointly carried out. In reactor containment vessels, the internal pressure and temperature rise occurring at the time of LOCA become the main design load. The features of PCCVs as prestressed concrete structures are the utilization of prestress for large scale membrane structures, the vessels are subjected to thermal stress since those contain nuclear reactors, and the sufficient safety to the severe aseismatic condition peculiar to nuclear power stations. The prestressed concrete structure achieves the function of pressure vessels, the steel liner achieves the function of leak prevention, and the concrete achieves the function of radiation shield, and these are combined in one body as PCCVs. A PCCV is composed of a prestressed concrete upper shell and a reinforced concrete foundation. The structure, the standards to be applied, the allowable values of load and materials, the stress analysis, the cross section design and the design reinforcement are reported. (Kako, I.)

  6. How power is generated in a nuclear reactor

    International Nuclear Information System (INIS)

    Swaminathan, V.

    1978-01-01

    Power generation by nuclear fission as a result of chain reaction caused by neutrons interacting with fissile material such as 235 U, 233 U and 239 Pu is explained. Electric power production by reactor is schematically illustrated. Materials used in thermal reactor and breeder reactor are compared. Fuel reprocessing and disposal of radioactive waste coming from reprocessing plant is briefly described. Nuclear activities in India are reviewed. Four heavy water plants and two power reactors are under construction and will be operative in the near future. Two power reactors are already in operation. Nuclear Fuel Complex at Hyderabad supplies fuel element to the reactors. Fuel reprocessing and waste management facility has been set up at Tarapur. Bhabha Atomic Research Centre at Bombay and Reactor Research Centre at Kalpakkam near Madras are engaged in applied and basic research in nuclear science and engineering. (B.G.W.)

  7. The concept of the innovative power reactor

    Directory of Open Access Journals (Sweden)

    Sang Won Lee

    2017-10-01

    Full Text Available The Fukushima accident reveals the vulnerability of existing active nuclear power plant (NPP design against prolonged loss of external electricity events. The passive safety system is considered an attractive alternative to cope with this kind of disaster. Also, the passive safety system enhances both the safety and the economics of NPPs. The adoption of a passive safety system reduces the number of active components and can minimize the construction cost of NPPs. In this paper, reflecting on the experience during the development of the APR+ design in Korea, we propose the concept of an innovative Power Reactor (iPower, which is a kind of passive NPP, to enhance safety in a revolutionary manner. The ultimate goal of iPower is to confirm the feasibility of practically eliminating radioactive material release to the environment in all accident conditions. The representative safety grade passive system includes a passive emergency core cooling system, a passive containment cooling system, and a passive auxiliary feedwater system. Preliminary analysis results show that these concepts are feasible with respect to preventing and/or mitigating the consequences of design base accidents and severe accidents.

  8. Experience in using a research reactor for the training of power reactor operators

    International Nuclear Information System (INIS)

    Blotcky, A.J.; Arsenaut, L.J.

    1972-01-01

    A research reactor facility such as the one at the Omaha Veterans Administration Hospital would have much to offer in the way of training reactor operators. Although most of the candidates for the course had either received previous training in the Westinghouse Reactor Operator Training Program, had operated nuclear submarine reactors or had operated power reactors, they were not offered the opportunity to perform the extensive manipulations of a reactor that a small research facility will allow. In addition the AEC recommends 10 research reactor startups per student as a prerequisite for a cold operator?s license and these can easily be obtained during the training period

  9. The future of nuclear power and fourth-generation reactors

    International Nuclear Information System (INIS)

    Carre, F.; Renault, C.

    2006-01-01

    Faced with the exhaustion of fossil fuel resources, the output of existing nuclear power must quadruple between now and 2050, and the Commissariat a l'Energie atomique (CEA) and its industrial partners are cooperating in a programme of R and D on future nuclear power. France strategy puts rapid neutron reactors (RNR) at the forefront, in view of their possible introduction by 2040. These reactors allow a more efficient use of uranium resources and minimise the production of long-life nuclear waste. Two technologies which use respectively, sodium and gas as their coolant are being studied. For the sodium RNR, which benefits from significant existing experience, the key is to first improve its economic performance. For the gas RNR, which draws on the principles and the generic assets of the RNR, for those using helium as the coolant, and those with applications at high temperature, it is important firstly to demonstrate the key technologies such as the fuel. The decision of President Chirac to launch the study of a prototype, fourth-generation reactor for 2020 is stimulating the research effort into France future nuclear power. (author)

  10. Nuclear Power: Outlook for New U.S. Reactors

    National Research Council Canada - National Science Library

    Parker, Larry; Holt, Mark

    2007-01-01

    .... The renewed interest in nuclear power has resulted primarily from higher prices for natural gas, improved operation of existing reactors, and uncertainty about future restrictions on coal emissions...

  11. Management of radioactive wastes at power reactor sites in India

    International Nuclear Information System (INIS)

    Amalraj, R.V.; Balu, K.

    Indian nuclear power programme, at the present stage, is based on natural uranium fuelled heavy water moderated CANDU type reactors except for the first nuclear power station consisting of two units of enriched uranium fuelled, light water moderated, BWR type of reactors. Some of the salient aspects of radioactive waste management at power reactor sites in India are discussed. Brief reviews are presented on treatment of wastes, their disposal and environmental aspects. Indian experience in power reactor waste management is also summarised identifying some of the areas needing further work. (auth.)

  12. Instrumentation and control for reactor power setback in PFBR

    International Nuclear Information System (INIS)

    Upadhyay, Chandra Kant; Vasal, Tanmay; Nagaraj, C.P.; Madhusoodanan, K.

    2013-01-01

    In Prototype Fast Breeder Reactor (PFBR), a 500 MWe plant, Reactor Power Setback is a special operation envisaged for bulk power reduction on occurrence of certain events in Balance of Plant. The bulk power reduction requires a large negative reactivity perturbation if reactor is operating on nominal power. This necessitates a reliable monitoring system with fault tolerant I and C architecture in order to inhibit reactor SCRAM on negative reactivity trip signal. The impact of above events on the process is described. Design of a functional prototype module to carry out RPSB logic operation and its interface with other instruments has been discussed. (author)

  13. Technical Evaluation on Ageing Management in Nuclear Power in Japan -Life Extension over 40 Years at Japanese Nuclear Power Plants-

    International Nuclear Information System (INIS)

    Tanaka, H.; Iwasaki, M.; Miyama, S.

    2012-01-01

    The Japanese commercial reactors have accumulated more than 40 years operating experience since Tsuruga-1 started commercial operation in March 1970. However, the severe accident occurred at Fukushima nuclear power plants triggered by East Japan Great Earthquake on March 11 last year. The facts that all the reactors having experienced core meltdown had operated for more than 30 years and Fukushima Daiichi unit 1 had just received the approval for its 40th year Ageing Management Technical Evaluation results from the Japanese government caused increasing distrust among the public in nuclear power plants operating for a long period of time. However, investigations of the accident conducted so far have not revealed any evidence that ageing degradation accelerated the accident. In addition, the analysis of seismic accelerations of the earthquake did not show that any component function was lost due to the accident. Considering these facts, I would like to discuss the issues to be continuously pursued and to be additionally implemented as part of the plat life management activities. In addition, I will introduce the efforts made by the Japanese utilities following the accident. (author)

  14. Pellet bed reactor for multi-modal space power

    International Nuclear Information System (INIS)

    Buden, D.; Williams, K.; Mast, P.; Mims, J.

    1987-01-01

    A review of forthcoming space power needs for both civil and military missions indicates that power requirements will be in the tens of megawatts. The electrical power requirements are envisioned to be twofold: long-duration lower power levels will be needed for station keeping, communications, and/or surveillance; short-duration higher power levels will be required for pulsed power devices. These power characteristics led to the proposal of a multi-modal space power reactor using a pellet bed design. Characteristics desired for such a multimegawatt reactor power source are standby, alert, and pulsed power modes; high-thermal output heat source (approximately 1000 MWt peak power); long lifetime station keeping power (10 to 30 years); high temperature output (1500 K to 1800 K); rapid-burst power transition; high reliability (above 95 percent); and stringent safety standards compliance. The proposed pellet bed reactor is designed to satisfy these characteristics

  15. Enabling autonomous control for space reactor power systems

    International Nuclear Information System (INIS)

    Wood, R. T.

    2006-01-01

    The application of nuclear reactors for space power and/or propulsion presents some unique challenges regarding the operations and control of the power system. Terrestrial nuclear reactors employ varying degrees of human control and decision-making for operations and benefit from periodic human interaction for maintenance. In contrast, the control system of a space reactor power system (SRPS) employed for deep space missions must be able to accommodate unattended operations due to communications delays and periods of planetary occlusion while adapting to evolving or degraded conditions with no opportunity for repair or refurbishment. Thus, a SRPS control system must provide for operational autonomy. Oak Ridge National Laboratory (ORNL) has conducted an investigation of the state of the technology for autonomous control to determine the experience base in the nuclear power application domain, both for space and terrestrial use. It was found that control systems with varying levels of autonomy have been employed in robotic, transportation, spacecraft, and manufacturing applications. However, autonomous control has not been implemented for an operating terrestrial nuclear power plant nor has there been any experience beyond automating simple control loops for space reactors. Current automated control technologies for nuclear power plants are reasonably mature, and basic control for a SRPS is clearly feasible under optimum circumstances. However, autonomous control is primarily intended to account for the non optimum circumstances when degradation, failure, and other off-normal events challenge the performance of the reactor and near-term human intervention is not possible. Thus, the development and demonstration of autonomous control capabilities for the specific domain of space nuclear power operations is needed. This paper will discuss the findings of the ORNL study and provide a description of the concept of autonomy, its key characteristics, and a prospective

  16. Aging countermeasures for nuclear power plants in Japan and PLEC's activities

    International Nuclear Information System (INIS)

    Tajima, K.; Ishikawa, M.; Koyama, M.

    2002-01-01

    Full text: This summarizes aging countermeasure program of the nuclear power plants performed by the Japanese Government and industries, and describes current R and D program and utilization of the research results for the aged plants. The regulatory bodies (NISA of METI: the Nuclear Industry Safety Agency of the Ministry of Economy, Trade and Industry, MITI was reorganized to the Ministry of Economy, Trade and Industry (METI) in 2001) reviewed the results of technical assessments of aging phenomena of structures and components of nuclear power plants, conducted by the electric utilities. Major technical research projects related to aging of nuclear power plants in Japan have been conducted by Japan Power Engineering and Inspection Corporation (JAPEIC) under the asepses of the regulatory bodies. The results of the research projects were applied to the review and technical assessments. Environmental Fatigue Tests of Nuclear Power Plants Materials for Reliability Verification Project (EFT Project), has been conducting the fatigue tests for carbon steels, low alloy steels and austenitic steels under the simulated light water reactor (LWR) water environmental conditions. Based on the results, EFT project developed the models for evaluating fatigue initiation life reduction in LWR environment. The models were applied to the fatigue evaluation in the technical assessments mentioned above. On the other hand, the Nuclear Power Plant Life Engineering Center (PLEC) entrusted by NISA is conducting to coordinate planning of technology research on aging, to integrate technologies and knowledge of aging for management and assessment, to promote practical applications of results of technical research, and to promote sharing of information on aging technology. As to codes and standards the PLEC plans to develop them, based on the research results of the national projects performed by JAPEIC, with completion of the projects in order. Current planning technical standard is based to

  17. Cooperation ability of Japan to China in nuclear power industries. Present status and future

    International Nuclear Information System (INIS)

    Murakami, Tomoko

    2006-01-01

    Japan is superior to China in the field of LWR plant operation and maintenance, FBR cycle included operation and control of reactor and reprocessing facility, and measures of safeguards and non-proliferation of all commercial nuclear power facilities from the point of view that Japanese technologies are better than the other countries and China needs the technologies. It is important that Japanese electric power companies, plant makers, fuel industries and research organizations developed their business in China in the above fields on the basis of their knowledge, strategies and/or trough network of negotiation of two governments such as forum for nuclear cooperation in Asia (FNCA)·Generation IV International Forum (GIF), and World Association of Nuclear Operators (WANO)·World Nuclear Association (WNA). Outline of finding new market and technical cooperation in the industry and future of nuclear power industry in China are stated. As the supplementary materials, table of operating, building and planning nuclear power plants, estimation of demand for uranium enrichment on the basis of estimation and plans of expansion of power plant facilities, and results of calculation of Separative Work Unit (SWU) from demand for uranium are illustrated. (S.Y.)

  18. Feasibility study of full-reactor gas core demonstration test

    Science.gov (United States)

    Kunze, J. F.; Lofthouse, J. H.; Shaffer, C. J.; Macbeth, P. J.

    1973-01-01

    Separate studies of nuclear criticality, flow patterns, and thermodynamics for the gas core reactor concept have all given positive indications of its feasibility. However, before serious design for a full scale gas core application can be made, feasibility must be shown for operation with full interaction of the nuclear, thermal, and hydraulic effects. A minimum sized, and hence minimum expense, test arrangement is considered for a full gas core configuration. It is shown that the hydrogen coolant scattering effects dominate the nuclear considerations at elevated temperatures. A cavity diameter of somewhat larger than 4 ft (122 cm) will be needed if temperatures high enough to vaporize uranium are to be achieved.

  19. Radiation shielding provided by residential houses in Japan in reactor accidents accompanied with atmospheric release

    International Nuclear Information System (INIS)

    Yamaguchi, Yasuhiro; Minami, Kentaro

    1991-01-01

    The present report describes the radiation shielding effect of houses in Japan against the radioactive cloud resulting from a major reactor accident accompanied with atmospheric release. The shielding factor of houses, the ratio of indoor exposure rate to outdoor one, has been studied for the semi-infinite and finite clouds which contain γ-emitting radionuclides released from a reactor facility. The shielding factor of houses against γ-rays from the radioactive cloud decreases gradually with release delay time and keeps a minimum during the period from 50 to 1000 hours after reactor shutdown while 133 Xe predominates in the cloud. Radioiodines mixed in the cloud raise slightly the shielding factor, and the factor depends little on the shape of the cloud. A set of shielding factors for the use of emergency planning was consequently proposed as 0.4 for simple ferroconcrete residential house and 0.9 for other ordinary ones. (author)

  20. JENDL reactor constant and its application. The 40th anniversary of Japan nuclear data committee

    International Nuclear Information System (INIS)

    Zukeran, Atsushi

    2004-01-01

    The status of reactor constants about 27 years ago is briefly reviewed from the criticality predictions and nuclear data processing codes. In the second section, status of current users of JENDL-3.3 and/or JENDL-3.2 is consulted with the 2003 Fall Meeting of Atomic Energy Society of Japan. In the third section, the reliabilities of JENDL-3.3 and -3.2 are reviewed mainly from the application to light water reactor (LWR) mockup experiments; MISTRAL and BASALA made on EOLE critical facility of Cadarache Laboratory in France, since an extensive evaluation for nuclear data applicability to LWR have been scarcely performed in relative to for FBR. The results of international benchmark cores and criticality safety analyses are briefly reviewed. In the concluding remarks, overall applicability is shown as a summary with respect to all reactor parameters obtained in the LWR mockup experiment and some remarks are noted. (author)