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Sample records for frg-1 research reactor

  1. Upgrading of the research reactors FRG-1 and FRG-2

    International Nuclear Information System (INIS)

    Krull, W.

    1981-01-01

    In 1972 for the research reactor FRG-2 we applied for a license to increase the power from 15 MW to 21 MW. During this procedure a public laying out of the safety report and an upgrading procedure for both research reactors - FRG-1 (5 MW) and FRG-2 - were required by the licensing authorities. After discussing the legal background for licensing procedures in the Federal Republic of Germany the upgrading for both research reactors is described. The present status and future licensing aspects for changes of our research reactors are discussed, too. (orig.) [de

  2. Continuous backfitting measures for the FRG-1 and FRG-2 research reactors

    International Nuclear Information System (INIS)

    Blom, K.H.; Falck, K.; Krull, W.

    1990-01-01

    The GKSS-Research Centre Geesthacht GmbH has been operating the research reactors FRG-1 and FRG-2 with power levels of 5 MW and 15 MW for 31 and 26 years respectively. Safe operation at full power levels over so many years with an average utilization between 180 d to 250 d per year is possible only with great efforts in modernization and upgrading of the research reactors. Emphasis has been placed on backfitting since around 1975. At that time within the Federal Republic of Germany many new guidelines, rules, ordinances, and standards in the field of (power) reactor safety were published. Much work has been done on the modernization of the FRG-1 and FRG-2 research reactors therefore within the last ten years. Work done within the last two years and presently underway includes: measures against water leakage through the concrete and along the beam tubes; repair of both cooling towers; modernization of the ventilation system; measures for fire protection; activities in water chemistry and water quality; installation of a double tubing for parts of the primary piping of the FRG-1; replacement of instrumentation, process control systems (operation and monitoring system) and alarm system; renewal of the emergency power supply; installation of internal lightning protection; installation of a cold neutron source; enrichment reduction for FRG-1. These efforts will continue to allow safe operation of our research reactors over their whole operational life

  3. Quality assurance measures at the Geesthacht research reactor FRG-1

    International Nuclear Information System (INIS)

    Voss, J.; Krull, W.; Schmidt, K.

    1995-01-01

    The major part of the quality system for the FRG is already in practical use; other parts require extensive preparations and therefore transition periods of different lengths before they can be introduced. GKSS is aware that beside the other upgrading and refurbishment activities, these duality assurance measures will be very important in ensuring the operation of the FRG-1 research reactor over the coming 15 years or more. (orig.)

  4. Compilation of backfitting measures for the FRG-1 and FRG-2 research reactors

    International Nuclear Information System (INIS)

    Krull, W.

    1989-01-01

    The efforts for the modernization of the FRG-reactors within the last two years and at present are: Measures against water leakage through the concrete and along beam tubes, repair of both cooling towers, modernization of the ventilation system, measures for fire protection, activities in water chemistry and water quality, installation of a double tubing for parts of the primary piping of the FRG-1, replacement of instrumentation, process control system (operation and monitoring system) and alarm system, installation of a cold neutron source, enrichment reduction for the FRG-1. Planned activities are: Renewal of the emergency power supply, installation for internal lightning protection, compressed air system. (orig.) With 26 figs., 1 tab [de

  5. Backfitting of the FRG reactors

    Energy Technology Data Exchange (ETDEWEB)

    Krull, W [GKSS-Forschungszentrum Geesthacht GmbH, Geesthacht (Germany)

    1990-05-01

    The FRG-research reactors The GKSS-research centre is operating two research reactors of the pool type fueled with MTR-type type fuel elements. The research reactors FRG-1 and FRG-2 having power levels of 5 MW and 15 MW are in operation for 31 year and 27 years respectively. They are comparably old like other research reactors. The reactors are operating at present at approximately 180 days (FRG-1) and between 210 and 250 days (FRG-2) per year. Both reactors are located in the same reactor hall in a connecting pool system. Backfitting measures are needed for our and other research reactors to ensure a high level of safety and availability. The main backfitting activities during last ten years were concerned with: comparison of the existing design with today demands (criteria, guidelines, standards etc.); and probability approach for events from outside like aeroplane crashes and earthquakes; the main accidents were rediscussed like startup from low and full power, loss of coolant flow, loss of heat sink, loss of coolant and fuel plate melting; a new reactor protection system had to be installed, following today's demands; a new crane has been installed in the reactor hall. A cold neutron source has been installed to increase the flux of cold neutrons by a factor of 14. The FRG-l is being converted from 93% enriched U with Alx fuel to 20% enriched U with U{sub 3}Si{sub 2} fuel. Both cooling towers were repaired. Replacement of instrumentation is planned.

  6. Backfitting of the FRG reactors

    International Nuclear Information System (INIS)

    Krull, W.

    1990-01-01

    The FRG-research reactors The GKSS-research centre is operating two research reactors of the pool type fueled with MTR-type type fuel elements. The research reactors FRG-1 and FRG-2 having power levels of 5 MW and 15 MW are in operation for 31 year and 27 years respectively. They are comparably old like other research reactors. The reactors are operating at present at approximately 180 days (FRG-1) and between 210 and 250 days (FRG-2) per year. Both reactors are located in the same reactor hall in a connecting pool system. Backfitting measures are needed for our and other research reactors to ensure a high level of safety and availability. The main backfitting activities during last ten years were concerned with: comparison of the existing design with today demands (criteria, guidelines, standards etc.); and probability approach for events from outside like aeroplane crashes and earthquakes; the main accidents were rediscussed like startup from low and full power, loss of coolant flow, loss of heat sink, loss of coolant and fuel plate melting; a new reactor protection system had to be installed, following today's demands; a new crane has been installed in the reactor hall. A cold neutron source has been installed to increase the flux of cold neutrons by a factor of 14. The FRG-l is being converted from 93% enriched U with Alx fuel to 20% enriched U with U 3 Si 2 fuel. Both cooling towers were repaired. Replacement of instrumentation is planned

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

    International Nuclear Information System (INIS)

    Krull, W.

    1980-01-01

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

  8. FRG-1: new millenium - new compact core

    International Nuclear Information System (INIS)

    Schreiner, P.; Knop, W.

    2001-01-01

    The GKSS research center Geesthacht GmbH operates the MTR-type swimming pool research reactor FRG-1 (5 MW) for more than 40 years. The FRG-1 has been converted in February 1991 from HEU (93 %) to LEU (20 %) in one step and at that time the core size was reduced from 49 to 26 fuel elements. Consequently the thermal neutron flux in beam tube positions could be increased by more than a factor of two. It is the strong intention of GKSS to continue the operation of the FRG-1 research reactor for at least an additional 15 years with high availability and utilization. The reactor has been operated during the last years for approximately 250 full power days per year. To prepare the FRG-1 for an efficient future use, the core size has been reduced in a second step from 26 fuel elements to 12 fuel elements. (author)

  9. Main refurbishment activities on electronic and electrical equipment for the FRG-1 research reactor

    International Nuclear Information System (INIS)

    Blom, K.H.; Krull, W.

    1997-01-01

    As GKSS intends to operate the research reactor FRG-1 safely and reliably for many years to come, the plant is constantly refurbished and upgraded both in the interests of safety and operational reasons. The following electronic and electrical systems have been replaced or improved since 1990: Information and signalling systems; Emergency power plant (permit applied for); External and internal lightning protection system; Reactor protection system (in part); Safety lighting; Alarm and staff locating system; Control room telephone system; Closed-circuit television system; Beam tube controls; Storage plant for radioactive liquid waste; Ambient dose rate measuring system; Meteorological measuring system; Control and measuring system for the primary cooling circuit; Control rod drives; Control rod control system; Soft start for the secondary pumps; Control and switching devices for the emergency power plant; Trailing cable installation for the reactor bridge; Main-voltage distribution systems/cable routes. (author). 13 figs, 1 tab

  10. FRG compact core - one year experience

    International Nuclear Information System (INIS)

    Knop, W.; Schreiner, P.

    2001-01-01

    The GKSS research centre Geesthacht GmbH operates the MTR-type swimming pool reactor FRG-1 (5 MW) for more than 40 years. The FRG-1 has been upgraded and refurbished many times to follow the changing demands of safe operation and today's needs of high neutron flux for scientific research. High neutron fluxes with highest availability is the permanent demand of the science on the operation of a neutron source. (orig.)

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

    International Nuclear Information System (INIS)

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

    1981-11-01

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

  12. New cooling system of the FRG-1 two barrier system of the primary coolant cycle

    International Nuclear Information System (INIS)

    Knop, W.; Schreiner, P.

    2003-01-01

    The GKSS research center operates the swimming pool reactor FRG-1 with a thermal power of 5 MW as national neutron source for neutron scattering experiments and sample irradiation as well. Before changing the primary coolant cycle consisted of the reactor core and the closed piping including pumps, heat exchanger and delay tank. The closed cooling circuit was located underneath the reactor pool, in the so-called radioactive cellar. This piping system served secondary coolant system. Due to the location of the primary coolant cycle below the operation pool a postulated 2-F line break and simultaneous failure of the pool slide gate valve could lead to a falling dry of the total reactor core. the new primary coolant system was built in the beginning 2002 in a partitioned cell all within the radioactive cellar, so that the reactor core remains with water with the assumed incident. Due to the new two barrier-inclusion of the primary circuit only the melting of two fuel plates (from total 252 fuel plates) has to be taken into account. This measure and the core compactness in 2000 with a neutron flux gain of a factor of 2 makes the FRG-1 ready for the next 15 years of reactor operation. (author)

  13. FRG activities in the framework of IWGATWR

    International Nuclear Information System (INIS)

    Meyer, P.J.

    1989-01-01

    In the FRG a total capacity of 21,470 MW of nuclear power stations are in operation (May 88). One remaining convoy plant with a capacity of 1,300 MW is under construction and will start its commercial operation in 1989. The SNR-300 fast breeder reactor is still not allowed for political reasons to start operation. With the mentioned capacity the nuclear power plant programme in the FRG is more or less finalized. However, in the 90s further nuclear power plants are expected to be ordered to start operation at the end of the century. The entire situation for new nuclear power plants in the FRG is at the moment unclear. Nevertheless a number of projects are being performed in the field of advanced reactor technology in the FRG. A short description of some of these programs is given in this paper. 4 figs

  14. Theoretical investigations on the determination of cavitation-free primary flow after compacting the reactor FRG-1

    International Nuclear Information System (INIS)

    Pihowicz, W.

    1997-01-01

    Basing on an extensive analysis of the influence of compacting the FRG-1 reactor core upon the change in flow, heat exchange, as well as cavitation behavior of the primary circuit the underlying principles of the determination procedure for a cavitations-free primary flow have been developed theoretically. It was found that the problem has to be treated in a complex manner, i.e. considering the coupled flow, temperature, as well as cavitation fields, and that for a successful solution of the problem it is absolutely necessary to simultaneously induce an optimized fixation of the main primary coolant pump. (orig.) [de

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

    International Nuclear Information System (INIS)

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

    1980-11-01

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

  16. FHL1 reduces dystrophy in transgenic mice overexpressing FSHD muscular dystrophy region gene 1 (FRG1.

    Directory of Open Access Journals (Sweden)

    Sandra J Feeney

    Full Text Available Facioscapulohumeral muscular dystrophy (FSHD is an autosomal-dominant disease with no effective treatment. The genetic cause of FSHD is complex and the primary pathogenic insult underlying the muscle disease is unknown. Several disease candidate genes have been proposed including DUX4 and FRG1. Expression analysis studies of FSHD report the deregulation of genes which mediate myoblast differentiation and fusion. Transgenic mice overexpressing FRG1 recapitulate the FSHD muscular dystrophy phenotype. Our current study selectively examines how increased expression of FRG1 may contribute to myoblast differentiation defects. We generated stable C2C12 cell lines overexpressing FRG1, which exhibited a myoblast fusion defect upon differentiation. To determine if myoblast fusion defects contribute to the FRG1 mouse dystrophic phenotype, this strain was crossed with skeletal muscle specific FHL1-transgenic mice. We previously reported that FHL1 promotes myoblast fusion in vitro and FHL1-transgenic mice develop skeletal muscle hypertrophy. In the current study, FRG1 mice overexpressing FHL1 showed an improvement in the dystrophic phenotype, including a reduced spinal kyphosis, increased muscle mass and myofiber size, and decreased muscle fibrosis. FHL1 expression in FRG1 mice, did not alter satellite cell number or activation, but enhanced myoblast fusion. Primary myoblasts isolated from FRG1 mice showed a myoblast fusion defect that was rescued by FHL1 expression. Therefore, increased FRG1 expression may contribute to a muscular dystrophy phenotype resembling FSHD by impairing myoblast fusion, a defect that can be rescued by enhanced myoblast fusion via expression of FHL1.

  17. Backfitting swimming pool reactors

    International Nuclear Information System (INIS)

    Roebert, G.A.

    1978-01-01

    Calculations based on measurements in a critical assembly, and experiments to disclose fuel element surface temperatures in case of accidents like stopping of primary coolant flow during full power operation, have shown that the power of the swimming pool type research reactor FRG-2 (15 MW, operating since 1967) might be raised to 21 MW within the present rules of science and technology, without major alterations of the pool buildings and the cooling systems. A backfitting program is carried through to adjust the reactor control systems of FRG-2 and FRG-1 (5 MW, housed in the same reactor hall) to the present safety rules and recommendations, to ensure FRG-2 operation at 21 MW for the next decade. (author)

  18. IGORR 1: Proceedings of the 1. meeting of the International Group On Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    West, C D [comp.

    1990-05-01

    Descriptions of the ongoing projects presented at this Meeting were concerned with: New Research Reactor FRM-II at Munich; MITR-II reactor; The Advanced. Neutron Source (ANS) Project; The high Flux Reactor Petten, Status and Prospects; The High Flux Beam Reactor Instrumentation Upgrade; BER-II Upgrade; The BR2 Materials Testing Reactor Past, Ongoing and Under-Study Upgrades; The ORPHEE, Reactor Current Status and Proposed Enhancement of Experimental Variabilities; Construction of the Upgraded JRR-3; Status of the University of Missouri-Columbia Research Reactor Upgrade; the Reactor and Cold Neutron Facility at NIST; Upgrade of Materials Irradiation Facilities in HFIR; Backfitting of the FRG Reactors; University Research Reactors in the United States; and Organization of the ITER Project - Sharing of Informational Procurements. Topics of interest were: Thermal-hydraulic tests and correlations, Corrosion tests and analytical models , Multidimensional kinetic analysis for small cores, Fuel plates fabrication, Fuel plates stability, Fuel irradiation, Burnable poison irradiation, Structural materials irradiation, Neutron guides irradiation, Cold Source materials irradiation, Cold Source LN{sub 2} test, Source LH2-H{sub 2}O reaction (H or D), Instrumentation upgrading and digital control system, Man-machine interface.

  19. IGORR 1: Proceedings of the 1. meeting of the International Group On Research Reactors

    International Nuclear Information System (INIS)

    West, C.D.

    1990-05-01

    Descriptions of the ongoing projects presented at this Meeting were concerned with: New Research Reactor FRM-II at Munich; MITR-II reactor; The Advanced. Neutron Source (ANS) Project; The high Flux Reactor Petten, Status and Prospects; The High Flux Beam Reactor Instrumentation Upgrade; BER-II Upgrade; The BR2 Materials Testing Reactor Past, Ongoing and Under-Study Upgrades; The ORPHEE, Reactor Current Status and Proposed Enhancement of Experimental Variabilities; Construction of the Upgraded JRR-3; Status of the University of Missouri-Columbia Research Reactor Upgrade; the Reactor and Cold Neutron Facility at NIST; Upgrade of Materials Irradiation Facilities in HFIR; Backfitting of the FRG Reactors; University Research Reactors in the United States; and Organization of the ITER Project - Sharing of Informational Procurements. Topics of interest were: Thermal-hydraulic tests and correlations, Corrosion tests and analytical models , Multidimensional kinetic analysis for small cores, Fuel plates fabrication, Fuel plates stability, Fuel irradiation, Burnable poison irradiation, Structural materials irradiation, Neutron guides irradiation, Cold Source materials irradiation, Cold Source LN 2 test, Source LH2-H 2 O reaction (H or D), Instrumentation upgrading and digital control system, Man-machine interface

  20. The qualification of U3O8 as research reactor fuel

    International Nuclear Information System (INIS)

    Krull, W.

    1983-01-01

    This report summarizes the today knowledge of the qualification status of U 3 O 8 as low enriched ( 3 O 8 is so far qualified to start testing of ten (10) fuel elements with an U-density of 3.1 g U/cc in the FRG-2 research reactor. (orig.) [de

  1. The increase of the subthermal neutron flux by using a cold neutron source at the FRG-1

    International Nuclear Information System (INIS)

    Krueger, A.; Turgut, M.H.

    1986-01-01

    The increase of the subthermal neutron flux (wavelength range 4-6 A by a cold neutron source (CNS) at a radial beam tube of the FRG-1 reactor is investigated in combination with different reflectors (H 2 O, C, Be, D 2 O). Advantage factors on the basis of the directed neutron flux, resulting from the use of the CNS, are calculated for various configurations. In addition, the influence of different scattering models (gas, Koppel/Young) for the CNS, group structure, and structural materials are described. Finally, the CNS assembly which is going to be installed at the FRG-1 is treated in detail. For the calculations the transport code NEUTRA and the spectral code GGC-4 are used. (orig.) [de

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

    International Nuclear Information System (INIS)

    Vollradt, J.

    1977-01-01

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

  3. Comparison of LOCA safety analysis in the USA, FRG, and Japan

    International Nuclear Information System (INIS)

    Leach, L.P.; Ybarrondo, L.J.; Hicken, E.F.; Tasaka, K.

    1983-01-01

    The bases for loss-of-coolant accident (LOCA) safety analysis required by reactor licensing regulations in the United States of America (USA), Federal Republic of Germany (FRG), and Japan are investigated and related to new data obtained since the regulations were established. The licensing approaches used in the three countries are similar in that a conservative calculation is called for, the necessary conservatism is unspecified, and new research data have had only limited effect on changing the regulations

  4. List of the reports from reactor safety research of the BMFT and USNRC

    International Nuclear Information System (INIS)

    1977-02-01

    This list presents a survey of reports from the FRG and USA issued on special problems of reactor safety research. The problems include emergency core cooling, the containment during loss of coolant, core meltdown, component safety, external impacts, risk and reliability, quality assurance, fission product transport, radiation exposure, as well as the LWR, HTR and the fast sodium-cooled reactor in general. (orig./HK) [de

  5. Status and aspects of fuel element development for advanced high-temperature reactors in the FRG

    International Nuclear Information System (INIS)

    Nickel, H.; Balthesen, E.

    1975-01-01

    In the FRG three basic fuel element designs for application in high temperature gas cooled reactors are being persued: the spherical element, the graphite block element, and the moulded block element (monolith). This report gives the state of development reached with the three types of elements but also views their specific merits and performance margin and presents aspects of their future development potential for operation in advanced HTGR plants. The development of coated feed and breed particles for application in all HTGR fuel elements is treated in more detail. Summarizing it can be said that all the fuel elements as well as their components have proved their aptitude for the dual cycle systems in numerous fuel element and particle performance tests. To adapt these fuel elements and coated particles for advanced reactor concepts and to develop them up to full technical maturity further testing is still necessary, however. Ways of overcoming problems arising from the more stringent requirements are shown. (orig.) [de

  6. Report to the 12th. Meeting of the Brazil-FRG Mixed Commission

    International Nuclear Information System (INIS)

    1983-01-01

    This report presents the main agreements between the Governments of the Federal Republic of Germany for Nuclear Research and Technology Development Cooperative Program, including the CNEN-FRG cooperation activities during 1982 and the CNEN-FRG proposed cooperation program for 1983. (E.O.)

  7. Conversion planning for the two FRGs of the GKSS-Research Centre

    International Nuclear Information System (INIS)

    Roegler, H.-J.

    1984-01-01

    The German Research Centre GKSS at Geesthacht operates two research reactors, the FRG-1 with 5 MW power and the FRG-2 with 15 MW power. Both plants use the same fuel element for a number of reasons. Such a case is the background of the following report. The conversion planning for both plants has to be taken into account these reasons and to result in a common fuel element again. Moreover, the efforts for the conversion and the related licensing steps will be linked to some other improvements of the performance of the FRGs the operator planned long ago such as (1) upgrading of the FRG-2 to 21 MW and improvement of fuel economy (2) reduction of core size of FRG-1 to improve the effectiveness of the beam tubes. The report deals with all the nuclear and thermal hydraulic investigations carried out for these tasks at INTERATOM including the harmonizing with the operators. (author)

  8. Residual stress analysis by neutron time-of-flight at a reactor source

    International Nuclear Information System (INIS)

    Priesmeyer, H.G.; Schroder, J.

    1990-01-01

    Non-destructive neutron diffractometry for stress analysis will be a powerful experimental tool in material science research performed at the GKSS 5 MW reactor FRG-1. Arguments which show the advantages of the time-of-flight method are given and a suitable high-resolution neutron-efficient type of spectrometer is introduced. First results derived from this method are presented

  9. Status and development programme of the fast breeder reactor system in the Federal Republic of Germany

    International Nuclear Information System (INIS)

    Daeunert, U.; Kessler, G.

    1977-01-01

    The Federal Republic of Germany depends on imports of raw materials for energy purposes, and special importance is therefore given to nuclear energy for reasons of energy policy. Against the background of a nuclear power station capacity amounting to 24,000MW(e) in 1976 (power stations being either on order, under construction or actually in operation), the fast breeder reactor (FBR) is outstandingly significant in the Federal Republic for the economic use both of the uranium and the plutonium accumulating in light-water reactors (LWRs). Work in this sector started in the FRG in 1960. Important stages in development were: the initial activities on the plutonium critical SNEAK at Karlsruhe; successful operation of the KNK 20-MW(e) experimental reactor (sodium-cooled) with the KNK II (plutonium core) follow-up stage; costruction of the 300-MW(e) FBR prototype nuclear power station at Kalkar; and initial planning carried out for a 1300-MW(e) FBR demonstration nuclear power station. A viable infrastructure for safety assessment must go hand in hand with the development of new reactor systems. Prototype reactors in the FRG are therefore already subjected to the same strict licensing procedures under nuclear legislation as are commercial nuclear power stations. Several years' practical experience have shown that this was the right approach. It was recognized in the FRG early on that major long-term technological projects could no longer be implemented by unaided single nations. The SNR-300 FBR prototype project combines the efforts of the governments, power supply companies, reactor manufacturers and nuclear research centres of the FRG, Belgium and the Netherlands. The agreement signed in 1971 by the Rheinisch-Westfaelische Elektrizitaetswerke AG (FRG), Ente Nazionale per l'Energia Elettrica (Italy) and Electricite de France (France) power supply companies on the joint construction of the Super-Phenix and SNR-2 FBR demonstration nuclear power stations, and the German

  10. Current status of the Thai Research Reactor (TRR-1/M1)

    International Nuclear Information System (INIS)

    Chueinta, Siripone; Julanan, Mongkol; Charncanchee, Decharchai

    2006-01-01

    The first Thai Research Reactor, TRR-1 went critical on 27 October 1962 at the maximum power of 1 MW. It was located at Office of Atoms for Peace (OAP) in Bangkok. Since then, TRR-1 was continuously operated and eventually shut down in 1975. Plate type, high-enriched uranium (HEU) and U 3 O 8 A1 cladding were used as the reactor fuel. Light water was used as moderator and coolant as well. In 1975, because of the problem from fuel supplier and also to supporting the Treaty of Non Proliferation of Nuclear Weapon or NPT, TRR-1 was shut down for modification. The reactor core and control system were disassembled and replaced by TRIGA Mark III. A new core was a hexagonal core shape designed by General Atomics (GA). Afterwards, TRR-1 was officially renamed to the Thai Research Reactor-1/Modification 1 (TRR-1/M1). TRR-1/M1 is a multipurpose swimming pool type reactor with nominal power of 2 MW. The TRR-1/M1 uses uranium enriched at 20% in U-235 (LEU) and ZrH alloy as fuel. Light water is also used as coolant and moderator. At present, the reactor is operating with core No.14. The reactor has been serving for various kinds of utilization namely, radioisotope production, neutron activation analysis, beam experiments and reactor physics experiments. (author)

  11. Utilization of brown coal in FRG power plants

    Energy Technology Data Exchange (ETDEWEB)

    Kotler, V.R.

    1985-07-01

    FRG methods are studied for utilizing brown coal in view of the development of Kansk-Achinsk brown coal deposits. The use of brown coal in FRG power plants has increased from 15% in 1950- 1960 to 85% (total output) in 1982, providing 79.4 TWh of electrical energy. The remainder was used for briquetting, pulverization and breeze coke. In 1982 nearly 100 million tons of brown coal were burned by six large power stations (rated capacity 11,400 MW) to produce nearly 80 billion kWh of energy. Measures are discussed taken to reduce slagging and to remove excessive moisture content. Problems are analyzed associated with increased contamination of the atmosphere in areas with high population density (412/km/sup 2/) and cost of suppression is reviewed. According to available data, the cost of preventive measures taken by FRG, USA, Japan and the Netherlands is equal to 30% of the total cost of the energy. The most critical problem is suppression of sulfur dioxide, either by dry or wet scrubbers or by the addition of dry dolomite or lime to the furnace (75% of all SO/sub 2/ emissions in FRG comes from power stations). A method is described developed by RWE based on a series of distribution headers in the upper part of combustion chambers. At best, 70-80% reduction can be achieved. 14 references.

  12. Experimental and clinical radiotherapy of malignant tumours in FRG (scientific records of Soviet specialists professional business trip)

    International Nuclear Information System (INIS)

    Yarmonenko, S.P.; Berdov, B.A.; Akademiya Meditsinskikh Nauk SSSR, Obninsk. Nauchno-Issledovatel'skij Inst. Meditsinskoj Radiologii)

    1986-01-01

    A report is made on the business trip of Soviet radiologists in the FRG to study the experience in the organization and realization of radiologic and radiobiologic researches in oncology. Clinico-radiobiological studies in the FRG on radiobiology of human tumours are pointed out. Efficiency of the extencive use of computer-tomography equipment is emphasized. The methods of local electromagnetic hyperthermia are actively used

  13. Future plans for the design and construction of fast reactor power stations in the Federal Republic of Germany

    International Nuclear Information System (INIS)

    Kempken, M.; Koehler, M.; Wolff, M.

    1978-01-01

    Some important design features of future fast reactors in the Federal Republic of Germany (FRG) are presented, in particular for the demonstration plant SNR 2 which is to follow the prototype SNR 300, presently under construction in Kalkar. The SNR 2 conceptual design will be based on the SNR 300 design as far as possible. Programmes for the introduction of fast breeder reactor power stations on the part of the governments, the utilities and suppliers are based on broad international co-operation. The FRG is a country which imports a high proportion of its primary energy and it has rather small resources of natural uranium. The natural uranium realistically available to the FRG will allow nuclear energy to play a substantial role in the long-term energy supply only if present uranium utilization based on LWRs is supplemented and replaced by breeder reactor utilization later. To maintain this option, efforts towards the development, design and construction of fast breeder reactors have to be intensively continued in the FRG. The construction of the first large power station with a fast breeder reactor, SNR 2, will, according to present planning, start in the middle of the 80s. Operation can be expected to start at the beginning of the 90s. The present fast breeder programme in the FRG promises to develop reactors, reprocessing and fuel manufacturing plants to such a degree that by the end of this century the introduction of a substantial number of fast reactor power stations will be possible. (author)

  14. The Chernobyl reactor accident. Pt. 1 and 2

    International Nuclear Information System (INIS)

    1986-06-01

    The report first summarizes the available information on the various incidents of the whole accident scenario, and combines the information to present a first general outline and a basis for appraisal. The most significant incidents reported, namely power excursion, core meltdown, and fire, are discussed with a view to the reactor design and safety of reactors installed in the FRG. The main differences and advantages of German reactor designs are shown, as e.g.: Power excursions are mastered by inherent physical conditions; far better redundancy of engineered safety systems; enclosure of the complete reactor cooling system in a pressure-retaining steel containment; reactor buildings being made of reinforced concrete. The second part of the report deals with the radiological effects to be expected for our country. Data are given on the varying radiological exposure of the different regions. The fate and uptake of radioactivity in the human body are discussed. The conclusion drawn from the data presented is that the individual exposure due to the reactor accident will remain within the variations and limits of natural radioactivity and effects. (orig./HP) [de

  15. Gas reactor international cooperative program interim report: United States/Federal Republic of Germany nuclear licensing comparison

    International Nuclear Information System (INIS)

    1978-09-01

    In order to compare US and FRG Nuclear Licensing, a summary description of United States Nuclear Licensing is provided as a basis. This is followed by detailed information on the participants in the Nuclear Licensing process in the Federal Republic of Germany (FRG). FRG licensing procedures are described and the rules and regulations imposed are summarized. The status of gas reactor licensing in both the U.S. and the FRG is outlined and overall conclusions are drawn as to the major licensing differences. An appendix describes the most important technical differences between US and FRG criteria

  16. Stability analysis of the Ghana Research Reactor-1 (GHARR-1)

    International Nuclear Information System (INIS)

    Della, R.; Alhassan, E.; Adoo, N.A.; Bansah, C.Y.; Nyarko, B.J.B.; Akaho, E.H.K.

    2013-01-01

    Highlights: • We developed a theoretical model to study the stability of the Ghana Research Reactor-1. • The neutronics transfer function was described by the point kinetics model for a single group of delayed neutrons. • The thermal hydraulics transfer function was based on the modified lumped parameter concept. • A computer code, RESA (REactor Stability Analysis) was developed. • Results show that the closed-loop transfer function was stable and well damped for variable operating power levels. - Abstract: A theoretical model has been developed to study the stability of the Ghana Research Reactor one (GHARR-1). The closed-loop transfer function of GHARR-1 was established based on the model, which involved the neutronics and the thermal hydraulics transfer functions. The reactor kinetics was described by the point kinetics model for a single group of delayed neutrons, whilst the thermal hydraulics transfer function was based on the modified lumped parameter concept. The inherent internal feedback effect due to the fuel and the coolant was represented by the fuel temperature coefficient and the moderator temperature coefficient respectively. A computer code, RESA (REactor Stability Analysis), entirely in Java was developed based on the model for systems analysis. Stability analysis of the open-loop transfer function of GHARR-1 based on the Nyquist criterion and Bode diagrams using RESA, has shown that the closed-loop transfer function was marginally stable for variable operating power levels. The relative stability margins of GHARR-1 were also identified

  17. Fast breeder reactor research

    International Nuclear Information System (INIS)

    1975-01-01

    reactors of the future, the body of research aimed at developing liquid metal cooled fast reactors, national plans for work in 1976 on developing fast reactors - these were some of the topics discussed in connection with the national programmes. The development of power reactors involves a wide range of problems in the fields of nuclear and reactor physics, the thermophysics, chemistry, physics and technology of the cooling system, structural materials and nuclear fuel, the fabrication of reliable fuel elements and operating equipment, reactor monitoring and control, spent fuel reprocessing, the economics of constructing fast power reactors, nuclear safety, etc. The IWGFR, as at previous meetings, therefore paid great attention to the matter of holding international specialists' meetings. The working group recommended that the IAEA should organize the following IWGFR meetings in 1976: (1) In-Service Inspection and Monitoring (Bensberg, FRG, March 1976). (2) Cavitation in Sodium and Studies of Analogy with Water as Compared to Sodium (Cadarache, France, April 1976). (3) High Temperature Structural Design Technology (United States, May 1976) (4) Aerosol Formation, Vapour Deposits and Sodium Vapour Trapping (France, September-December 1976). The Group welcomed the IAEA's proposal to hold specialists' meetings on 'Fast Reactor Instrumentation' and 'Fuel Reprocessing and Recycling Techniques' within the framework of the Agency's programme of working groups in 1976. After discussing questions of co-ordinating and organizing international conferences on fast reactors, the IWGFR agreed to send representatives to the joint meeting of the American Nuclear Society and the American Institute of Metallurgical Engineers on 'Liquid Metal Technology', to be held at Champion, Pennsylvania, U.S.A. from 3-6 May 1976, and recommended that the IAEA should organize an international symposium on the 'Design, Construction and Operating Experience of Demonstration Fast Power Reactors' at Bologna

  18. Stability of Schwarzschild-like solutions in f(R,G) gravity models

    International Nuclear Information System (INIS)

    De Felice, Antonio; Suyama, Teruaki; Tanaka, Takahiro

    2011-01-01

    We study linear metric perturbations around a spherically symmetric static spacetime for general f(R,G) theories, where R is the Ricci scalar and G is the Gauss-Bonnet term. We find that, unless the determinant of the Hessian of f(R,G) is zero, even-type perturbations have a ghost for any multipole mode. In order for these theories to be plausible alternatives to general relativity, the theory should satisfy the condition that the ghost is massive enough to effectively decouple from the other fields. We study the requirement on the form of f(R,G) which satisfies this condition. We also classify the number of propagating modes both for the odd-type and the even-type perturbations and derive the propagation speeds for each mode.

  19. Transport of radioactive materials in the Federal Republic of Germany (FRG) since 3. October 1990

    International Nuclear Information System (INIS)

    Alter, U.; Collin, F.W.; Fasten, C.

    1993-01-01

    The paper presents a survey on the transport of radioactive materials in the FRG. For shipments of nuclear material and large sources in the FRG - and also in the former German Democratic Republic - a license from the Federal Office for Radiation Protection (the competent authority in the FRG) due to the Atomic law and Radiological Protection Requirements are necessary. (J.P.N.)

  20. Medium and long-term supply of the FRG with natural uranium

    Energy Technology Data Exchange (ETDEWEB)

    Stephany, M

    1976-08-01

    A short survey is presented of the situation concerning the FRG's uranium supply taking into special consideration the economical aspects. Characteristic for the demand-supply-situation is the fact that the development of prices does not influence the costs as much as this is the case with other fuels, and that assurance of supply for the FRG is underpinned by participation in foreign enterprises.

  1. Proceedings of the US/FRG research symposium: effects of atmospheric pollutants on the spruce-fir forests of the Eastern United States and the Federal Republic of Germany

    Science.gov (United States)

    Gerard, tech. coord. Hertel; Gerard Hertel

    1988-01-01

    Includes 66 papers presented at the US/FRG research symposium: effects of atmospheric pollutants on the spruce-fir forests of the Eastern United States and the Federal Republic of Germany, which was held October 19-23, 1987, in Burlington, Vermont.

  2. Exposure calculations for the FRG isotopic heat source project environmental assessment

    International Nuclear Information System (INIS)

    Metcalf, I.L.

    1997-01-01

    The report documents the maximum exposure for transfer of the Federal Republic of Germany (FRG) Isotopic Heat Sources from the 324 Building and placed in interim storage at the Central Waste Complex (CWC). These results are to be reported in the Environmental Assessment DOE-EA- 1 21 1

  3. Safeguarding research reactors

    International Nuclear Information System (INIS)

    Powers, J.A.

    1983-03-01

    The report is organized in four sections, including the introduction. The second section contains a discussion of the characteristics and attributes of research reactors important to safeguards. In this section, research reactors are described according to their power level, if greater than 25 thermal megawatts, or according to each fuel type. This descriptive discussion includes both reactor and reactor fuel information of a generic nature, according to the following categories. 1. Research reactors with more than 25 megawatts thermal power, 2. Plate fuelled reactors, 3. Assembly fuelled reactors. 4. Research reactors fuelled with individual rods. 5. Disk fuelled reactors, and 6. Research reactors fuelled with aqueous homogeneous fuel. The third section consists of a brief discussion of general IAEA safeguards as they apply to research reactors. This section is based on IAEA safeguards implementation documents and technical reports that are used to establish Agency-State agreements and facility attachments. The fourth and last section describes inspection activities at research reactors necessary to meet Agency objectives. The scope of the activities extends to both pre and post inspection as well as the on-site inspection and includes the examination of records and reports relative to reactor operation and to receipts, shipments and certain internal transfers, periodic verification of fresh fuel, spent fuel and core fuel, activities related to containment and surveillance, and other selected activities, depending on the reactor

  4. Burnup measurements at the RECH-1 research reactor

    International Nuclear Information System (INIS)

    Henriquez, C.; Navarro, G.; Pereda, C.; Torres, H.; Pena, L.; Klein, J.; Calderon, D.; Kestelman, A.J.

    2002-01-01

    The Chilean Nuclear Energy Commission has decided to produce LEU fuel elements for the RECH-1 research reactor. During December 1998, the Fuel Fabrication Plant delivered the first four fuel elements, called leaders, to the RECH-1 reactor. The set was introduced into the reactor's core, following the normal routine, but performing a special follow-up on their behavior inside and outside the core. In order to measure the burn-up of the leader fuel elements, it was decided to develop a burn-up measurements system to be installed into the RECH-1 reactor pool, and to decline the use of a similar system, which operates in a hot cell. The main reason to build this facility was to have the capability to measure the burn-up of fuel elements without waiting for long decay period. This paper gives a brief description of the facility to measure the burn-up of spent fuel elements installed into the reactor pool, showing the preliminary obtained spectra and briefly discussing them. (author)

  5. Necessity of research reactors

    International Nuclear Information System (INIS)

    Ito, Tetsuo

    2016-01-01

    Currently, only three educational research reactors at two universities exist in Japan: KUR, KUCA of Kyoto University and UTR-KINKI of Kinki University. UTR-KINKI is a light-water moderated, graphite reflected, heterogeneous enriched uranium thermal reactor, which began operation as a private university No. 1 reactor in 1961. It is a low power nuclear reactor for education and research with a maximum heat output of 1 W. Using this nuclear reactor, researches, practical training, experiments for training nuclear human resources, and nuclear knowledge dissemination activities are carried out. As of October 2016, research and practical training accompanied by operation are not carried out because it is stopped. The following five items can be cited as challenges faced by research reactors: (1) response to new regulatory standards and stagnation of research and education, (2) strengthening of nuclear material protection and nuclear fuel concentration reduction, (3) countermeasures against aging and next research reactor, (4) outflow and shortage of nuclear human resources, and (5) expansion of research reactor maintenance cost. This paper would like to make the following recommendations so that we can make contribution to the world in the field of nuclear power. (1) Communication between regulatory authorities and business operators regarding new regulatory standards compliance. (2) Response to various problems including spent fuel measures for long-term stable utilization of research reactors. (3) Personal exchanges among nuclear experts. (4) Expansion of nuclear related departments at universities to train nuclear human resources. (5) Training of world-class nuclear human resources, and succession and development of research and technologies. (A.O.)

  6. Experience and research with the IEA-R1 Brazilian reactor

    International Nuclear Information System (INIS)

    Fulfaro, R.; Sousa, J.A. de; Nastasi, M.J.C.; Vinhas, L.A.; Lima, F.W.

    1982-06-01

    The IEA-R1 reactor of the Instituto de Pesquisas Energeticas e Nucleares, IPEN, of Sao Paulo, Brazil, a lightwater moderated swimming-pool research reactor of MTR type, went critical for the first time on September 16, 1957. In a general way, in these twenty four years the reactor was utilized without interruption by users of IPEN and other institutions, for the accomplishment of work in the field of applied and basic research, for master and doctoral thesis and for technical development. Some works performed and the renewal programme established for the IEA-R1 research reactor in which several improvements and changes were made. Recent activities in terms of production of radioisotopes and some current research programm in the field of Radiochemistry are described, mainly studies and research on chemical reactions and processes using radioactive tracers and development of radioanalytical methods, such as neutron activation and isotopic dilution. The research programmes of the Nuclear Physics Division of IPEN, which includes: nuclear spectroscopy studies and electromagnetic hyperfine interactions; neutron diffraction; neutron inelastic scattering studies in condensed matter; development and application of the technique of fission track register in solid state detectors; neutron radioactive capture with prompt gamma detection and, finally, research in the field of nuclear metrology, are presented. (Author) [pt

  7. Experience and research with the IEA-R1 Brazilian reactor

    International Nuclear Information System (INIS)

    Fulfaro, R.; Sousa, J.A. de; Nastasi, M.J.C.; Vinhas, L.A.; Lima, F.W. de.

    1982-06-01

    The IEA-R1 reactor of the Instituto de Pesquisas Energeticas e Nucleares, IPEN, of Sao Paulo, Brazil, a lighwater moderated swimming-pool research reactor of MTR type, went critical for the first time on September 16, 1957. In a general way, in these twenty four years the reactor was utilized without interruption by users of IPEN and other institutions, for the accomplishment of work in the field of applied and basic research, for master and doctoral thesis and for technical development. Some works performed and the renewal programme established for the IEA-R1 research reactor in which several improvements and changes were made. Recent activities in terms of production of radioisotopes and some current research programm in the field of Radiochemistry are described, mainly studies and research on chemical reactions and processes using radioactive tracers and development of radioanalytical methods, such as neutron activation and isotopic dilution. It is also presented the research programmes of the Nuclear Physics Division of IPEN, which includes: nuclear spectroscopy studies and electromagnetic hyperfine interactions; neutron diffraction; neutron inelastic scattering studies in condensed matter; development and application of the technique of fission track register in solid state detectors; neutron radioactive capture with prompt gamma detection and, finally, research in the field of nuclear metrology. (Author) [pt

  8. Research reactors in Argentina

    International Nuclear Information System (INIS)

    Carlos Ruben Calabrese

    1999-01-01

    Argentine Nuclear Development started in early fifties. In 1957, it was decided to built the first a research reactor. RA-1 reactor (120 kw, today licensed to work at 40 kW) started operation in January 1958. Originally RA-1 was an Argonaut (American design) reactor. In early sixties, the RA-1 core was changed. Fuel rods (20% enrichment) was introduced instead the old Argonaut core design. For that reason, a critical facility named RA-0 was built. After that, the RA-3 project started, to build a multipurpose 5 MW nuclear reactor MTR pool type, to produce radioisotopes and research. For that reason and to define the characteristics of the RA-3 core, another critical facility was built, RA-2. Initially RA-3 was a 90 % enriched fuel reactor, and started operation in 1967. When Atucha I NPP project started, a German design Power Reactor, a small homogeneous reactor was donated by the German Government to Argentina (1969). This was RA-4 reactor (20% enrichment, 1W). In 1982, RA-6 pool reactor achieved criticality. This is a 500 kW reactor with 90% enriched MTR fuel elements. In 1990, RA-3 started to operate fueled by 20% enriched fuel. In 1997, the RA-8 (multipurpose critical facility located at Pilcaniyeu) started to operate. RA-3 reactor is the most important CNEA reactor for Argentine Research Reactors development. It is the first in a succession of Argentine MTR reactors built by CNEA (and INVAP SE ) in Argentina and other countries: RA-6 (500 kW, Bariloche-Argentina), RP-10 (10MW, Peru), NUR (500 kW, Algeria), MPR (22 MW, Egypt). The experience of Argentinian industry permits to compete with foreign developed countries as supplier of research reactors. Today, CNEA has six research reactors whose activities have a range from education and promotion of nuclear activity, to radioisotope production. For more than forty years, Argentine Research Reactors are working. The experience of Argentine is important, and argentine firms are able to compete in the design and

  9. Design of a new research reactor : 1st year conceptual design

    International Nuclear Information System (INIS)

    Park, Cheol; Lee, B. C.; Chae, H. T.

    2004-01-01

    A new research reactor model satisfying the strengthened regulatory environments and the changed circumstances around nuclear society should be prepared for the domestic and international demand of research reactor. This can also lead to the improvement of technologies and fostering manpower obtained during the construction and the operation of HANARO. In this aspect, this study has been launched and the 1st year conceptual design has been carried out in 2003. The major tasks performed at the first year of conceptual design stage are as follows; Establishments of general design requirements of research reactors and experimental facilities, Establishment of fuel and reactor core concepts, Preliminary analysis of reactor physics and thermal-hydraulics for conceptual core, Conceptual design of reactor structure and major systems, International cooperation to establish foundations for exporting

  10. Refurbishment of Pakistan research reactor (PARR-1) for stainless steel lining of the reactor pool

    International Nuclear Information System (INIS)

    Salahuddin, A.; Israr, M.; Hussain, M.

    2002-01-01

    Pakistan Research Reactor-1 (PARR-1) is a pool-type research reactor. Reactor aging has resulted in the increase of water seepage from the concrete walls of the reactor pool. To stop the seepage, it was decided to augment the existing pool walls with an inner lining of stainless steel. This could be achieved only if the pool walls could be accessed unhindered and without excessive radiation doses. For this purpose a partial decommissioning was done by removing all active core components including standard/control fuel elements, reflector elements, beam tubes, thermal shield, core support structure, grid plate and the pool's ceramic tiles, etc. An overall decommissioning program was devised which included procedures specific to each item. This led to the development of a fuel transport cask for transportation, and an interim fuel storage bay for temporary storage of fuel elements (until final disposal). The safety of workers and the environment was ensured by the use of specially designed remote handling tools, appropriate shielding and pre-planned exposure reduction procedures based on the ALARA principle. During the implementation of this program, liquid and solid wastes generated were legally disposed of. It is felt that the experience gained during the refurbishment of PARR-1 to install the stainless steel liner will prove useful and better planning and execution for the future decommissioning of PARR-1, in particular, and for other research reactors like PARR-2 (27 kW MNSR), in general. Furthermore, due to the worldwide activities on decommissioning, especially those communicated through the IAEA CRP on 'Decommissioning Techniques for Research Reactors', the importance of early planning has been well recognized. This has made possible the implementation of some early steps like better record keeping, rehiring of trained manpower, and creation of interim and final waste storage. (author)

  11. Multipurpose research reactors

    International Nuclear Information System (INIS)

    1988-01-01

    The international symposium on the utilization of multipurpose research reactors and related international co-operation was organized by the IAEA to provide for information exchange on current uses of research reactors and international co-operative projects. The symposium was attended by about 140 participants from 36 countries and two international organizations. There were 49 oral presentations of papers and 24 poster presentations. The presentations were divided into 7 sessions devoted to the following topics: neutron beam research and applications of neutron scattering (6 papers and 1 poster), reactor engineering (6 papers and 5 posters), irradiation testing of fuel and material for fission and fusion reactors (6 papers and 10 posters), research reactor utilization programmes (13 papers and 4 posters), neutron capture therapy (4 papers), neutron activation analysis (3 papers and 4 posters), application of small reactors in research and training (11 papers). A separate abstract was prepared for each of these papers. Refs, figs and tabs

  12. Research reactors

    International Nuclear Information System (INIS)

    Merchie, Francois

    2015-10-01

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

  13. Research reactors

    International Nuclear Information System (INIS)

    Kowarski, L.

    1955-01-01

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

  14. IAEA safety standards for research reactors

    International Nuclear Information System (INIS)

    Abou Yehia, H.

    2007-01-01

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

  15. How long does the reactor in a nuclear submarine remain intact on the sea bed

    International Nuclear Information System (INIS)

    Anon.

    1978-01-01

    In a brief report from a conference held by OECD in Hamburg on the safety of nuclear powered ships, the need for international conventions in this field was stressed by a representative of the Ministry of Research and Technology of the FRG. It was also maintained that nuclear ships are the safest in the world. A representative of the FRG's reactor safety dept. said, however, that nuclear warships will never become subject to international control and that the safety of such ships was not reassuring. At a press conference a U.S. Coastguard representative refused to reply to a question on how long it would take before radioactive materials ieaked from the wreck of the 'Thresher', as this was classified information. A German worker maintained that the oxygen content in the water was too low at that depth for corrosion to take place, while another maintained that there was evidence that the reactor was already leaking. (JIW)

  16. A review of fast reactor activities in Italy

    International Nuclear Information System (INIS)

    Pierantoni, F.; Tavoni, R.

    1990-01-01

    Italy's total primary energy demand increased in 1989 by 2.8% continuing the trend initiated in 1984. Net electricity import increased by 9% reaching the total of 34 TWh, equivalent to the electrical production of 5 power stations in base load operation. As known, after the Chernobyl accident, a moratorium of at least 5 years was imposed on the construction of new nuclear reactors in Italy. The PEN (National Energy Plan), already approved by the government and submitted to parliament for discussion, calls for a programme of research and industrial development of reactors with inherent and passive safety features. In this framework during 1989, a collaboration of ENEL, Ansaldo and ENEA with some designers of innovative reactors, mainly water-cooled, has taken place in many fields. The Italian participation and contribution in international groups has continued with the aim of maintaining and updating the know-how on the most relevant themes of research. Contacts are underway in the area of advanced reactors with several countries such as the USA, the FRG, the UK and France. With France an agreement for cooperation in the area of future nuclear reactors and advanced technologies for the upgrading of industries was signed in April 1990. (author). 1 fig., 2 tabs

  17. Irradiation experience of IPEN fuel at IEA-R1 research reactor

    International Nuclear Information System (INIS)

    Perrotta, Jose A.; Neto, Adolfo; Durazzo, Michelangelo; Souza, Jose A.B. de; Frajndlich, Roberto

    1998-01-01

    IPEN/CNEN-SP produces, for its IEA-R1 Research Reactor, MTR fuel assemblies based on U 3 O 8 -Al dispersion fuel type. Since 1985 a qualification program on these fuel assemblies has been performed. Average 235 U burnup of 30% and peak burnup of 50% was already achieved by these fuel assemblies. This paper presents some results acquire, by these fuel assemblies, under irradiation at IEA-R1 Research Reactor. (author)

  18. Demolition of the FRJ-1 research reactor (MERLIN)

    International Nuclear Information System (INIS)

    Stahn, B.; Matela, K.; Zehbe, C.; Poeppinghaus, J.; Cremer, J.

    2003-01-01

    FRJ-2 (MERLIN), the swimming pool reactor cooled and moderated by light water, was built at the then Juelich Nuclear Research Establishment (KFA) between 1958 and 1962. In the period between 1964 and 1985, it was used for. The reactor was decommissioned in 1985. Since 1996, most of the demolition work has been carried out under the leadership of a project team. The complete secondary cooling system was removed by late 1998. After the cooling loops and experimental installations had been taken out, the reactor vessel internals were removed in 2000 after the water had been drained from the reactor vessel. After the competent authority had granted a license, demolition of the reactor block, the central part of the research reactor, was begun in October 2001. In a first step, the reactor operating floor and the reactor attachment structures were removed by the GNS/SNT consortium charged with overall planning and execution of the job. This phase gave rise to approx. The reactor block proper is dismantled in a number of steps. A variety of proven cutting techniques are used for this purpose. Demolition of the reactor block is to be completed in the first half of 2003. (orig.) [de

  19. Research reactor core conversion guidebook. V.1: Summary

    International Nuclear Information System (INIS)

    1992-04-01

    In view of the proliferation concerns caused by the use of highly enriched uranium (HEU) and in anticipation that the supply of HEU to research and test reactors will be more restricted in the future, this guidebook has been prepared to assist research reactor operators in addressing the safety and licensing issues for conversion of their reactor cores from the use of HEU fuel to the use of low enriched uranium fuel. This Guidebook, in five volumes, addresses the effects of changes in the safety-related parameters of mixed cores and the converted core. It provides an information base which should enable the appropriate approvals processes for implementation of a specific conversion proposal, whether for a light or for a heavy water moderated research reactor. Refs, figs, bibliographies and tabs

  20. Water cooled reactor technology: Safety research abstracts no. 1

    International Nuclear Information System (INIS)

    1990-01-01

    The Commission of the European Communities, the International Atomic Energy Agency and the Nuclear Energy Agency of the OECD publish these Nuclear Safety Research Abstracts within the framework of their efforts to enhance the safety of nuclear power plants and to promote the exchange of research information. The abstracts are of nuclear safety related research projects for: pressurized light water cooled and moderated reactors (PWRs); boiling light water cooled and moderated reactors (BWRs); light water cooled and graphite moderated reactors (LWGRs); pressurized heavy water cooled and moderated reactors (PHWRs); gas cooled graphite moderated reactors (GCRs). Abstracts of nuclear safety research projects for fast breeder reactors are published independently by the Nuclear Energy Agency of the OECD and are not included in this joint publication. The intention of the collaborating international organizations is to publish such a document biannually. Work has been undertaken to develop a common computerized system with on-line access to the stored information

  1. Dose limits and licensing requirements for the limitation of the emission of radioactive materials from nuclear power stations in the FRG and the USA

    Energy Technology Data Exchange (ETDEWEB)

    Schwibach, J; Huber, O

    1975-08-01

    In licensing the operation of nuclear power plants in the FRG and USA, particular limitations of the release of radioactive materials in the air and water are layed down to correspond to the protective laws of radiation and environmental protection. The first limiting recommendations for the removal of radioactive waste waters were worked out in 1965 in the FRG and in 1968/69 for the removal of radioactive exhaust air of nuclear power plants. Based on this, in 1975 these relevant regulations were included in the draft of the new radiation protection specification. In 1971, these type of guidelines were put to discussion in the USA and were dismissed in 1975 in a licensing regulation of the NRC. These regulations or guidelines differ in their various dose limits. For example, the German dose limits of 30 mrem/a whole body dose for radioactive materials in the exhaust air of nuclear power plants and of 90 mrem/a for the thyroid dose through radioiodine via the exposure exhaust air-pasture-cow-milk-infant are often compared to the American dose limits of 5 mrem/a whole body dose and 15 mrem/a skin dose as well as 15 mrem/a thyroid dose. Such a numerical comparison is, howewer, wrong. The dose limits used in the FRG are, e.g., not to be exceeded. Furthermore, in the FRG, all contributions to be calculated on one site are to be considered. In the USA, the corresponding values are only valid for actual exposure paths due to the emission of a power reactor. They can be multiply exceeded. Thus the German licensing practise is clearly more restrictive.

  2. Research reactor DHRUVA

    International Nuclear Information System (INIS)

    Veeraraghaven, N.

    1990-01-01

    DHRUVA, a 100 MWt research reactor located at the Bhabha Atomic Research Centre, Bombay, attained first criticality during August, 1985. The reactor is fuelled with natural uranium and is cooled, moderated and reflected by heavy water. Maximum thermal neutron flux obtained in the reactor is 1.8 X 10 14 n/cm 2 /sec. Some of the salient design features of the reactor are discussed in this paper. Some important features of the reactor coolant system, regulation and protection systems and experimental facilities are presented. A short account of the engineered safety features is provided. Some of the problems that were faced during commissioning and the initial phase of power operation are also dealt upon

  3. Reactor Materials Research

    International Nuclear Information System (INIS)

    Van Walle, E.

    2001-01-01

    The activities of the Reactor Materials Research Department of the Belgian Nuclear Research Centre SCK-CEN in 2000 are summarised. The programmes within the department are focussed on studies concerning (1) fusion, in particular mechanical testing; (2) Irradiation Assisted Stress Corrosion Cracking (IASCC); (3) nuclear fuel; and (4) Reactor Pressure Vessel Steel (RPVS)

  4. Reactor Materials Research

    Energy Technology Data Exchange (ETDEWEB)

    Van Walle, E

    2001-04-01

    The activities of the Reactor Materials Research Department of the Belgian Nuclear Research Centre SCK-CEN in 2000 are summarised. The programmes within the department are focussed on studies concerning (1) fusion, in particular mechanical testing; (2) Irradiation Assisted Stress Corrosion Cracking (IASCC); (3) nuclear fuel; and (4) Reactor Pressure Vessel Steel (RPVS)

  5. IPR-R1 TRIGA research reactor decommissioning plan

    International Nuclear Information System (INIS)

    Andrade Grossi, Pablo; Oliveira de Tello, Cledola Cassia; Mesquita, Amir Zacarias

    2008-01-01

    The International Atomic Energy Agency (IAEA) is concerning to establish or adopt standards of safety for the protection of health, life and property in the development and application of nuclear energy for peaceful purposes. In this way the IAEA recommends that decommissioning planning should be part of all radioactive installation licensing process. There are over 200 research reactors that have either not operated for a considerable period of time and may never return to operation or, are close to permanent shutdown. Many countries do not have a decommissioning policy, and like Brazil not all installations have their decommissioning plan as part of the licensing documentation. Brazil is signatory of Joint Convention on the safety of spent fuel management and on the safety of radioactive waste management, but until now there is no decommissioning policy, and specifically for research reactor there is no decommissioning guidelines in the standards. The Nuclear Technology Development Centre (CDTN/CNEN) has a TRIGA Mark I Research Reactor IPR-R1 in operation for 47 years with 3.6% average fuel burn-up. The original power was 100 k W and it is being licensed for 250 k W, and it needs the decommissioning plan as part of the licensing requirements. In the paper it is presented the basis of decommissioning plan, an overview and the end state / final goal of decommissioning activities for the IPR-R1, and the Brazilian ongoing activities about this subject. (author)

  6. Ozone profiles at Juelich, FRG, during 1988 and 1989

    International Nuclear Information System (INIS)

    Smit, H.G.J.; Straeter, W.; Loup, H.; Kley, D.

    1989-12-01

    Ozone soundings were performed regular at Juelich, FRG (50deg 41' N, 6deg 24' E). This report, the first one of an intended series, contains information on technical aspects and presents vertical profiles obtained during 1988 and 1989. (orig.) [de

  7. Parmenidesa kula bytu: fragm. B 8.43-45 (PARMENIDES' SPHERE OF BEING: FRG. 8. 43-45

    Directory of Open Access Journals (Sweden)

    Jerzy Pawliszcze

    2006-12-01

    Full Text Available The main topic of the paper is a problem of locative sense of Parmenidean 'esti' - 'is' in frg. 8 and other places. The question whether being in Parmenides' poem is spherical, and in result, spatial is one of the most 'pregnant' questions in the investigations of the Pre-Socratic philosophy before and after this Father of Ontology. The author is analyzing the use of metaphors in Parmenides and Heraclitus from the Aristotelian point of view (Po. 1457 b 25-33 in order to prove that sphere of being in frg. 8.43 in comparison with line 1.29 about 'well-rounded Truth' is only an antilocative metaphor (cf. frg.5. He is examining many different interpretations of old and contemporary authors and textual testimonies of Simplicius. He is also analyzing pro et contra arguments. The conclusion might also serve as an premise in old controversy of modern semantics about primary locative presupposition of existential constructions in many languages, especially IE (Lyons, Kahn - maybe in all of them.

  8. Major update of Safety Analysis Report for Thai Research Reactor-1/Modification 1

    Energy Technology Data Exchange (ETDEWEB)

    Tippayakul, Chanatip [Thailand Institute of Nuclear Technology, Bangkok (Thailand)

    2013-07-01

    Thai Research Reactor-1/Modification 1 (TRR-1/M1) was converted from a Material Testing Reactor in 1975 and it had been operated by Office of Atom for Peace (OAP) since 1977 until 2007. During the period, Office of Atom for Peace had two duties for the reactor, that is, to operate and to regulate the reactor. However, in 2007, there was governmental office reformation which resulted in the separation of the reactor operating organization from the regulatory body in order to comply with international standard. The new organization is called Thailand Institute of Nuclear Technology (TINT) which has the mission to promote peaceful utilization of nuclear technology while OAP remains essentially the regulatory body. After the separation, a new ministerial regulation was enforced reflecting a new licensing scheme in which TINT has to apply for a license to operate the reactor. The safety analysis report (SAR) shall be submitted as part of the license application. The ministerial regulation stipulates the outlines of the SAR almost equivalent to IAEA standard 35-G1. Comparing to the IAEA 35-G1 standard, there were several incomplete and missing chapters in the original SAR of TRR1/M1. The major update of the SAR was therefore conducted and took approximately one year. The update work included detail safety evaluation of core configuration which used two fuel element types, the classification of systems, structures and components (SSC), the compilation of detail descriptions of all SSCs and the review and evaluation of radiation protection program, emergency plan and emergency procedure. Additionally, the code of conduct and operating limits and conditions were revised and finalized in this work. A lot of new information was added to the SAR as well, for example, the description of commissioning program, information on environmental impact assessment, decommissioning program, quality assurance program and etc. Due to the complexity of this work, extensive knowledge was

  9. Thai Research Reactor (TRR-1/M1) Neutron Beam Measurements

    International Nuclear Information System (INIS)

    Ratanatongchai, Wichian

    2009-07-01

    Full text: Neutron beam tube of neutron radiography facility at Thai Research Reactor (TRR-1/M1) Thailand Institute of Nuclear Technology (public organization) is a divergent beam. The rectangular open-end of the beam tube is 16 cm x 17 cm while the inner-end is closed to the reactor core. The neutron beam size was measured using 20 cm x 40 cm neutron imaging plate. The measurement at the position 100 cm from the end of the collimator has shown that the beam size was 18.2 cm x 19.0 cm. Gamma ray in neutron the beam was also measured by the identical position using industrial X ray film. The area of gamma ray was 27.8 cm x 31.1 cm with the highest intensity found to be along the neutron beam circumference

  10. Nuclear research reactors

    International Nuclear Information System (INIS)

    1985-01-01

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

  11. Generating the flux map of Nigeria Research Reactor-1 for efficient ...

    African Journals Online (AJOL)

    One of the main uses to which the Nigeria Research Reactor-1 (NIRR-1) will be put is neutron activation analysis. The activation analyst requires information about the flux level at various points within and around the reactor core to enable him identify the point of optimum flux (at a given operating power) for any irradiation ...

  12. Nuclear reactor and materials science research: Technical report, May 1, 1985-September 30, 1986

    International Nuclear Information System (INIS)

    1987-01-01

    Throughout the 17-month period of its grant, May 1, 1985-September 30, 1986, the MIT Research Reactor (MITR-II) was operated in support of research and academic programs in the physical and life sciences and in related engineering fields. The reactor was operated 4115 hours during FY 1986 and for 6080 hours during the entire 17-month period, an average of 82 hours per week. Utilization of the reactor during that period may be classified as follows: neutron beam tube research; nuclear materials research and development; radiochemistry and trace analysis; nuclear medicine; radiation health physics; computer control of reactors; dose reduction in nuclear power reactors; reactor irradiations and services for groups outside MIT; MIT Research Reactor. Data on the above utilization for FY 1986 show that the MIT Nuclear Reactor Laboratory (NRL) engaged in joint activities with nine academic departments and interdepartmental laboratories at MIT, the Charles Stark Draper Laboratory in Cambridge, and 22 other universities and nonprofit research institutions, such as teaching hospitals

  13. Lawsuits concerning nuclear power generation in FRG

    International Nuclear Information System (INIS)

    Saito, Osamu

    1980-01-01

    The confirmation of the courts of justice is required for the permission of power stations. This proposition is not in the laws in FRG, but in view of the recent judicatory regulation, it seems to be the norm established experimentally. From the character of German nation, more than 40 specialists and the committees independent of administration take part in the procedure of administrative permission, but considering the temporary procedure, the processes of five classes of courts join in these. Based on the background of such situation, the author outlined the traditional practice in the legislation and administration in the field of nuclear power generation, then investigated into the decisions of Freiburg and Wuerzburg courts of administrative litigation in 1977 and the decision of the federal constitutional court in 1978. Confronting the same technology of light water reactors, the Freiburg court said that the device protecting from the burst of a pressure vessel is necessary, but the Wuerzburg court did not demand it. The confrontations similar to it were seen in the requirements for the utilization of radioactive substances and the final storage of them. The recent decision of the federal constitutional court is concerned with FBRs, and the court discussed the problem of ''residual risks''. The studies on the German decisions are useful for Japan. (Kako, I.)

  14. IEA-R1 research reactor: operational life extension and considerations regarding future decommissioning

    International Nuclear Information System (INIS)

    Frajndlich, Roberto

    2009-01-01

    The IEA-R1 reactor is a pool type research reactor moderated and cooled by light water and uses graphite and beryllium reflectors. The reactor is located at the Instituto de Pesquisas Energeticas e Nucleares (IPEN-CNEN/SP), in the city of Sao Paulo, Brazil. It is the oldest research reactor in the southern hemisphere and one of the oldest of this kind in the world. The first criticality of the reactor was obtained on September 16, 1957. Given the fact that Brazil does not have yet a definitive radioactive waste repository and a national policy establishing rules for the spent fuel storage, the institutions which operate the research reactors for more than 50 years in the country have searched internal solutions for continued operation. This paper describes the spent fuel assemblies and radioactive waste management process for the IEA-R1 reactor and the refurbishment and modernization program adopted to extend its lifetime. Some considerations about the future decommissioning of the reactor are also discussed which, in my opinion, might help the operating organization to make decisions about financial, legal and technical aspects of the decommissioning procedures in a time frame of 10-15 years(author)

  15. Thai research reactor

    International Nuclear Information System (INIS)

    Aramrattana, M.

    1987-01-01

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

  16. Dynamics and inherent safety features of small modular high temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Harrington, R.M.; Ball, S.J.; Cleveland, J.C.

    1986-01-01

    Investigations were made at Oak Ridge National Laboratory to characterize the dynamics and inherent safety features of various modular high temperature gas-cooled reactor (HTGR) designs. This work was sponsored by the US Nuclear Regulatory Commission's HTGR Safety Research program. The US Department of Energy (DOE) and the Gas Cooled Reactor Associates (GCRA) have sponsored studies of several modular HTGR concepts, each having it own unique advantageous economic and inherent safety features. The DOE design team has recently choses a 350-MW(t) annular core with prismatic, graphite matrix fuel for its reference plant. The various safety features of this plant and of the pebble-bed core designs similar to those currently being developed and operated in the Federal Republic of Germany (FRG) are described. A varity of postulated accident sequences involving combinations of loss of forced circulation of the helium primary coolant, loss of primary coolant pressurization, and loss of normal and backup heat sinks were studied and are discussed. Results demonstrate that each concept can withstand an uncontrolled heatup accident without reaching excessive peak fuel temperatures. Comparisons of calculated and measured response for a loss of forced circulation test on the FRG reactor, AVR, are also presented. 10 refs

  17. Instrumentation renewal at the FIR 1 research reactor in Finland

    International Nuclear Information System (INIS)

    Bars, Bruno; Kall, Leif

    1982-01-01

    The Finnish TRIGA Mark II reactor (FIR 1 100 kW, later 250 kW steady state power and pulsing capability up to 250 MW) has been in operation for 20 years. The reactor is the only research reactor in Finland and is an important research training and service facility, which obviously will be operated for 10...20 years ahead. The mechanical parts of the reactor are in good shape. Some minor modifications have previously been made in the instrumentation. However, the original instrumentation could hardly have been used for 10...20 years ahead without extensive modifications and modernization. After a careful evaluation and planning process the whole reactor instrumentation was renewed in 1981 at a cost of about 400 000 dollar. The renewal was carried out in cooperation with the Central Research Institute for Physics (KFKI) at the Hungarian Academy of Sciences, which delivered the nuclear part of the instrumentation and with the Finnish company Valmet Oy Instrument Works, which delivered the conventional instrumentation, including the automatic power control system and the control console. The instrumentation, which is located in-a new isolated control room is based on modern industrial standard modular units with standardized signal ranges, electronic testing possibilities, galvanically isolated outputs etc. The instrument renewal project was brought successfully to completion in November 1981 after only about 10 working days of shut down time. The reactor is now in routine operation and the experiences gained from the new instrumentation are excellent. (author)

  18. Nuclear research reactors in Brazil

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

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

  19. Nuclear research reactors in Brazil

    International Nuclear Information System (INIS)

    Cota, Anna Paula Leite; Mesquita, Amir Zacarias

    2011-01-01

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

  20. Neutronic studies in the enrichment reduction of research reactor IEAR-1

    International Nuclear Information System (INIS)

    Maiorino, J.R.; Fanaro, L.C.B.; Mai, L.A.; Ferreira, P.S.B.; Garone, J.G.M.

    1987-01-01

    In the present work the codes used by the Reactor Physics Division of IPEN-CNEN-SP in calculations for plate-type reactors are described analyzing research reactor IEAR-1. The IAEA model problem for a plate-type reactor 10 MW with high, medium and low enrichment is solved through different methodologies now in use at the RTF/IPEN-CNEN-SP (HAMMER and HAMMER-TECH-CITATION and LEO4-2DBP-UM) looking into the calculation capability for high to low enrichment conversion within the contract held with the IAEA (BRA-4661). Finally, present reactor configuration calculations are compared with experimental measurements with the aim to validate the calculation method. (Author)

  1. Modelling of protective actions in the German Risk Study (FRG)

    International Nuclear Information System (INIS)

    Burkart, A.K.

    1981-01-01

    An emergency response model for nuclear accidents has to allow for a great number of widely different emergency conditions. In addition, it should be compatible with the pertinent laws, regulations, ordinances, guidelines, criteria and reference levels. The German (FRG) guidelines are basic and flexible rather than precise, many decisions being left to the emergency management. In the Risk Study these decisions had to be anticipated. After a brief discussion of the basis of the emergency response model employed in the German Risk Study (FRG), the essential requirements to be met are listed. The main part of the paper deals with the rationale and specification of protective actions. As a result of the calculations the numbers of persons and sizes of areas involved in protective actions are presented. The last section deals with the variation of input data. (author)

  2. Digital control of research reactors

    International Nuclear Information System (INIS)

    Crump, J.C. III.; Richards, W.J.; Heidel, C.C.

    1991-01-01

    Research reactors provide an important service for the nuclear industry. Developments and innovations used for research reactors can be later applied to larger power reactors. Their relatively inexpensive cost allows research reactors to be an excellent testing ground for the reactors of tomorrow. One area of current interest is digital control of research reactor systems. Digital control systems offer the benefits of implementation and superior system response over their analog counterparts. At McClellan Air Force Base in Sacramento, California, the Stationary Neutron Radiography System (SNRS) uses a 1,000-kW TRIGA reactor for neutron radiography and other nuclear research missions. The neutron radiography beams generated by the reactor are used to detect corrosion in aircraft structures. While the use of the reactor to inspect intact F-111 wings is in itself noteworthy, there is another area in which the facility has applied new technology: the instrumentation and control system (ICS). The ICS developed by General Atomics (GA) contains several new and significant items: (a) the ability to servocontrol on three rods, (b) the ability to produce a square wave, and (c) the use of a software configurator to tune parameters affected by the actual reactor core dynamics. These items will probably be present in most, if not all, future research reactors. They were developed with increased control and overall usefulness of the reactor in mind

  3. Advanced development and operating experience with a canned motor pump under pressurized water reactor conditions

    International Nuclear Information System (INIS)

    Dittmer, H.; Reymann, A.; Seibig, B.; Reinecker, H.

    1988-01-01

    At the research reactor FRG-2, Geesthacht, an irradiation device is in operation for testing defective light-water-reactor (LWR) test fuel rods under pressurized water reactor conditions (320 0 C, 160 bar). The requirements to the canned motor pump for cooling water circulation: medium: Demineralized water, operating temperature 320 0 C, operating pressure 155 bar, radiation field of the reactor, integration in the irradiation capsule, helium leak rate -6 mbar.dm 3 .s -1 , minimum working life 3000 hours, were high and caused difficulties in the acquisition of this component. First test runs with supplied pumps showed that the desired working life could not be achieved. The results of the development steps, the test runs, and the performance in service show that for our range of applications, the best combination of materials for the radial bearings is silicon-infiltrated SiC (8% free Si) against the same material. These bearings allowed a good working life for the pump to be achieved. (orig./GL) [de

  4. Planning the Decommissioning of Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Podlaha, J., E-mail: pod@ujv.cz [Nuclear Research Institute Rez, 25068 Rez (Czech Republic)

    2013-08-15

    In the Czech Republic, three research nuclear reactors are in operation. According to the valid legislation, preliminary decommissioning plans have been prepared for all research reactors in the Czech Republic. The decommissioning plans shall be updated at least every 5 years. Decommissioning funds have been established and financial resources are regularly deposited. Current situation in planning of decommissioning of research reactors in the Czech Republic, especially planning of decommissioning of the LVR-15 research reactor is described in this paper. There appeared new circumstances having wide impact on the decommissioning planning of the LVR-15 research reactor: (1) Shipment of spent fuel to the Russian Federation for reprocessing and (2) preparation of processing of radioactive waste from reconstruction of the VVR-S research reactor (now LVR-15 research reactor). The experience from spent fuel shipment to the Russian Federation and from the process of radiological characterization and processing of radioactive waste from reconstruction of the VVR-S research reactor (now the LVR-15 research reactor) and the impact on the decommissioning planning is described in this paper. (author)

  5. Nuclear data usage for research reactors

    International Nuclear Information System (INIS)

    Nakano, Yoshihiro; Soyama, Kazuhiko; Amano, Toshio

    1996-01-01

    In the department of research reactor, many neutronics calculations have been performed to construct, to operate and to modify research reactors of JAERI with several kinds of nuclear data libraries. This paper presents latest two neutronic analyses on research reactors. First one is design work of a low enriched uranium (LEU) fuel for JRR-4 (Japan Research Reactor No.4). The other is design of a uranium silicon dispersion type (silicide) fuel of JRR-3M (Japan Research Reactor No.3 Modified). Before starting the design work, to estimate the accuracy of computer code and calculation method, experimental data are calculated with several nuclear data libraries. From both cases of calculations, it is confirmed that JENDL-3.2 gives about 1 %Δk/k higher excess reactivity than JENDL-3.1. (author)

  6. IEA-R1 Nuclear Research Reactor: 58 Years of Operating Experience and Utilization for Research, Teaching and Radioisotopes Production

    Energy Technology Data Exchange (ETDEWEB)

    Cardenas, Jose Patricio Nahuel; Filho, Tufic Madi; Saxena, Rajendra; Filho, Walter Ricci [Nuclear and Energy Research Institute, IPEN-CNEN/SP, Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP, Av. Prof. Lineu Prestes 2242 Cid Universitaria CEP: 05508-000- Sao Paulo-SP (Brazil)

    2015-07-01

    IEA-R1 research reactor at the Instituto de Pesquisas Energeticas e Nucleares (Nuclear and Energy Research Institute) IPEN, Sao Paulo, Brazil is the largest power research reactor in Brazil, with a maximum power rating of 5 MWth. It is being used for basic and applied research in the nuclear and neutron related sciences, for the production of radioisotopes for medical and industrial applications, and for providing services of neutron activation analysis, real time neutron radiography, and neutron transmutation doping of silicon. IEA-R1 is a swimming pool reactor, with light water as the coolant and moderator, and graphite and beryllium as reflectors. The reactor was commissioned on September 16, 1957 and achieved its first criticality. It is currently operating at 4.5 MWth with a 60-hour cycle per week. In the early sixties, IPEN produced {sup 131}I, {sup 32}P, {sup 198}Au, {sup 24}Na, {sup 35}S, {sup 51}Cr and labeled compounds for medical use. During the past several years, a concerted effort has been made in order to upgrade the reactor power to 5 MWth through refurbishment and modernization programs. One of the reasons for this decision was to produce {sup 99}Mo at IPEN. The reactor cycle will be gradually increased to 120 hours per week continuous operation. It is anticipated that these programs will assure the safe and sustainable operation of the IEA-R1 reactor for several more years, to produce important primary radioisotopes {sup 99}Mo, {sup 125}I, {sup 131}I, {sup 153}Sm and {sup 192}Ir. Currently, all aspects of dealing with fuel element fabrication, fuel transportation, isotope processing, and spent fuel storage are handled by IPEN at the site. The reactor modernization program is slated for completion by 2015. This paper describes 58 years of operating experience and utilization of the IEA-R1 research reactor for research, teaching and radioisotopes production. (authors)

  7. Effect of FRG-8813, a new-type histamine H(2)-receptor antagonist, on the recurrence of gastric ulcer after healing by drug treatment in rats.

    Science.gov (United States)

    Ajioka, H; Miyake, H; Matsuura, N

    2000-08-01

    We investigated the recurrence of ulcers in rats after treatment with FRG-8813, (+/-)-2-(furfurylsulfinyl)-N-[4- [4-(piperidinomethyl)-2-pyridyl] oxy-(Z)-2-butenyl] acetamide, a novel histamine H(2)-receptor antagonist. Chronic gastric ulcers were induced by serosa-searing with a hot metal bar, and the ulcer healing and recurrence after treatment with FRG-8813 or famotidine were evaluated by endoscopy for 160 days. At the dose of 30 mg/kg p. o., once daily, the treatment with FRG-8813 or famotidine for 60 days, which was stopped earlier if the ulcer had healed, accelerated the ulcer healing significantly. A subsequent follow-up study on the healed rats showed that the cumulative recurrence rate of rats healed by FRG-8813 was lower than that of naturally healed rats or rats healed by famotidine. In many cases of rats healed by FRG-8813, the regenerated mucosa was normal in contrast with the control of famotidine-healed animals. The mucosal regeneration index of the gastric ulcer after 10 days' administration of FRG-8813 was significantly higher than that obtained with famotidine. After cessation of the treatment with famotidine for 7 days, rebound hyperacidity was induced; but such rebound did not occur with FRG-8813. Considering the low recurrence rate of ulcers after FRG-8813 treatment, we suggest that FRG-8813 treatment may provide additional benefits in peptic ulcer therapy. Copyright 2000 S. Karger AG, Basel

  8. Core calculations for the upgrading of the IEA-R1 research reactor

    International Nuclear Information System (INIS)

    Santos, Adimir dos; Perrotta, Jose A.; Bastos, Jose Luis F.; Yamaguchi, Mitsuo; Umbehaun, Pedro E.

    1998-01-01

    The IEA-R1 Research Reactor is a multipurpose reactor. It has been used for basic and applied research in the nuclear area, training and radioisotopes production since 1957. In 1995, the Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP) took the decision to modernize and upgrade the power from 2 to 5 MW and increase the operational cycle. This work presents the design requirements and the calculations effectuated to reach this goal. (author)

  9. Mimic of OSU research reactor

    International Nuclear Information System (INIS)

    Lu, Hong; Miller, D.W.

    1991-01-01

    The Ohio State University research reactor (OSURR) is undergoing improvements in its research and educational capabilities. A computer-based digital data acquisition system, including a reactor system mimic, will be installed as part of these improvements. The system will monitor the reactor system parameters available to the reactor operator either in digital parameters available to the reactor operator either in digital or analog form. The system includes two computers. All the signals are sent to computer 1, which processes the data and sends the data through a serial port to computer 2 with a video graphics array VGA monitor, which is utilized to display the mimic system of the reactor

  10. Safety of research reactors (Design and Operation)

    International Nuclear Information System (INIS)

    Dirar, H. M.

    2012-06-01

    The primary objective of this thesis is to conduct a comprehensive up-to-date literature review on the current status of safety of research reactor both in design and operation providing the future trends in safety of research reactors. Data and technical information of variety selected historical research reactors were thoroughly reviewed and evaluated, furthermore illustrations of the material of fuel, control rods, shielding, moderators and coolants used were discussed. Insight study of some historical research reactors was carried with considering sample cases such as Chicago Pile-1, F-1 reactor, Chalk River Laboratories,. The National Research Experimental Reactor and others. The current status of research reactors and their geographical distribution, reactor category and utilization is also covered. Examples of some recent advanced reactors were studied like safety barriers of HANARO of Korea including safety doors of the hall and building entrance and finger print identification which prevent the reactor from sabotage. On the basis of the results of this research, it is apparent that a high quality of safety of nuclear reactors can be attained by achieving enough robust construction, designing components of high levels of efficiency, replacing the compounds of the reactor in order to avoid corrosion and degradation with age, coupled with experienced scientists and technical staffs to operate nuclear research facilities.(Author)

  11. Nuclear material control at IEA-R1 nuclear research reactor

    International Nuclear Information System (INIS)

    1988-01-01

    The control measurements system and verification of physical inventory for fuel elements used in the operation of IEA-R1 nuclear research reactor are described. The computer code used for burn-up calculation are shown. (E.G.) [pt

  12. Thermal hydraulic analysis of the IPR-R1 TRIGA research reactor using a RELAP5 model

    International Nuclear Information System (INIS)

    Costa, Antonella L.; Reis, Patricia Amelia L.; Pereira, Claubia; Veloso, Maria Auxiliadora F.; Mesquita, Amir Z.; Soares, Humberto V.

    2010-01-01

    The RELAP5 code is widely used for thermal hydraulic studies of commercial nuclear power plants. Current investigations and code adaptations have demonstrated that the RELAP5 code can be also applied for thermal hydraulic analysis of nuclear research reactors with good predictions. Therefore, as a contribution to the assessment of RELAP5/MOD3.3 for research reactors analysis, this work presents steady-state and transient calculation results performed using a RELAP5 model to simulate the IPR-R1 TRIGA research reactor at 50 kilowatts (kW) of power operation. The reactor is located in the Nuclear Technology Development Center (CDTN), Brazil. It is a 250 kW, light water moderated and cooled, graphite-reflected, open pool type research reactor. The development and the assessment of a RELAP5 model for the IPR-R1 TRIGA are presented. Experimental data were considered in the process of the RELAP5 model validation. The RELAP5 results were also compared with calculated data from the STHIRP-1 (Research Reactors Thermal Hydraulic Simulation) code. The results obtained have shown that the RELAP5 model for the IPR-R1 TRIGA reproduces the actual steady-state reactor behavior in good agreement with the available data.

  13. International topical meeting. Research Reactor Fuel Management (RRFM) and meeting of the International Group on Reactor Research (IGORR)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2007-07-01

    Nuclear research and test reactors have been in operation for over 60 years, over 270 research reactors are currently operating in more than 50 countries. This meeting is dedicated to different aspects of research reactor fuels: new fuels for new reactors, the conversion to low enriched uranium fuels, spent fuel management and computational tools for core simulation. About 80 contributions are reported in this document, they are organized into 7 sessions: 1) international topics and overview on new projects and fuel, 2) new projects and upgrades, 3) fuel development, 4) optimisation and research reactor utilisation, 5) innovative methods in research reactors physics, 6) safety, operation and research reactor conversion, 7) fuel back-end management, and a poster session. Experience from Australian, Romanian, Libyan, Syrian, Vietnamese, South-African and Ghana research reactors are reported among other things. The Russian program for research reactor spent fuel management is described and the status of the American-driven program for the conversion to low enriched uranium fuels is presented. (A.C.)

  14. Utilization of research reactors - A global perspective

    International Nuclear Information System (INIS)

    Muranaka, R.G.

    1988-01-01

    This paper presents 1) a worldwide picture of research reactors, operable, shutdown, under construction and planned, 2) statistics on utilization of research reactors including TRIGA reactors, and 3) some results of a survey conducted during 1988 on the utilization of research reactors in developing Member States in the Asia-Pacific Region

  15. Reactivity feedback coefficients Pakistan research reactor-1 using PRIDE code

    Energy Technology Data Exchange (ETDEWEB)

    Mansoor, Ali; Ahmed, Siraj-ul-Islam; Khan, Rustam [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan). Dept. of Nuclear Engineering; Inam-ul-Haq [Comsats Institute of Information Technology, Islamabad (Pakistan). Dept. of Physics

    2017-05-15

    Results of the analyses performed for fuel, moderator and void's temperature feedback reactivity coefficients for the first high power core configuration of Pakistan Research Reactor - 1 (PARR-1) are summarized. For this purpose, a validated three dimensional model of PARR-1 core was developed and confirmed against the reference results for reactivity calculations. The ''Program for Reactor In-Core Analysis using Diffusion Equation'' (PRIDE) code was used for development of global (3-dimensional) model in conjunction with WIMSD4 for lattice cell modeling. Values for isothermal fuel, moderator and void's temperature feedback reactivity coefficients have been calculated. Additionally, flux profiles for the five energy groups were also generated.

  16. Reactor Materials Research

    Energy Technology Data Exchange (ETDEWEB)

    Van Walle, E

    2002-04-01

    The activities of SCK-CEN's Reactor Materials Research Department for 2001 are summarised. The objectives of the department are: (1) to evaluate the integrity and behaviour of structural materials used in nuclear power industry; (2) to conduct research to unravel and understand the parameters that determine the material behaviour under or after irradiation; (3) to contribute to the interpretation, the modelling of the material behaviour and to develop and assess strategies for optimum life management of nuclear power plant components. The programmes within the department are focussed on studies concerning (1) Irradiation Assisted Stress Corrosion Cracking (IASCC); (2) nuclear fuel; and (3) Reactor Pressure Vessel Steel.

  17. Reactor Materials Research

    International Nuclear Information System (INIS)

    Van Walle, E.

    2002-01-01

    The activities of SCK-CEN's Reactor Materials Research Department for 2001 are summarised. The objectives of the department are: (1) to evaluate the integrity and behaviour of structural materials used in nuclear power industry; (2) to conduct research to unravel and understand the parameters that determine the material behaviour under or after irradiation; (3) to contribute to the interpretation, the modelling of the material behaviour and to develop and assess strategies for optimum life management of nuclear power plant components. The programmes within the department are focussed on studies concerning (1) Irradiation Assisted Stress Corrosion Cracking (IASCC); (2) nuclear fuel; and (3) Reactor Pressure Vessel Steel

  18. Assessment of a RELAP5 model for the IPR-R1 TRIGA research reactor

    International Nuclear Information System (INIS)

    Reis, Patricia A.L.; Costa, Antonella L.; Pereira, Claubia; Veloso, Maria A.F.; Mesquita, Amir Z.; Soares, Humberto V.

    2010-01-01

    RELAP5 code was developed at the Idaho National Environmental and Engineering Laboratory and it is widely used for thermal hydraulic studies of commercial nuclear power plants and, currently, it has been also applied for thermal hydraulic analysis of nuclear research systems with good predictions. This work is a contribution to the assessment of RELAP5/3.3 code for research reactors analysis. It presents steady-state and transient calculation results performed using a RELAP5 model to simulate the IPR-R1 TRIGA research reactor conditions operating at 50 and 100 kW. The reactor is located at the Nuclear Technology Development Centre (CDTN), Brazil. The development and the assessment of a RELAP5 model for the IPR-R1 TRIGA are presented. Experimental data were considered in the process of code-to-data validation. The RELAP5 results were also compared with calculation performed using the STHIRP-1 (Research Reactors Thermal Hydraulic Simulation) code. The use of a cross flow model has been essential to improve results in the transient condition respect to preceding investigations.

  19. Radiochemical analysis in the nuclear research establishment (KFA) Juelich, FRG

    International Nuclear Information System (INIS)

    Anon.

    1975-01-01

    KFA Juelich is one of the two great nuclear research centres of the Federal Republic of Germany. About 3700 employees including about 700 scientists are engaged in a great number of programs and projects belonging to six main fields of research and development: high temperature reactor and energy techniques; nuclear fusion; properties of materials; materials research; life and environment; methods. In the article the radiochemical analysis work of the former Central Institute of Analytical Chemistry and its two successors is described: activation analysis, application of tracer techniques, fission product analysis. Further on the irradiation facilities are described, a short survey is given on the instrumentation, and the future work is outlined. (T.G.)

  20. New about research reactors

    International Nuclear Information System (INIS)

    Egorenkov, P.M.

    2001-01-01

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

  1. Impact of proposed research reactor standards on reactor operation

    Energy Technology Data Exchange (ETDEWEB)

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

    1974-07-01

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

  2. Impact of proposed research reactor standards on reactor operation

    International Nuclear Information System (INIS)

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

    1974-01-01

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

  3. Simulation of channel blockage for the IEA-R1 research reactor using RELAP/MOD 3

    International Nuclear Information System (INIS)

    Oliveira, Eduardo C.F. de; Castrillo, Lazara Silveira

    2015-01-01

    Research reactors have great importance in the area of nuclear technology, such as radioisotope production, research in nuclear physics, development of new technologies and staff training for reactor operation. The IEA-R1 is a Brazilian research reactor type pool, located at the IPEN (Instituto de Pesquisas Energeticas e Nucleares). In this work is simulated with computer code RELAP5 / MOD 3.3.2 gamma, the effect caused by partial and complete blockage of a channel in MTR fuel element of the IEA-R1 core, in order to analyzed the thermal hydraulic parameters on adjacent channels. (author)

  4. Simulation of channel blockage for the IEA-R1 research reactor using RELAP/MOD 3

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, Eduardo C.F. de; Castrillo, Lazara Silveira, E-mail: ecfoliveira@hotmail.com, E-mail: lazara.castrillo@upe.br [Universidade de Pernambuco (UPE), Recife, PE (Brazil). Escola Politecnica de Pernambuco

    2015-07-01

    Research reactors have great importance in the area of nuclear technology, such as radioisotope production, research in nuclear physics, development of new technologies and staff training for reactor operation. The IEA-R1 is a Brazilian research reactor type pool, located at the IPEN (Instituto de Pesquisas Energeticas e Nucleares). In this work is simulated with computer code RELAP5 / MOD 3.3.2 gamma, the effect caused by partial and complete blockage of a channel in MTR fuel element of the IEA-R1 core, in order to analyzed the thermal hydraulic parameters on adjacent channels. (author)

  5. Research reactor support

    International Nuclear Information System (INIS)

    2005-01-01

    Research reactors (RRs) have been used in a wide range of applications including nuclear power development, basic physics research, education and training, medical isotope production, geology, industry and other fields. However, many research reactors are fuelled with High Enriched Uranium (HEU), are underutilized and aging, and have significant quantities of spent fuel. HEU inventories (fresh and spent) pose security risks Unavailability of a high-density-reprocessable fuel hinders conversion and limits back-end options and represents a survival dilemma for many RRs. Improvement of interim spent fuel storage is required at some RRs. Many RRs are under-utilized and/or inadequately funded and need to find users for their services, or permanently shut down and eventually decommission. Reluctance to decommission affect both cost and safety (loss of experienced staff ) and many shut down but not decommissioned RR with fresh and/or spent fuel at the sites invoke serious concern. The IAEA's research reactor support helps to ensure that research reactors can be operated efficiently with fuels and targets of lower proliferation and security concern and that operators have appropriate technology and options to manage RR fuel cycle issues, especially on long term interim storage of spent research reactor fuel. Availability of a high-density-reprocessable fuel would expand and improve back end options. The International Atomic Energy Agency provides assistance to Member States to convert research reactors from High Enriched Uranium fuel and targets (for medical isotope production) to qualified Low Enriched Uranium fuel and targets while maintaining reactor performance levels. The assistance includes provision of handbooks and training in the performance of core conversion studies, advice for the procurement of LEU fuel, and expert services for LEU fuel acceptance. The IAEA further provides technical and administrative support for countries considering repatriation of its

  6. Thermal hydraulic and safety analyses for Pakistan Research Reactor-1

    International Nuclear Information System (INIS)

    Bokhari, I.H.; Israr, M.; Pervez, S.

    1999-01-01

    Thermal hydraulic and safety analysis of Pakistan Research Reactor-1 (PARR-1) utilizing low enriched uranium (LEU) fuel have been performed using computer code PARET. The present core comprises of 29 standard and 5 control fuel elements. Results of the thermal hydraulic analysis show that the core can be operated at a steady-state power level of 10 MW for a flow rate of 950 m 3 /h, with sufficient safety margins against ONB (onset of nucleate boiling) and DNB (departure from nucleate boiling). Safety analysis has been carried out for various modes of reactivity insertions. The events studied include: start-up accident; accidental drop of a fuel element in the core; flooding of a beam tube with water; removal of an in-pile experiment during reactor operation etc. For each of these transients, time histories of reactor power, energy released and clad surface temperature etc. were calculated. The results indicate that the peak clad temperatures remain well below the clad melting temperature during these accidents. It is therefore concluded that the reactor can be safely operated at 10 MW without compromising safety. (author)

  7. Thermal-hydraulic modelling of the SAFARI-1 research reactor using RELAP/SCDAPSIM/MOD3.4

    International Nuclear Information System (INIS)

    Sekhri, Abdelkrim; Graham, Andy; D'Arcy, Alan; Oliver, Melissa

    2008-01-01

    The SAFARI-1 reactor is a tank-in-pool MTR type research reactor operated at a nominal core power of 20 MW. It operates exclusively in the single phase liquid water regime with nominal water and fuel temperatures not exceeding 100 deg. C. RELAP/SCDAPSIM/MOD3.4 is a Best Estimate Code for light water reactors as well as for low pressure transients, as part of the code validation was done against low pressure facilities and research reactor experimental data. The code was used to simulate SAFARI-1 in normal and abnormal operation and validated against the experimental data in the plant and was used extensively in the upgrading of the Safety Analysis Report (SAR) of the reactor. The focus of the following study is the safety analysis of the SAFARI-1 research reactor and describes the thermal hydraulic modelling and analysis approach. Particular emphasis is placed on the modelling detail, the application of the no-boiling rule and predicting the Onset of Nucleate Boiling and Departure from Nucleate Boiling under Loss of Flow conditions. Such an event leads the reactor to switch to a natural convection regime which is an adequate mode to maintain the clad and fuel temperature within the safety margin. It is shown that the RELAP/SCDAPSIM/MOD3.4 model can provide accurate predictions as long as the clad temperature remains below the onset of nucleate boiling temperature and the DNB ratio is greater than 2. The results are very encouraging and the model is shown to be appropriate for the analysis of SAFARI-1 research reactor. (authors)

  8. FRG sealed isotopic heat sources project (C-229) project management plan

    International Nuclear Information System (INIS)

    Metcalf, I.L.

    1997-01-01

    This Project Management Plan defines the cost, scope, schedule, organizational responsibilities, and work breakdown structure for the removal of the Federal Republic of Germany (FRG) Sealed Isotopic Heat Sources from the 324 Building and placed in interim storage at the Central Waste Complex (CWC)

  9. On the research activities in reactor and neutron physics using the first egyptian research reactor

    International Nuclear Information System (INIS)

    Hassan, A.M.

    2000-01-01

    A review on the most important research activities in reactor and neutron physics using the first Egyptian Research Reactor (ET-RR-1) is given. An out look on: neutron cross-sections, neutron flux, neutron capture gamma-ray spectroscopy, neutron activation analysis, neutron diffraction and radiation shielding experiments, is presented

  10. Commissioning of research reactors. Safety guide

    International Nuclear Information System (INIS)

    2006-01-01

    The objective of this Safety Guide is to provide recommendations on meeting the requirements for the commissioning of research reactors on the basis of international best practices. Specifically, it provides recommendations on fulfilling the requirements established in paras 6.44 and 7.42-7.50 of International Atomic Energy Agency, Safety of Research Reactors, IAEA Safety Standards Series No. NS-R-4, IAEA, Vienna (2005) and guidance and specific and consequential recommendations relating to the recommendations presented in paras 615-621 of International Atomic Energy Agency, Safety in the Utilization and Modification of Research Reactors, Safety Series No. 35-G2, IAEA, Vienna (1994) and paras 228-229 of International Atomic Energy Agency, Safety Assessment of Research Reactors and Preparation of the Safety Analysis Report, Safety Series No. 35-G1, IAEA, Vienna (1994). This Safety Guide is intended for use by all organizations involved in commissioning for a research reactor, including the operating organization, the regulatory body and other organizations involved in the research reactor project

  11. Proceedings of the international meeting on research and test reactor core conversions from HEU to LEU fuels

    Energy Technology Data Exchange (ETDEWEB)

    Travelli, A [Argonne National Laboratory, Argonne, IL (United States)

    1983-09-01

    Conversion of research and test reactor cores from the use of high enrichment uranium to the use of low enrichment uranium depends on the cooperation of many research organizations, reactor operators, and government agencies. At a technical level, it involves almost all aspects of the fuel cycle, including fuel development, testing, shipping and reprocessing; experiment performance; economics; and safety and licensing aspects. The reactors involved and the conversion activities are distributed among approximately 25 countries, making this a subject which is best dealt with on an international basis. To foster direct communication in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at the Argonne National Laboratory, sponsored this meeting as the fifth of a series which began in 1978. The previous meetings were held at Argonne (International Meeting of Research Reactor Fuel Designers, Developers, and Fabricators, Argonne National Laboratory, Argonne, U.S.A., November 910, 1978), at Saclay (IAEA Consultants' Meeting on Research Reactor Core Conversions from HEU to LEU, Centre d'etudes Nucleaires de Saclay, Saclay, France, December 12-14, 1979), at Argonne (International Meeting on Development, Fabrication and Application of Reduced Enrichment Fuels for Research and Test Reactors, Argonne National Laboratory, Argonne, U.S.A., November 12-14, 1980) and at Juelich (Seminar on Research Reactor Operation and Use, Juelich Nuclear Research Center, Juelich, F.R.G., September 48, 1981). Proceedings from the two most recent previous meetings were published as ANL/RERTR/TM-3 (CONF-801144) and IAEA-SR-77. The spirit of this meeting differs slightly from that of the previous meetings. The advances which have been made and the growing maturity of the effort have caused a gradual shift of emphasis away from those topics which dominated the floor during the first meetings, such as fuel and methods development, and towards topics which concern more

  12. Proceedings of the international meeting on research and test reactor core conversions from HEU to LEU fuels

    International Nuclear Information System (INIS)

    Travelli, A.

    1983-09-01

    Conversion of research and test reactor cores from the use of high enrichment uranium to the use of low enrichment uranium depends on the cooperation of many research organizations, reactor operators, and government agencies. At a technical level, it involves almost all aspects of the fuel cycle, including fuel development, testing, shipping and reprocessing; experiment performance; economics; and safety and licensing aspects. The reactors involved and the conversion activities are distributed among approximately 25 countries, making this a subject which is best dealt with on an international basis. To foster direct communication in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at the Argonne National Laboratory, sponsored this meeting as the fifth of a series which began in 1978. The previous meetings were held at Argonne (International Meeting of Research Reactor Fuel Designers, Developers, and Fabricators, Argonne National Laboratory, Argonne, U.S.A., November 910, 1978), at Saclay (IAEA Consultants' Meeting on Research Reactor Core Conversions from HEU to LEU, Centre d'etudes Nucleaires de Saclay, Saclay, France, December 12-14, 1979), at Argonne (International Meeting on Development, Fabrication and Application of Reduced Enrichment Fuels for Research and Test Reactors, Argonne National Laboratory, Argonne, U.S.A., November 12-14, 1980) and at Juelich (Seminar on Research Reactor Operation and Use, Juelich Nuclear Research Center, Juelich, F.R.G., September 48, 1981). Proceedings from the two most recent previous meetings were published as ANL/RERTR/TM-3 (CONF-801144) and IAEA-SR-77. The spirit of this meeting differs slightly from that of the previous meetings. The advances which have been made and the growing maturity of the effort have caused a gradual shift of emphasis away from those topics which dominated the floor during the first meetings, such as fuel and methods development, and towards topics which concern more

  13. Physics and safety of advanced research reactors

    International Nuclear Information System (INIS)

    Boening, K.; Hardt, P. von der

    1987-01-01

    Advanced research reactor concepts are presently being developed in order to meet the neutron-based research needs of the nineties. Among these research reactors, which are characterized by an average power density of 1-10 MW per liter, highest priority is now generally given to the 'beam tube reactors'. These provide very high values of the thermal neutron flux (10 14 -10 16 cm -2 s -1 ) in a large volume outside of the reactor core, which can be used for sample irradiations and, in particular, for neutron scattering experiments. The paper first discusses the 'inverse flux trap concept' and the main physical aspects of the design and optimization of beam tube reactors. After that two examples of advanced research reactor projects are described which may be considered as two opposite extremes with respect to the physical optimization principle just mentioned. The present situation concerning cross section libraries and neutronic computer codes is more or less satisfactory. The safety analyses of advanced research reactors can largely be updated from those of current new designs, partially taking advantage of the immense volume of work done for power reactors. The paper indicates a few areas where generic problems for advanced research reactor safety are to be solved. (orig.)

  14. Progress in the neutronic core conversion (HEU-LEU) analysis of Ghana research reactor-1.

    Energy Technology Data Exchange (ETDEWEB)

    Anim-Sampong, S.; Maakuu, B. T.; Akaho, E. H. K.; Andam, A.; Liaw, J. J. R.; Matos, J. E.; Nuclear Engineering Division; Ghana Atomic Energy Commission; Kwame Nkrumah Univ. of Science and Technology

    2006-01-01

    The Ghana Research Reactor-1 (GHARR-1) is a commercial version of the Miniature Neutron Source Reactor (MNSR) and has operated at different power levels since its commissioning in March 1995. As required for all nuclear reactors, neutronic and thermal hydraulic analysis are being performed for the HEU-LEU core conversion studies of the Ghana Research Reactor-1 (GHARR-1) facility, which is a commercial version of the Miniature Neutron Source Reactor (MNSR). Stochastic Monte Carlo particle transport methods and tools (MCNP4c/MCNP5) were used to fine-tune a previously developed 3-D MCNP model of the GHARR-1 facility and perform neutronic analysis of the 90.2% HEU reference and candidate LEU (UO{sub 2}, U{sub 3}Si{sub 2}, U-9Mo) fresh cores with varying enrichments from 12.6%-19.75%. In this paper, the results of the progress made in the Monte Carlo neutronic analysis of the HEU reference and candidate LEU fuels are presented. In particular, a comparative performance assessment of the LEU with respect to neutron flux variations in the fission chamber and experimental irradiation channels are highlighted.

  15. Applied research into direct numerical control of A-1 reactor temperature

    International Nuclear Information System (INIS)

    Karpeta, C.; Volf, K.

    1974-01-01

    Partial results of research efforts aimed at applying modern control theory in the control of the reactor of the A-1 nuclear power station are presented. A mathematical model of the process dynamics was developed. Some parameters of the model were determined using the results of an experimentally performed reactor scram. The optimal stochastic discrete regulator was determined and closed-loop transients were studied. The possibilities of implementing control routines were investigated using the RPP-16 computer. (author)

  16. Introduction to Safety Analysis Approach for Research Reactors

    International Nuclear Information System (INIS)

    Park, Suki

    2016-01-01

    The research reactors have a wide variety in terms of thermal powers, coolants, moderators, reflectors, fuels, reactor tanks and pools, flow direction in the core, and the operating pressure and temperature of the cooling system. Around 110 research reactors have a thermal power greater than 1 MW. This paper introduces a general approach to safety analysis for research reactors and deals with the experience of safety analysis on a 10 MW research reactor with an open-pool and open-tank reactor and a downward flow in the reactor core during normal operation. The general approach to safety analysis for research reactors is described and the design features of a typical open-pool and open-tank type reactor are discussed. The representative events expected in research reactors are investigated. The reactor responses and the thermal hydraulic behavior to the events are presented and discussed. From the minimum CHFR and the maximum fuel temperature calculated, it is ensured that the fuel is not damaged in the step insertion of reactivity by 1.8 mk and the failure of all primary pumps for the reactor with a 10 MW thermal power and downward core flow

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  18. Backfitting of research reactors

    International Nuclear Information System (INIS)

    Delrue, R.; Noesen, T.

    1985-01-01

    The backfitting of research reactors covers a variety of activities. 1. Instrumentation and control: Control systems have developed rapidly and many reactor operators wish to replace obsolete equipment by new systems. 2. Pool liners: Some pools are lined internally with ceramic tiles. These may become pervious with time necessitating replacement, e.g. by a new stainless steel liner. 3. Heat removal system: Deficiencies can occur in one or more of the cooling system components. Upgrading may require modifications of the system such as addition of primary loops, introduction of deactivation tanks, pump replacement. Recent experience in such work has shown that renewal, backfitting and upgrading of an existing reactor is economically attractive since the related costs and delivery times are substantially lower than those required to install a new research reactor

  19. Digital Systems Implemented at the IPEN Nuclear Research Reactor (IEA-R1): Results and Necessities

    International Nuclear Information System (INIS)

    Nahuel-Cardenas, Jose-Patricio; Madi-Filho, Tufic; Ricci-Filho, Walter; Rodrigues-de-Carvalho, Marcos; Lima-Benevenuti, Erion-de; Gomes-Neto, Jose

    2013-06-01

    (Nuclear and Energy Research Institute) was founded in 1956 with the main purpose of doing research and development in the field of nuclear energy and its applications. It is located at the campus of University of Sao Paulo (USP), in the city of Sao Paulo, in an area of nearly 500, 000 m2. It has over 1.000 employees and 40% of them have qualification at master or doctor level The institute is recognized as a national leader institution in research and development (R and D) in the areas of radiopharmaceuticals, industrial applications of radiation, basic nuclear research, nuclear reactor operation and nuclear applications, materials science and technology, laser technology and applications. Along with the R and D, it has a strong educational activity, having a graduate program in Nuclear Technology, in association with the University of Sao Paulo, ranked as the best university in the country. The Federal Government Evaluation institution CAPES, granted to this course grade 6, considering it a program of Excellence. This program started at 1976 and has awarded 458 Ph.D. degrees and 937 master degrees since them. The actual graduate enrollment is around 400 students. One of major nuclear installation at IPEN is the IEA-R1 research reactor; it is the only Brazilian research reactor with substantial power level suitable for its utilization in researches concerning physics, chemistry, biology and engineering as well as for producing some useful radioisotopes for medical and other applications. IEA-R1 reactor is a swimming pool type reactor moderated and cooled by light water and uses graphite and beryllium as reflectors. The first criticality was achieved on September 16, 1957. The reactor is currently operating at 4.5 MW power level with an operational schedule of continuous 64 hours a week. In 1996 a Modernization Program was started to establish recommendations in order to mitigate equipment and structures ageing effects in the reactor components, detect and evaluate

  20. The IAEA programme on research reactor safety

    International Nuclear Information System (INIS)

    Abou Yehia, H.

    2007-01-01

    According to the research reactor database of IAEA (RRDB), 250 reactors are operating worldwide, 248 have been shut down and 170 have been decommissioned. Among the 248 reactors that do not run, some will resume their activities, others will be dismantled and the rest do not face a clear future. The analysis of reported incidents shows that the ageing process is a major cause of failures, more than two thirds of operating reactors are over 30 years old. It also appears that the lack of adequate regulations or safety standards for research reactors is an important issue concerning reactor safety particularly when reactors are facing re-starting or upgrading or modifications. The IAEA has launched a 4-axis program: 1) to set basic safety regulations and standards for research reactors, 2) to provide IAEA members with an efficient help for the application of these safety regulations to their reactors, 3) to foster international exchange of information on research reactor safety, and 4) to provide IAEA members with a help concerning safety issues linked to malicious acts or sabotage on research reactors

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

    International Nuclear Information System (INIS)

    Richards, W.J.

    1980-01-01

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

  2. Future role of hydrogen in FRG

    International Nuclear Information System (INIS)

    Bradke, H.

    1992-01-01

    Relative to the Federal Republic of Germany energy-economy framework, this paper prepares supply and demand assessments for a set of energy source diversification strategy alternatives involving the substantial use of hydrogen fuels, with the aim of reducing the strain on the the earth's limited supplies of fossil fuels and limiting carbon dioxide emissions into the atmosphere. These assessments include forecasts of population dynamics, GNP, and sectoral energy consumption, production, imports and prices for fossil fuels and renewable energy sources. The comparative evaluation of the diversification scenarios includes sensitivity analyses to establish the optimum mix of economy-energy planning criteria that would allow for the successful evolution of a hydrogen based economy in the FRG by the year 2040

  3. Modernization of Safety and Control Instrumentation of the IEA-R1 Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    De Carvalho, P.V., E-mail: paulov@ien.gov.br [Institute of Nuclear Engineering (IEN), National Nuclear Energy Commission (CNEN), Rio de Janeiro (Brazil)

    2014-08-15

    The research reactor IEA-R1 located in the Institute of Energy and Nuclear Research (IPEN), São Paulo, Brazil, obtained its first criticality on 16 September 1957 and since then has served the scientific and medical community in the performance of experiments in applied nuclear physics, as well as the provision of radioisotopes for production of radiopharmaceuticals. The reactor produces radioisotopes {sup 82}Br and {sup 41}Ar for special processes in industrial inspection and {sup 192}Ir and {sup 198}Au as sources of radiation used in brachytherapy, {sup 153}Sm for pain relief in patients with bone metastasis, and calibrated sources of {sup 133}Ba, {sup 137}Cs, {sup 57}Co, {sup 60}Co, {sup 241}Am and {sup 152}Eu used in medical clinics and hospitals practicing nuclear medicine and research laboratories. Services are offered in regular non-destructive testing by neutron radiography, neutron irradiation of silicon for phosphorous doping and other various irradiations with neutrons. The reactor is responsible for producing approximately 70% of radiopharmaceutical {sup 131}I used in Brazil, which saves about US$ 800 000 annually for the country. After more than 50 years of use, most of its equipment and systems have been modernized, and recently the reactor power was increased to 5 MW in order to enhance radioisotope production capability. However, the control room and nuclear instrumentation system used for reactor safety have operated more than 30 years and require constant maintenance. Many equipment and electronic components are obsolete, and replacements are not available in the market. The modernization of the nuclear safety and control instrumentation systems of IEA-R1 is being carried out with consideration for the internationally recognized criteria for safety and reliable reactor operations and the latest developments in nuclear electronic technology. The project for the new reactor instrumentation system specifies three wide range neutron monitoring

  4. Neutronics analysis of Nigerian Research Reactor-1

    International Nuclear Information System (INIS)

    Azande, T.S.; Balogun, G.I.

    2010-01-01

    Feasibility studies for the conversion of the Nigerian Research Reactor-1 (NIRR-1) have been performed using WIMS and CITATION codes (Azande et al, 2009 and Balogun, 2003) at the Centre for Energy Research and Training (CERT), Ahmadu Bello University, Zaria Kaduna State. In this work, the neutronics analysis of NIRR-1 core concerning mass loading of U-235 in the core, shut down margin (SDM), safety reactivity factor (SRF), control rod worth, and control rod critical depth of insertion were investigated at low enrichment. Two fuel types (UAl 4 and UO 2 ) were considered and the uranium densities required for the conversion of NIRR-1 core to low enrichment were computed to be 1201g/cc with 20% enrichment, 1144 g/cc with 19.75% enrichment, 1274 g/cc with 15% enrichment, 1448 g/cc with 10% enrichment for UAl 4 fuel type and 1141g/cc with 20% enrichment, 1144 g/cc with 19.75% enrichment, 1216 g/cc with 15% enrichment, and 1389 g/cc with 10% enrichment for UO 2 fuel type. Signi ficantly, higher uranium densities are required to convert NIRR-1 from HEU to LEU - indicating a drastic review of the NIRR-1 core.

  5. Strengthening IAEA Safeguards for Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Reid, Bruce D. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Anzelon, George A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Budlong-Sylvester, Kory [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-09-01

    During their December 10-11, 2013, workshop in Grenoble France, which focused on the history and future of safeguarding research reactors, the United States, France and the United Kingdom (UK) agreed to conduct a joint study exploring ways to strengthen the IAEA’s safeguards approach for declared research reactors. This decision was prompted by concerns about: 1) historical cases of non-compliance involving misuse (including the use of non-nuclear materials for production of neutron generators for weapons) and diversion that were discovered, in many cases, long after the violations took place and as part of broader pattern of undeclared activities in half a dozen countries; 2) the fact that, under the Safeguards Criteria, the IAEA inspects some reactors (e.g., those with power levels under 25 MWt) less than once per year; 3) the long-standing precedent of States using heavy water research reactors (HWRR) to produce plutonium for weapons programs; 4) the use of HEU fuel in some research reactors; and 5) various technical characteristics common to some types of research reactors that could provide an opportunity for potential proliferators to misuse the facility or divert material with low probability of detection by the IAEA. In some research reactors it is difficult to detect diversion or undeclared irradiation. In addition, infrastructure associated with research reactors could pose a safeguards challenge. To strengthen the effectiveness of safeguards at the State level, this paper advocates that the IAEA consider ways to focus additional attention and broaden its safeguards toolbox for research reactors. This increase in focus on the research reactors could begin with the recognition that the research reactor (of any size) could be a common path element on a large number of technically plausible pathways that must be considered when performing acquisition pathway analysis (APA) for developing a State Level Approach (SLA) and Annual Implementation Plan (AIP). To

  6. RMB. The new Brazilian multipurpose research reactor

    International Nuclear Information System (INIS)

    Perrotta, Jose Augusto; Soares, Adalberto Jose

    2015-01-01

    Brazil has four research reactors (RR) in operation: IEA-R1, a 5 MW pool type RR; IPR-R1, a 100 kW TRIGA type RR; ARGONAUTA, a 500 W Argonaut type RR, and IPEN/MB-01, a 100 W critical facility. The first three were constructed in the 50's and 60's, for teaching, training, and nuclear research, and for many years they were the basic infrastructure for the Brazilian nuclear developing program. The last, IPEN/MB-01, is the result of a national project developed specifically for qualification of reactor physics codes. Considering the relative low power of Brazilian research reactors, with exception of IEAR1, none of the other reactors are feasible for radioisotope production, and even IEA-R1 has a limited capacity. As a consequence, since long ago, 100% of the Mo-99 needed to attend Brazilian nuclear medicine services has been imported. Because of the high dependence on external supply, the international Moly-99 supply crisis that occurred in 2008/2009 affected significantly Brazilian nuclear medicine services, and as presented in previous IAEA events, in 2010 Brazilian government formalized the decision to build a new research reactor. The new reactor named RMB (Brazilian Multipurpose Reactor) will be a 30 MW open pool type reactor, using low enriched uranium fuel. The facility will be part of a new nuclear research centre, to be built about 100 kilometres from Sao Paulo city, in the southern part of Brazil. The new nuclear research centre will have several facilities, to use thermal and cold neutron beams; to produce radioisotopes; to perform neutron activation analysis; and to perform irradiations tests of materials and fuels of interest for the Brazilian nuclear program. An additional facility will be used to store, for at least 100 years, all the fuel used in the reactor. The paper describes the main characteristics of the new centre, emphasising the research reactor and giving a brief description of the laboratories that will be constructed, It also presents the

  7. RMB. The new Brazilian multipurpose research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Perrotta, Jose Augusto; Soares, Adalberto Jose [Comissao Nacional de Energia Nuclear (CNEN) (Brazil)

    2015-01-15

    Brazil has four research reactors (RR) in operation: IEA-R1, a 5 MW pool type RR; IPR-R1, a 100 kW TRIGA type RR; ARGONAUTA, a 500 W Argonaut type RR, and IPEN/MB-01, a 100 W critical facility. The first three were constructed in the 50's and 60's, for teaching, training, and nuclear research, and for many years they were the basic infrastructure for the Brazilian nuclear developing program. The last, IPEN/MB-01, is the result of a national project developed specifically for qualification of reactor physics codes. Considering the relative low power of Brazilian research reactors, with exception of IEAR1, none of the other reactors are feasible for radioisotope production, and even IEA-R1 has a limited capacity. As a consequence, since long ago, 100% of the Mo-99 needed to attend Brazilian nuclear medicine services has been imported. Because of the high dependence on external supply, the international Moly-99 supply crisis that occurred in 2008/2009 affected significantly Brazilian nuclear medicine services, and as presented in previous IAEA events, in 2010 Brazilian government formalized the decision to build a new research reactor. The new reactor named RMB (Brazilian Multipurpose Reactor) will be a 30 MW open pool type reactor, using low enriched uranium fuel. The facility will be part of a new nuclear research centre, to be built about 100 kilometres from Sao Paulo city, in the southern part of Brazil. The new nuclear research centre will have several facilities, to use thermal and cold neutron beams; to produce radioisotopes; to perform neutron activation analysis; and to perform irradiations tests of materials and fuels of interest for the Brazilian nuclear program. An additional facility will be used to store, for at least 100 years, all the fuel used in the reactor. The paper describes the main characteristics of the new centre, emphasising the research reactor and giving a brief description of the laboratories that will be constructed, It also

  8. Experiment on continuous operation of the Brazilian IEA-R1 research reactor

    International Nuclear Information System (INIS)

    Freitas Pintaud, M. de

    1994-01-01

    In order to increase the radioisotope production in the IEA-R1 research reactor at IPEN/CNEN-SP, it has been proposed a change in its operation regime from 8 hours per day and 5 days per week to continuous 48 hours per week. The necessary reactor parameters for this new operation regime were obtained through an experiment in which the reactor was for the first time operated in the new regime. This work presents the principal results from this experiment: xenon reactivity, new shutdown margins, and reactivity loss due to fuel burnup in the new operation regime. (author)

  9. Power Trip Set-points of Reactor Protection System for New Research Reactor

    International Nuclear Information System (INIS)

    Lee, Byeonghee; Yang, Soohyung

    2013-01-01

    This paper deals with the trip set-point related to the reactor power considering the reactivity induced accident (RIA) of new research reactor. The possible scenarios of reactivity induced accidents were simulated and the effects of trip set-point on the critical heat flux ratio (CHFR) were calculated. The proper trip set-points which meet the acceptance criterion and guarantee sufficient margins from normal operation were then determined. The three different trip set-points related to the reactor power are determined based on the RIA of new research reactor during FP condition, over 0.1%FP and under 0.1%FP. Under various reactivity insertion rates, the CHFR are calculated and checked whether they meet the acceptance criterion. For RIA at FP condition, the acceptance criterion can be satisfied even if high power set-point is only used for reactor trip. Since the design of the reactor is still progressing and need a safety margin for possible design changes, 18 MW is recommended as a high power set-point. For RIA at 0.1%FP, high power setpoint of 18 MW and high log rate of 10%pp/s works well and acceptance criterion is satisfied. For under 0.1% FP operations, the application of high log rate is necessary for satisfying the acceptance criterion. Considering possible decrease of CHFR margin due to design changes, the high log rate is suggested to be 8%pp/s. Suggested trip set-points have been identified based on preliminary design data for new research reactor; therefore, these trip set-points will be re-established by considering design progress of the reactor. The reactor protection system (RPS) of new research reactor is designed for safe shutdown of the reactor and preventing the release of radioactive material to environment. The trip set point of RPS is essential for reactor safety, therefore should be determined to mitigate the consequences from accidents. At the same time, the trip set-point should secure margins from normal operational condition to avoid

  10. Multi purpose research reactor

    International Nuclear Information System (INIS)

    Raina, V.K.; Sasidharan, K.; Sengupta, Samiran; Singh, Tej

    2006-01-01

    At present Dhruva and Cirus reactors provide the majority of research reactor based facilities to cater to the various needs of a vast pool of researchers in the field of material sciences, physics, chemistry, bio sciences, research and development work for nuclear power plants and production of radio isotopes. With a view to further consolidate and expand the scope of research and development in nuclear and allied sciences, a new 20 MWt multi purpose research reactor is being designed. This paper describes some of the design features and safety aspects of this reactor

  11. Research reactor job analysis - A project description

    International Nuclear Information System (INIS)

    Yoder, John; Bessler, Nancy J.

    1988-01-01

    Addressing the need of the improved training in nuclear industry, nuclear utilities established training program guidelines based on Performance-Based Training (PBT) concepts. The comparison of commercial nuclear power facilities with research and test reactors owned by the U.S. Department of Energy (DOE), made in an independent review of personnel selection, training, and qualification requirements for DOE-owned reactors pointed out that the complexity of the most critical tasks in research reactors is less than that in power reactors. The U.S. Department of Energy (DOE) started a project by commissioning Oak Ridge Associated Universities (ORAU) to conduct a job analysis survey of representative research reactor facilities. The output of the project consists of two publications: Volume 1 - Research Reactor Job Analysis: Overview, which contains an Introduction, Project Description, Project Methodology,, and. An Overview of Performance-Based Training (PBT); and Volume 2 - Research Reactor Job Analysis: Implementation, which contains Guidelines for Application of Preliminary Task Lists and Preliminary Task Lists for Reactor Operators and Supervisory Reactor Operators

  12. Development of a training simulator to operators of the IEA-R1 research reactor

    International Nuclear Information System (INIS)

    Carvalho, Ricardo Pinto de

    2006-01-01

    This work reports the development of a Simulator for the IEA-R1 Research Reactor. The Simulator was developed with Visual C++ in two stages: construction of the mathematics models and development and configuration of graphics interfaces in a Windows XP executable. A simplified modeling was used for main physics phenomena, using a point kinetics model for the nuclear process and the energy and mass conservation laws in the average channel of the reactor for the thermal hydraulic process. The dynamics differential equations were solved by using finite differences through the 4th order Runge- Kutta method. The reactivity control, reactor cooling, and reactor protection systems were also modeled. The process variables are stored in ASCII files. The Simulator allows navigating by screens of the systems and monitoring tendencies of the operational transients, being an interactive tool for teaching and training of IEA-R1 operators. It also can be used by students, professors, and researchers in teaching activities in reactor and thermal hydraulics theory. The Simulator allows simulations of operations of start up, power maneuver, and shut down. (author)

  13. Research reactors in Austria - Present situation

    International Nuclear Information System (INIS)

    Boeck, H.; Musilek, A.; Villa, M.

    2005-01-01

    In the past decades Austria operated three research reactors, the 10 MW ASTRA reactor at Seibersdorf, the 250 kW TRIGA reactor at the Atominstitut and the 1 kW Argonaut reactor at the Technical University in Graz. Since the shut down of the ASTRA on July 31th, 1999 and its immediate decommissioning reactor and the shut down of the Argonaut reactor in Graz on August 31st, 2004 only one reactor remains operational for keeping nuclear competence in Austria which is the 250 kW TRIGA Mark II reactor. (author)

  14. Reactor safety research program. A description of current and planned reactor safety research sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research

    International Nuclear Information System (INIS)

    1975-06-01

    The reactor safety research program, sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research, is described in terms of its program objectives, current status, and future plans. Elements of safety research work applicable to water reactors, fast reactors, and gas cooled reactors are presented together with brief descriptions of current and planned test facilities. (U.S.)

  15. Coal upgrading based on the high temperature reactor - engineering and economic aspects

    International Nuclear Information System (INIS)

    Barnert, H.; Neis, H.

    1987-01-01

    The development of the high temperature reactor in the FRG was opened with the aim of economically producing electricity which might be achieved with the next project works. Moreover, the HTR supplies process heat of very high temperature which may be used with metallic heat exchangers in the range up to 1000 0 C. From these reasons and regarding the world power prospects and the special power situation in our own country, the FRG began to develop methods of upgrading fossil primary energy sources using high-temperature heat from the HTR about 15 years ago. Meanwhile great progress has been achieved in research, development and test works; the modules are developed to a great extent. Economical service is expected for future utilization of selected processes. Problems to be still accomplished are first of all related to system testing as well as further exhaustion of the HTR's temperature potential. The proposed extension of the AVR to some process heat facility offers the possibility of testing the system in a low-cost and accelerated manner. (author)

  16. Overview of the Dalat Nuclear Research Reactor

    International Nuclear Information System (INIS)

    Nguyen Nhi Dien; Nguyen Thai Sinh; Luong Ba Vien

    2016-01-01

    The present reactor called Dalat Nuclear Research Reactor (DNRR) has been reconstructed from the former TRIGA Mark II reactor which was designed by General Atomic (GA, San Diego, California, USA), started building in early 1960s, put into operation in 1963 and operated until 1968 at nominal power of 250 kW. In 1975, all fuel elements of the reactor were unloaded and shipped back to the USA. The DNRR is a 500-kW pool-type research reactor using light water as both moderator and coolant. The reactor is used as a neutron source for the purposes of: (1) radioactive isotope production; (2) neutron activation analysis; and (3) research and training

  17. CER. Research reactors in France

    International Nuclear Information System (INIS)

    Estrade, Jerome

    2012-01-01

    Networking and the establishment of coalitions between research reactors are important to guarantee a high technical quality of the facility, to assure well educated and trained personnel, to harmonize the codes of standards and the know-ledge of the personnel as well as to enhance research reactor utilization. In addition to the European co-operation, country-specific working groups have been established for many years, such as the French research reactor Club d'Exploitants des Reacteurs (CER). It is the association of French research reactors representing all types of research reactors from zero power up to high flux reactors. CER was founded in 1990 and today a number of 14 research reactors meet twice a year for an exchange of experience. (orig.)

  18. The 33 years of research reactors in JAERI

    International Nuclear Information System (INIS)

    1990-11-01

    The development and utilization of atomic energy in Japan began with the installation of JRR-1 reactor which attained the criticality in August, 1957, thus the third fire was lighted for the first time in Japan. JRR-2 was constructed as a full scale versatile research reactor, which attained the criticality in October, 1960, and since 1962, it has accomplished the role of the reactor for joint utilization. JRR-3 is the first reactor made in Japan by concentrating Japanese technologies in it, to develop and improve Japanese atomic energy technology. It attained the criticality in September, 1962, and has been used as a versatile research reactor. In 1960, Research Reactor Management Department was founded. JRR-4 was constructed as the research reactor for shielding for developing a nuclear-powered ship, which attained the criticality in January, 1965. The first hot laboratory in Japan for carrying out the post-irradiation test on the fuel and materials irradiated in these research reactors was installed in 1961. The JRR-1 was stopped in September, 1968, and is used as the commemorative exhibition hall. The JRR-3 was reconstructed, and attained the criticality in March, 1990, using 20 % enriched uranium fuel. The course of the research reactors for 33 years is reported. (K.I.)

  19. Study on the present and future training of managers and operators for reactor facilities

    International Nuclear Information System (INIS)

    Hagemann, G.; Preuss, W.; Tietze, A.; Wuest, S.

    1973-09-01

    The study gives a survey on the training methods for the operating personnel of reactors in operation or under construction in the FRG and compares them with the training and testing methods of other countries, in particular the USA. (RW/AK) [de

  20. Application of JAERI research reactors to education

    International Nuclear Information System (INIS)

    Ogawa, Shigeru; Morozumi, Minoru

    1987-01-01

    At the dawning of the atomic age in Japan, training on reactor operation and reactor engineering experiments has been started in 1958 using JRR-1 (a 50 kW water boiler type reactor with liquid fuel), which was the first research reactor in Japan. The role of the training has been transferred to JRR-4 (a 3500 kW swimming pool type reactor with ETR type fuel) since 1969 due to the decommission of JRR-1. The training courses which have been held are: JRR-1 Short-Term Course for Operation (1958 ∼ 1963) General Course (1961 ∼ ) Reactor Engineering Course (1976 ∼ ) Training Course in Nuclear Technology (International course)(1986 ∼ ). And individual training concerning research reactors for the participants of scientist exchange program sponsored by Science and Technology Agency and of bilateral agreement have been initiated in 1985. The graduates of these courses work as staff members in various fields in nuclear industry. (author)

  1. Quality assurance program plan for FRG sealed isotopic heat sources project (C-229)

    International Nuclear Information System (INIS)

    Tanke, J.M.

    1997-01-01

    This QAPP implements the Quality Assurance Program Plan for the FRG Sealed Isotopic Heat Sources Project (C-229). The heat source will be relocated from the 324 Building and placed in interim storage at the Central Waste Complex (CWC)

  2. Computerized reactor monitor and control for research reactors

    International Nuclear Information System (INIS)

    Buerger, L.; Vegh, E.

    1981-09-01

    The computerized process control system developed in the Central Research Institute for Physics, Budapest, Hungary, is described together with its special applications at research reactors. The nuclear power of the Hungarian research reactor is controlled by this computerized system, too, while in Lybia many interesting reactor-hpysical calculations are built into the computerized monitor system. (author)

  3. Nuclear reactor instrumentation at research reactor renewal

    International Nuclear Information System (INIS)

    Baers, B.; Pellionisz, P.

    1981-10-01

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

  4. Training and research on the nuclear reactor VR-1

    International Nuclear Information System (INIS)

    Matejka, K.

    1998-01-01

    The VR-1 training reactor is a light water reactor of the pool type using enriched uranium as the fuel. The moderator is demineralized light water, which also serves as the neutron reflector, biological shielding, and coolant. Heat evolved during the fission process is removed by natural convection. The reactor is used in the education of students in the field of reactor and neutron physics, dosimetry, nuclear safety, and instrumentation and control systems for nuclear facilities. Although primarily intended for students in various branches of technology (power engineering, nuclear engineering, physical engineering), this specialized facility is also used by students of faculties educating future natural scientists and teachers. Typical tasks trained at the VR-1 reactor include: measurement of delayed neutrons; examination of the effect of various materials on the reactivity of the reactor; measurement of the neutron flux density by various procedures; measurement of reactivity by various procedures; calibration of reactor control rods by various procedures; approaching the critical state; investigation of nuclear reactor dynamics; start-up, control and operation of a nuclear reactor; and investigation of the effect of a simulated nucleate boil on reactivity. In addition to the education of university-level students, training courses are also organized for specialists in the Czech nuclear programme

  5. Strategic planning for research reactors. Guidance for reactor managers

    International Nuclear Information System (INIS)

    2001-04-01

    The purpose of this publication is to provide guidance on how to develop a strategic plan for a research reactor. The IAEA is convinced of the need for research reactors to have strategic plans and is issuing a series of publications to help owners and operators in this regard. One of these covers the applications of research reactors. That report brings together all of the current uses of research reactors and enables a reactor owner or operator to evaluate which applications might be possible with a particular facility. An analysis of research reactor capabilities is an early phase in the strategic planning process. The current document provides the rationale for a strategic plan, outlines the methodology of developing such a plan and then gives a model that may be followed. While there are many purposes for research reactor strategic plans, this report emphasizes the use of strategic planning in order to increase utilization. A number of examples are given in order to clearly illustrate this function

  6. Expanding the storage capability at ET-RR-1 research reactor at Inshass

    International Nuclear Information System (INIS)

    Sultan, Mariy M.; Khattab, M.

    1999-01-01

    Storing of spent fuel from Test Reactor in developing countries has become a big dilemma for the following reasons: The transportation of spent fuel is very expensive; There are no reprocessing plants in most developing countries; The expanding of existing storage facilities in reactor building require experience that most of developing countries lack; Some political motivations from Nuclear Developed countries intervene which makes the transportation procedures and logistics to those countries difficult. This paper gives the conceptual design of a new spent fuel storage now under construction at Inshass research reactor (ET-RR-1). The location of the new storage facility is chosen to be within the premises of the reactor facility so that both reactor and the new storage are one Material Balance Area. The paper also proposes some ideas that can enhance the transportation and storage of spent fuel of test reactors, such as: Intensifying the role of IAEA in helping countries to get rid of the spent fuel; The initiation of regional spent fuel storage facilities in some developing countries. (author)

  7. Ageing of research reactors

    International Nuclear Information System (INIS)

    Ciocanescu, M.

    2001-01-01

    Historically, many of the research institutions were centred on a research reactor facility as main technological asset and major source of neutrons for research. Important achievements were made in time in these research institutions for development of nuclear materials technology and nuclear safety for nuclear energy. At present, ageing of nuclear research facilities among these research reactors and ageing of staff are considerable factors of reduction of competence in research centres. The safe way of mitigation of this trend deals with ageing management by so called, for power reactors, Plant Life Management and new investments in staff as investments in research, or in future resources of competence. A programmatic approach of ageing of research reactors in correlation with their actual and future utilisation, will be used as a basis for safety evaluation and future spending. (author)

  8. Safety of research reactors - A regulator's perspective

    International Nuclear Information System (INIS)

    Rahman, M.S.

    2001-01-01

    Due to historical reasons research reactors have received less regulatory attention in the world than nuclear power plants. This has given rise to several safety issues which, if not addressed immediately, may result in an undesirable situation. However, in Pakistan, research reactors and power reactors have received due attention from the regulatory authority. The Pakistan Research Reactor-1 has been under regulatory surveillance since 1965, the year of its commissioning. The second reactor has also undergone all the safety reviews and checks mandated by the licensing procedures. A brief description of the regulatory framework, the several safety reviews carried out have been briefly described in this paper. Significant activities of the regulatory authority have also been described in verifying the safety of research reactors in Pakistan along with the future activities. The views of the Pakistani regulatory authority on the specific issues identified by the IAEA have been presented along with specific recommendations to the IAEA. We are of the opinion that there are more Member States operating nuclear research reactors than nuclear power plants. Therefore, there should be more emphasis on the research reactor safety, which somehow has not been the case. In several recommendations made to the IAEA on the specific safety issues the emphasis has been, in general, to have a similar documentation and approach for maintaining and verifying operational safety at research reactors as is currently available for nuclear power reactors and may be planned for nuclear fuel cycle facilities. (author)

  9. Neutron irradiation embrittlement of reactor pressure vessel steel 20 MnMoNi55 weld

    International Nuclear Information System (INIS)

    Ghoneim, M.M.

    1987-05-01

    The effect of neutron irradiation on the mechanical and fracture properties of an 'improved' 20 MnMoNi 55 Pressure Vessel Steel (PVS) weld was investigated. In addition to very low residual element content, especially Cu (0.035 wt.%), and relatively higher Ni content (0.9 wt.%), this steel has higher strength (30% more) than the steels used currently in nuclear reactor pressure vessels. The material was irradiated to 3.5x10 19 and 7x10 19 n/cm 2 (E > 1 Mev) at 290 0 C and 2.5x10 19 n/cm 2 (E > 1 MeV) at 160 0 C in FRJ-1 and FRJ-2 research reactors at KFA, Juelich, F.R.G. Test methods used in the evaluation included instrumented impact testing of standard and precracked Charpy specimens, tensile, and fracture toughness testing. Instrumented impact testing provided load and energy vs. time (deflection) data in addition to energy absorption data. The results indicated that the investigated high strength improved steel is more resistant to irradiation induced embrittlement than conventional PVSs. (orig./IHOE)

  10. Reactor Safety Research: Semiannual report, January-June 1986: Reactor Safety Research Program

    International Nuclear Information System (INIS)

    1987-05-01

    Sandia National Laboratories is conducting, under USNRC sponsorship, phenomenological research related to the safety of commercial nuclear power reactors. The research includes experiments to simulate the phenomenology of accident conditions and the development of analytical models, verified by experiment, which can be used to predict reactor and safety systems performance behavior under abnormal conditions. The objective of this work is to provide NRC requisite data bases and analytical methods to (1) identify and define safety issues, (2) understand the progression of risk-significant accident sequences, and (3) conduct safety assessments. The collective NRC-sponsored effort at Sandia National Laboratories is directed at enhancing the technology base supporting licensing decisions

  11. Siting of research reactors

    International Nuclear Information System (INIS)

    1987-01-01

    The purpose of this document is to develop criteria for siting and the site-related design basis for research reactors. The concepts presented in this document are intended as recommendations for new reactors and are not suggested for backfitting purposes for facilities already in existence. In siting research reactors serious consideration is given to minimizing the effects of the site on the reactor and the reactor on the site and the potential impact of the reactor on the environment. In this document guidance is first provided on the evaluation of the radiological impact of the installation under normal reactor operation and accident conditions. A classification of research reactors in groups is then proposed, together with a different approach for each group, to take into account the relevant safety problems associated with facilities of different characteristics. Guidance is also provided for both extreme natural events and for man-induced external events which could affect the safe operation of the reactor. Extreme natural events include earthquakes, flooding for river or coastal sites and extreme meteorological phenomena. The feasibility of emergency planning is finally considered for each group of reactors

  12. Safety operation of training reactor VR-1

    International Nuclear Information System (INIS)

    Matejka, K.

    2001-01-01

    There are three nuclear research reactors in the Czech Republic in operation now: light water reactor LVR-15, maximum reactor power 10 MW t , owner and operator Nuclear Research Institute Rez; light water zero power reactor LR-0, maximum reactor power 5 kW t , owner and operator Nuclear Research Institute Rez and training reactor VR-1 Sparrow, maximum reactor power 5 kW t , owner and operate Faculty of Nuclear Sciences and Physical Engineering, CTU in Prague. The training reactor VR-1 Vrabec 'Sparrow', operated at the Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University in Prague, was started up on December 3, 1990. Particularly it is designed for training the students of Czech universities, preparing the experts for the Czech nuclear programme, as well as for certain research work, and for information programmes in the nuclear programme, as well as for certain research work, and for information programmes in sphere of using the nuclear energy (public relations). (author)

  13. Safety analysis for research reactors

    International Nuclear Information System (INIS)

    2008-01-01

    The aim of safety analysis for research reactors is to establish and confirm the design basis for items important to safety using appropriate analytical tools. The design, manufacture, construction and commissioning should be integrated with the safety analysis to ensure that the design intent has been incorporated into the as-built reactor. Safety analysis assesses the performance of the reactor against a broad range of operating conditions, postulated initiating events and other circumstances, in order to obtain a complete understanding of how the reactor is expected to perform in these situations. Safety analysis demonstrates that the reactor can be kept within the safety operating regimes established by the designer and approved by the regulatory body. This analysis can also be used as appropriate in the development of operating procedures, periodic testing and inspection programmes, proposals for modifications and experiments and emergency planning. The IAEA Safety Requirements publication on the Safety of Research Reactors states that the scope of safety analysis is required to include analysis of event sequences and evaluation of the consequences of the postulated initiating events and comparison of the results of the analysis with radiological acceptance criteria and design limits. This Safety Report elaborates on the requirements established in IAEA Safety Standards Series No. NS-R-4 on the Safety of Research Reactors, and the guidance given in IAEA Safety Series No. 35-G1, Safety Assessment of Research Reactors and Preparation of the Safety Analysis Report, providing detailed discussion and examples of related topics. Guidance is given in this report for carrying out safety analyses of research reactors, based on current international good practices. The report covers all the various steps required for a safety analysis; that is, selection of initiating events and acceptance criteria, rules and conventions, types of safety analysis, selection of

  14. Research reactors - an overview

    International Nuclear Information System (INIS)

    West, C.D.

    1997-01-01

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

  15. Welding of components of primary circuits of nuclear reactors in FRG

    International Nuclear Information System (INIS)

    Pehtts, P.; Iversen, K.

    1979-01-01

    Welding materials and methods, surfacing and soldering, applied when assembling nuclear reactors in the Federal Republic of Germany, are considered. It is noted that reactor vessel flux two-pass surfacing is mainly carried out, using the band electrode. The austenitic steel serves as filler material. Vessels are welded using electroslag flux method and nonconsumable electrodes. Tube plates claddina and tube welding during steam generator production are made by flux surfacing and inert gas shielded using nonconsumable electrode. When assembling fuel elements high temperature soldering with the solders, containing no boron of the Ni-Cr-Si and Ni-Cr-P systems is used

  16. . Effects of extended shutdown on the control rod drive mechanism of nigeria research reactor-1(NIRR-1)

    International Nuclear Information System (INIS)

    Yusuf, I; Mati, A. A.

    2010-01-01

    The control rod drive mechanism of the Nigeria Research Reactor-1 is being driven by a servo motor, type SDE-45 through a mechanical gear system. The servo motor ensures the position control of the control rod, and hence the stability of the neutron-flux of the nuclear research reactor. The control rod drive mechanism assembly is mounted on top of the reactor vessel, about 0.6m above 30m 3 volume of reactor pool water. The top of the pool is covered with a Perspex material to protect the water in the pool from environmental contamination and to reduce evaporation. Although most of the materials in the control rod drive mechanism assembly are made of stainless steel, the servo motor however contains corrodible materials. The paper reveals a practical experience of failure of the control rod drive mechanism as a result of corrosion growth between the rotor of the servo motor and its stator windings, due to an extended shutdown of the facility.

  17. Establishing a Radiation Protection Programme for a Research Reactor

    International Nuclear Information System (INIS)

    Abdallah, M. M.

    2014-04-01

    The nature and intensity of radiation from the operation of a research reactor depend on the type of reactor, its design features and its operational history. The protection of workers from the harmful effect of radiation must therefore be of paramount importance to any operating organization of a research reactor. This project report attempts to establish an operational radiation protection programme for a research reactor using the Ghana Research Reactor-1 as a case study. (au)

  18. Research reactor utilization in chemistry programmes

    International Nuclear Information System (INIS)

    Bautista, E.

    1983-01-01

    The establishment and roles of the Philippines Atomic Energy Commission in promoting and regulating the use of atomic energy are explained. The research reactor, PRR-1 is being converted to TRIGA to meet the increasing demands of high-flux. The activities of PAEC in chemistry research programs utilizing reactor are discussed in detail. The current and future plans of Research and Development programs are also included. (A.J.)

  19. Research reactors and materials testing

    International Nuclear Information System (INIS)

    Vidal, H.

    1986-01-01

    Research reactors can be classified in three main groups according to the moderator which is used. Their technical characteristics are given and the three most recent research and materials testing reactors are described: OSIRIS, ORPHEE and the high-flux reactor of Grenoble. The utilization of research reactors is reviewed in four fields of activity: training, fundamental or applied research and production (eg. radioisotopes) [fr

  20. Research reactor de-fueling and fuel shipment

    International Nuclear Information System (INIS)

    Ice, R.D.; Jawdeh, E.; Strydom, J.

    1998-01-01

    Planning for the Georgia Institute of Technology Research Reactor operations during the 1996 Summer Olympic Games began in early 1995. Before any details could be outlined, several preliminary administrative decisions had to be agreed upon by state, city, and university officials. The two major administrative decisions involving the reactor were (1) the security level and requirements and (2) the fuel status of the reactor. The Georgia Tech Research Reactor (GTRR) was a heavy-water moderated and cooled reactor, fueled with high-enriched uranium. The reactor was first licensed in 1964 with an engineered lifetime of thirty years. The reactor was intended for use in research applications and as a teaching facility for nuclear engineering students and reactor operators. Approximately one year prior to the olympics, the Georgia Tech administration decided that the GTRR fuel would be removed. In addition, a heightened, beyond regulatory requirements, security system was to be implemented. This report describes the scheduling, operations, and procedures

  1. Directory of Nuclear Research Reactors 1994

    International Nuclear Information System (INIS)

    1995-08-01

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

  2. Physical security at research reactors

    International Nuclear Information System (INIS)

    Clark, R.A.

    1977-01-01

    Of the 84 non-power research facilities licensed under 10 CFR Part 50, 73 are active (two test reactors, 68 research reactors and three critical facilities) and are required by 10 CFR Part 73.40 to provide physical protection against theft of SNM and against industrial sabotage. Each licensee has developed a security plan required by 10 CFR Part 50.34(c) to demonstrate the means of compliance with the applicable requirements of 10 CFR Part 73. In 1974, the Commission provided interim guidance for the organization and content of security plans for (a) test reactors, (b) medium power research and training reactors, and (c) low power research and training reactors. Eleven TRIGA reactors, with power levels greater than 250 kW and all other research and training reactors with power levels greater than 100 kW and less than or equal to 5,000 kW are designated as medium power research and training reactors. Thirteen TRIGA reactors with authorized power levels less than 250 kW are considered to be low power research and training reactors. Additional guidance for complying with the requirements of 73.50 and 73.60, if applicable, is provided in the Commission's Regulatory Guides. The Commission's Office of Inspection and Enforcement inspects each licensed facility to assure that an approved security plan is properly implemented with appropriate procedures and physical protection systems

  3. The rehabilitation/upgrading of Philippine Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Renato, T Banaga [Philippines Nuclear Research Inst., Quezon (Philippines)

    1998-10-01

    The Philippine Research Reactor (PRR-1) is the only research reactor in the Philippines. It was acquired through the Bilateral Agreement with the United States of America. The General Electric (G.E.) supplied PRR-1 first become operational in 1963 and used MTR plate type fuel. The original one-megawatt G.E. reactor was shutdown and converted into a 3 MW TRIGA PULSING REACTOR in 1984. The conversion includes the upgrading of the cooling system, replacement of new reactor coolant pumps, heat exchanger, cooling tower, replacement of new nuclear instrumentation and standard TRIGA console, TRIGA fuel supplied by General Atomic (G.A.). Philippine Nuclear Research Institute (PNRI) provided the old reactor, did the detailed design of the new cooling system, provided the new non-nuclear instrumentation and electrical power supply system and performed all construction, installation and modification work on site. The TRIGA conversion fuel is contained in a shrouded 4-rod cluster which fit into the original grid plate. The new fuel is a E{sub 1}-U-Z{sub 1}-H{sub 1.6} TRIGA fuel, has a 20% wt Uranium loading with 19.7% U-235 enrichment and about 0.5 wt % Erbium. The Start-up, calibration and Demonstration of Pulsing and Full Power Operation were completed during a three week start-up phase which were performed last March 1968. A few days after, a leak in the pool liner was discovered. The reactor was shutdown again for repair and up to present the reactor is still in the process of rehabilitation. This paper will describe the rehabilitation/upgrading done on the PRR-1 since 1988 up to present. (author)

  4. The rehabilitation/upgrading of Philippine Research Reactor

    International Nuclear Information System (INIS)

    Renato T, Banaga

    1998-01-01

    The Philippine Research Reactor (PRR-1) is the only research reactor in the Philippines. It was acquired through the Bilateral Agreement with the United States of America. The General Electric (G.E.) supplied PRR-1 first become operational in 1963 and used MTR plate type fuel. The original one-megawatt G.E. reactor was shutdown and converted into a 3 MW TRIGA PULSING REACTOR in 1984. The conversion includes the upgrading of the cooling system, replacement of new reactor coolant pumps, heat exchanger, cooling tower, replacement of new nuclear instrumentation and standard TRIGA console, TRIGA fuel supplied by General Atomic (G.A.). Philippine Nuclear Research Institute (PNRI) provided the old reactor, did the detailed design of the new cooling system, provided the new non-nuclear instrumentation and electrical power supply system and performed all construction, installation and modification work on site. The TRIGA conversion fuel is contained in a shrouded 4-rod cluster which fit into the original grid plate. The new fuel is a E 1 -U-Z 1 -H 1.6 TRIGA fuel, has a 20% wt Uranium loading with 19.7% U-235 enrichment and about 0.5 wt % Erbium. The Start-up, calibration and Demonstration of Pulsing and Full Power Operation were completed during a three week start-up phase which were performed last March 1968. A few days after, a leak in the pool liner was discovered. The reactor was shutdown again for repair and up to present the reactor is still in the process of rehabilitation. This paper will describe the rehabilitation/upgrading done on the PRR-1 since 1988 up to present. (author)

  5. Management of research reactor ageing

    International Nuclear Information System (INIS)

    1995-03-01

    As of December 1993, about one quarter of the operating research reactors were over 30 years old. The long life of research reactors has raised some concern amongst research reactor operators, regulators and, to some extent, the general public. The International Atomic Energy Agency commenced activities on the topic of research reactor ageing by appointing an internal working group in 1988 and convening a Consultants Meeting in 1989. The subject was also discussed at an international symposium and a regional seminar held in 1989 and 1992 respectively. A draft document incorporating information and experience exchanged at the above meetings was reviewed by a Technical Committee Meeting held in Vienna in 1992. The present TECDOC is the outcome of this meeting and contains recommendations, guidelines and information on the management of research reactor ageing, which should be used in conjunction with related publications of the IAEA Research Reactor Safety Programme, which are referenced throughout the text. This TECDOC will be of interest to operators and regulators involved with the safe operation of any type of research reactor to (a) understand the behaviour and influence of ageing mechanisms on the reactor structures, systems and components; (b) detect and assess the effect of ageing; (c) establish preventive and corrective measures to mitigate these effects; and (d) make decisions aimed at the safe and continued operation of a research reactor. 32 refs, tabs

  6. Management of research reactor ageing

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-03-01

    As of December 1993, about one quarter of the operating research reactors were over 30 years old. The long life of research reactors has raised some concern amongst research reactor operators, regulators and, to some extent, the general public. The International Atomic Energy Agency commenced activities on the topic of research reactor ageing by appointing an internal working group in 1988 and convening a Consultants Meeting in 1989. The subject was also discussed at an international symposium and a regional seminar held in 1989 and 1992 respectively. A draft document incorporating information and experience exchanged at the above meetings was reviewed by a Technical Committee Meeting held in Vienna in 1992. The present TECDOC is the outcome of this meeting and contains recommendations, guidelines and information on the management of research reactor ageing, which should be used in conjunction with related publications of the IAEA Research Reactor Safety Programme, which are referenced throughout the text. This TECDOC will be of interest to operators and regulators involved with the safe operation of any type of research reactor to (a) understand the behaviour and influence of ageing mechanisms on the reactor structures, systems and components; (b) detect and assess the effect of ageing; (c) establish preventive and corrective measures to mitigate these effects; and (d) make decisions aimed at the safe and continued operation of a research reactor. 32 refs, tabs.

  7. Research and materials irradiation reactors

    International Nuclear Information System (INIS)

    Ballagny, A.; Guigon, B.

    2004-01-01

    Devoted to the fundamental and applied research on materials irradiation, research reactors are nuclear installations where high neutrons flux are maintained. After a general presentation of the research reactors in the world and more specifically in France, this document presents the heavy water cooled reactors and the water cooled reactors. The third part explains the technical characteristics, thermal power, neutron flux, operating and details the Osiris, the RHF (high flux reactor), the Orphee and the Jules Horowitz reactors. The last part deals with the possible utilizations. (A.L.B.)

  8. Experimental study of the temperature distribution in the TRIGA IPR-R1 Brazilian research reactor

    International Nuclear Information System (INIS)

    Mesquita, Amir Zacarias

    2005-01-01

    The TRIGA-IPR-R1 Research Nuclear Reactor has completed 44 years in operation in November 2004. Its initial nominal thermal power was 30 kW. In 1979 its power was increased to 100 kW by adding new fuel elements to the reactor. Recently some more fuel elements were added to the core increasing the power to 250 kW. The TRIGA-IPR-R1 is a pool type reactor with a natural circulation core cooling system. Although the large number of experiments had been carried out with this reactor, mainly on neutron activation analysis, there is not many data on its thermal-hydraulics processes, whether experimental or theoretical. So a number of experiments were carried out with the measurement of the temperature inside the fuel element, in the reactor core and along the reactor pool. During these experiments the reactor was set in many different power levels. These experiments are part of the CDTN/CNEN research program, and have the main objective of commissioning the TRIGA-IPR-R1 reactor for routine operation at 250 kW. This work presents the experimental and theoretical analyses to determine the temperature distribution in the reactor. A methodology for the calibration and monitoring the reactor thermal power was also developed. This methodology allowed adding others power measuring channels to the reactor by using thermal processes. The fuel thermal conductivity and the heat transfer coefficient from the cladding to the coolant were also experimentally valued. lt was also presented a correlation for the gap conductance between the fuel and the cladding. The experimental results were compared with theoretical calculations and with data obtained from technical literature. A data acquisition and processing system and a software were developed to help the investigation. This system allows on line monitoring and registration of the main reactor operational parameters. The experiments have given better comprehension of the reactor thermal-fluid dynamics and helped to develop numerical

  9. The research reactors their contribution to the reactors physics

    International Nuclear Information System (INIS)

    Barral, J.C.; Zaetta, A.; Johner, J.; Mathoniere, G.

    2000-01-01

    The 19 october 2000, the french society of nuclear energy organized a day on the research reactors. This associated report of the technical session, reactors physics, is presented in two parts. The first part deals with the annual meeting and groups general papers on the pressurized water reactors, the fast neutrons reactors and the fusion reactors industry. The second part presents more technical papers about the research programs, critical models, irradiation reactors (OSIRIS and Jules Horowitz) and computing tools. (A.L.B.)

  10. Aspects of the Iea-R1 research reactor seismic evaluation

    International Nuclear Information System (INIS)

    Mattar Neto, Miguel

    1996-01-01

    Codes and standards for the seismic evaluation of the research reactor IEA-R1 are presented. An approach to define the design basis earthquake based on the local seismic map and on simplified analysis methods is proposed. The site seismic evaluation indicates that the design earthquake intensity is IV MM. Therefore, according to the used codes and standards, no buildings, systems, and components seismic analysis are required. (author)

  11. Ageing implementation and refurbishment development at the IEA-R1 nuclear research reactor: a 15 years experience

    International Nuclear Information System (INIS)

    Cardenas, Jose Patricio N.; Ricci Filho, Walter; Carvalho, Marcos R. de; Berretta, Jose Roberto; Marra Neto, Adolfo

    2011-01-01

    IPEN (Instituto de Pesquisas Energeticas e Nucleares) is a nuclear research center established into the Secretary of Science and Technology from the government of the state of Sao Paulo, and administered both technically and financially by Comissao Nacional de Energia Nuclear (CNEN), a federal government organization under the Ministry of Science and Technology. The institute is located inside the campus of the University of Sao Paulo, Sao Paulo city, Brazil. One of major nuclear facilities at IPEN is the IEA-R1 nuclear research reactor. It is the unique Brazilian research reactor with substantial power level suitable for application with research in physics, chemistry, biology and engineering, as well as radioisotope production for medical and other applications. Designed and built by Babcok-Wilcox, in accordance with technical specifications established by the Brazilian Nuclear Energy Commission, and financed by the US Atoms for Peace Program, it is a swimming pool type reactor, moderated and cooled by light water and uses graphite and beryllium as reflector elements. The first criticality was achieved on September 16, 1957 and the reactor is currently operating at 4.0 MW on a 64h per week cycle. Since 1996, an IEA-R1 reactor ageing study was established at the Research Reactor Center (CRPq) related with general deterioration of components belonging to some operational systems, as cooling towers from secondary cooling system, piping and pumps, sample irradiation devices, radiation monitoring system, fuel elements, rod drive mechanisms, nuclear and process instrumentation and safety operational system. Although basic structures are almost the same as the original design, several improvements and modifications in components, systems and structures had been made along reactor life. This work aims to show the development of the ageing program in the IEA-R1 reactor and the upgrading (modernization) that was carried out, concerning several equipment and system in the

  12. Ageing implementation and refurbishment development at the IEA-R1 nuclear research reactor: a 15 years experience

    Energy Technology Data Exchange (ETDEWEB)

    Cardenas, Jose Patricio N.; Ricci Filho, Walter; Carvalho, Marcos R. de; Berretta, Jose Roberto; Marra Neto, Adolfo, E-mail: ahiru@ipen.b, E-mail: wricci@ipen.b, E-mail: carvalho@ipen.b, E-mail: jrretta@ipen.b, E-mail: amneto@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    IPEN (Instituto de Pesquisas Energeticas e Nucleares) is a nuclear research center established into the Secretary of Science and Technology from the government of the state of Sao Paulo, and administered both technically and financially by Comissao Nacional de Energia Nuclear (CNEN), a federal government organization under the Ministry of Science and Technology. The institute is located inside the campus of the University of Sao Paulo, Sao Paulo city, Brazil. One of major nuclear facilities at IPEN is the IEA-R1 nuclear research reactor. It is the unique Brazilian research reactor with substantial power level suitable for application with research in physics, chemistry, biology and engineering, as well as radioisotope production for medical and other applications. Designed and built by Babcok-Wilcox, in accordance with technical specifications established by the Brazilian Nuclear Energy Commission, and financed by the US Atoms for Peace Program, it is a swimming pool type reactor, moderated and cooled by light water and uses graphite and beryllium as reflector elements. The first criticality was achieved on September 16, 1957 and the reactor is currently operating at 4.0 MW on a 64h per week cycle. Since 1996, an IEA-R1 reactor ageing study was established at the Research Reactor Center (CRPq) related with general deterioration of components belonging to some operational systems, as cooling towers from secondary cooling system, piping and pumps, sample irradiation devices, radiation monitoring system, fuel elements, rod drive mechanisms, nuclear and process instrumentation and safety operational system. Although basic structures are almost the same as the original design, several improvements and modifications in components, systems and structures had been made along reactor life. This work aims to show the development of the ageing program in the IEA-R1 reactor and the upgrading (modernization) that was carried out, concerning several equipment and system in the

  13. Guidelines for the Review of Research Reactor Safety: Revised Edition. Reference Document for IAEA Integrated Safety Assessment of Research Reactors (INSARR)

    International Nuclear Information System (INIS)

    2013-01-01

    The Integrated Safety Assessment of Research Reactors (INSARR) is an IAEA safety review service available to Member States with the objective of supporting them in ensuring and enhancing the safety of their research reactors. This service consists of performing a comprehensive peer review and an assessment of the safety of the respective research reactor. The reviews are based on IAEA safety standards and on the provisions of the Code of Conduct on the Safety of Research Reactors. The INSARR can benefit both the operating organizations and the regulatory bodies of the requesting Member States, and can include new research reactors under design or operating research reactors, including those which are under a Project and Supply Agreement with the IAEA. The first IAEA safety evaluation of a research reactor operated by a Member State was completed in October 1959 and involved the Swiss 20 MW DIORIT research reactor. Since then, and in accordance with its programme on research reactor safety, the IAEA has conducted safety review missions in its Member States to enhance the safety of their research reactor facilities through the application of the Code of Conduct on the Safety of Research Reactors and the relevant IAEA safety standards. About 320 missions in 51 Member States were undertaken between 1972 and 2012. The INSARR missions and other limited scope safety review missions are conducted following the guidelines presented in this publication, which is a revision of Guidelines for the Review of Research Reactor Safety (IAEA Services Series No. 1), published in December 1997. This publication details those IAEA safety standards and guidance publications relevant to the safety of research reactors that have been revised or published since 1997. The purpose of this publication is to give guidance on the preparation, implementation, reporting and follow-up of safety review missions. It is also intended to be of assistance to operators and regulators in conducting

  14. Improvement of research reactor sustainability

    International Nuclear Information System (INIS)

    Ciocanescu, M.; Paunoiu, C.; Toma, C.; Preda, M.; Ionila, M.

    2010-01-01

    The Research Reactors as is well known have numerous applications in a wide range of science technology, nuclear power development, medicine, to enumerate only the most important. The requirements of clients and stack-holders are fluctuating for the reasons out of control of Research Reactor Operating Organization, which may ensure with priority the safety of facility and nuclear installation. Sustainability of Research Reactor encompasses several aspects which finally are concentrated on safety of Research Reactor and economical aspects concerning operational expenses and income from external resources. Ensuring sustainability is a continuous, permanent activity and also it requests a strategic approach. The TRIGA - 14 MW Research Reactor detains a 30 years experience of safe utilization with good performance indicators. In the last 4 years the reactor benefited of a large investment project for modernization, thus ensuring the previous performances and opening new perspectives for power increase and for new applications. The previous core conversion from LEU to HEU fuel accomplished in 2006 ensures the utilization of reactor based on new qualified European supplier of TRIGA LEU fuel. Due to reduction of number of performed research reactors, the 14 MW TRIGA modernized reactor will play a significant role for the following two decades. (author)

  15. Development of Vibration Diagnostic System in Research Reactors

    International Nuclear Information System (INIS)

    EL-Kafas, A. A.

    1999-01-01

    Early failure detection and diagnosis system are an important group with increasing interest with the operating support system. Already existing system to monitor integrity of primary system components are vibration and acoustic monitoring system (2,3). The development of vibration diagnostic system for MARIA reactor (30 MW)-the second research reactor in Poland -was made. The new system is applied for the Egypt research reactor (ETRR-1). This paper represents the result obtained during the operation of this activity that carried out at MARIA and ETRR-1 reactors

  16. Research reactors; Les piles de recherche

    Energy Technology Data Exchange (ETDEWEB)

    Kowarski, L. [Commissariat a l' Energie Atomique, Paris (France). Centre d' Etudes Nucleaires]|[Organisation europeenne pour la Recherche Nucleaire, Geneve (Switzerland)

    1955-07-01

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

  17. Research reactors; Les piles de recherche

    Energy Technology Data Exchange (ETDEWEB)

    Kowarski, L [Commissariat a l' Energie Atomique, Paris (France). Centre d' Etudes Nucleaires; [Organisation europeenne pour la Recherche Nucleaire, Geneve (Switzerland)

    1955-07-01

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

  18. Calculation of the main neutron parameters of the IEA-R1 research reactor

    International Nuclear Information System (INIS)

    Ojima, Mario Katsuhiko

    1977-01-01

    The main neutron parameters of the research reactor IEA-R1 were calculated using computer programs to generate cross sections and criticality calculations. A calculation procedure based on the programs available in the Processing Center Data of IEA was established and centered in the HAMMER and CITATION system. A study was done in order to verify the validity and accuracy of the calculation method comparing the results with experimental data. Some operating parameters of the reactor, namely the distribution of neutron flux, the critical mass, the variation of the reactivity with the burning of fuel, and the dead time of the reactor were determined

  19. Industrial structure at research reactor suppliers

    International Nuclear Information System (INIS)

    Roegler, H.-J.; Bogusch, E.; Friebe, T.

    2001-01-01

    Due to the recent joining of the forces of Framatome S. A. from France and the Nuclear Division of Siemens AG Power Generation (KWU) from Germany to a Joint Venture named Framatome Advanced Nuclear Power S.A.S., the issue of the necessary and of the optimal industrial structure for nuclear projects as a research reactor is, was discussed internally often and intensively. That discussion took place also in the other technical fields such as Services for NPPs but also in the field of interest here, i. e. Research Reactors. In summarizing the statements of this presentation one can about state that: Research Reactors are easier to build than NPPs, but not standardised; Research Reactors need a wide spectrum of skills and experiences; to design and build Research Reactors needs an experienced team especially in terms of management and interfaces; Research Reactors need background from built reference plants more than from operating plants; Research Reactors need knowledge of suitable experienced subsuppliers. Two more essential conclusions as industry involved in constructing and upgrading research reactors are: Research Reactors by far are more than a suitable core that generates a high neutron flux; every institution that designs and builds a Research Reactor lacks quality or causes safety problems, damages the reputation of the entire community

  20. IAEA/CRP for decommissioning techniques for research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Won Zin; Won, H. J.; Kim, K. N.; Lee, K. W.; Jung, C. H

    2001-03-01

    The following were studied through the project entitled 'IAEA/CRP for decommissioning techniques for research reactors 1. Decontamination technology development for TRIGA radioactive soil waste - Electrokinetic soil decontamination experimental results and its mathematical simulation 2. The 2nd IAEA/CRP for decommissioning techniques for research reactors - Meeting results and program 3. Hosting the 2001 IAEA/RCA D and D training course for research reactors and small nuclear facilities.

  1. IAEA/CRP for decommissioning techniques for research reactors

    International Nuclear Information System (INIS)

    Oh, Won Zin; Won, H. J.; Kim, K. N.; Lee, K. W.; Jung, C. H.

    2001-03-01

    The following were studied through the project entitled 'IAEA/CRP for decommissioning techniques for research reactors 1. Decontamination technology development for TRIGA radioactive soil waste - Electrokinetic soil decontamination experimental results and its mathematical simulation 2. The 2nd IAEA/CRP for decommissioning techniques for research reactors - Meeting results and program 3. Hosting the 2001 IAEA/RCA D and D training course for research reactors and small nuclear facilities

  2. Research reactor`s role in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Choi, C-O [Korea Atomic Energy Research Inst., Taejon (Korea, Republic of)

    1996-12-31

    After a TRIGA MARK-II was constructed in 1962, new research activity of a general nature, utilizing neutrons, prevailed in Korea. Radioisotopes produced from the MARK-II played a good role in the 1960`s in educating people as to what could be achieved by a neutron source. Because the research reactor had implanted neutron science in the country, another TRIGA MARK-III had to be constructed within 10 years after importing the first reactor, due to increased neutron demand from the nuclear community. With the sudden growth of nuclear power, however, the emphasis of research changed. For a while research activities were almost all oriented to nuclear power plant technology. However, the specifics of nuclear power plant technology created a need for a more highly capable research reactor like HANARO 30MWt. HANARO will perform well with irradiation testing and other nuclear programs in the future, including: production of key radioisotopes, doping of silicon by transmutation, neutron activation analysis, neutron beam experiments, cold neutron source. 3 tabs., 2 figs.

  3. Generic Procedures for Response to a Nuclear or Radiological Emergency at Triga Research Reactors. Attachment 1 (2011)

    International Nuclear Information System (INIS)

    2011-01-01

    The publication provides guidance for response to emergencies at TRIGA research reactors in Threat Category II and III. It contains information on the unique behaviour of TRIGA fuel during accident conditions; it describes design characteristics of TRIGA research reactors and provides specific symptom-based emergency classification for this type of research reactor. This publication covers the determination of the appropriate emergency class and protective actions for a nuclear or radiological emergency at TRIGA research reactors. It does not cover nuclear security at TRIGA research reactors. The term 'threat category' is used in this publication as described in Ref. [6] and for the purposes of emergency preparedness and response only; this usage does not imply that any threat, in the sense of an intention and capability to cause harm, has been made in relation to facilities, activities or sources. The threat category is determined by an analysis of potential nuclear and radiological emergencies and the associated radiation hazard that could arise as a consequence of those emergencies. STRUCTURE. The attachment consists of an introduction which defines the background, objective, scope and structure, two sections covering technical aspects and appendices. Section 2 describes the characteristics of TRIGA fuel in normal and accident conditions. Section 3 contains TRIGA research reactor specific emergency classification tables for Threat Category II and III. These tables should be used instead of the corresponding emergency classification tables presented in Ref. [1] while developing the emergency response arrangements at TRIGA research reactors. The appendices present some historical overview and typical general data for TRIGA research reactor projects and the list of TRIGA installations around the world. The terms used in this document are defined in the IAEA Safety Glossary and the IAEA Code of Conduct on the Safety of Research Reactors.

  4. Australian research reactor studies

    International Nuclear Information System (INIS)

    McCulloch, D.B.

    1978-01-01

    The Australian AEC has two research reactors at the Lucas Heights Research Establishment, a 10 HW DIDO class materials testing reactor, HIFAR, and a smaller 100kW reactor MOATA, which was recently upgraded from 10kW power level. Because of the HIFAR being some 20 years old, major renewal and repair programmes are necessary to keep it operational. To enable meeting projected increases in demand for radioisotopes, plans for a new reactor to replace the HIFAR have been made and the design criteria are described in the paper. (author)

  5. Experience in utilizing research reactors in Yugoslavia

    Energy Technology Data Exchange (ETDEWEB)

    Pop-Jordanov, J.; Raisic, N. [Boris Kidric Institute of Nuclear Sciences VINCA, Belgrade (Yugoslavia); Copic, M.; Gabrovsek, Z. [Jozef Stefan Institute Ljubljana (Yugoslavia)

    1972-07-01

    The nuclear institutes in Yugoslavia possess three research reactors. Since 1958, two heavy-water reactors have been in operation at the 'Boris Kidric' Institute, a zero-power reactor RB and a 6. 5-MW reactor RA. At the Jozef Stefan Institute, a 250-kW TRIGA Mark II reactor has been operating since 1966. All reactors are equipped with the necessary experimental facilities. The main activities based on these reactors are: (1) fundamental research in solid-state and nuclear physics; (2) R and D activities related to nuclear power program; and (3) radioisotope production. In fundamental physics, inelastic neutron scattering and diffraction phenomena are studied by means of the neutron beam tubes and applied to investigations of the structures of solids and liquids. Valuable results are also obtained in n - γ reaction studies. Experiments connected with the fuel -element development program, owing to the characteristics of the existing reactors, are limited to determination of the fuel element parameters, to studies on the purity of uranium, and to a small number of capsule irradiations. All three reactors are also used for the verification of different methods applied in the analysis of power reactors, particularly concerning neutron flux distributions, the optimization of reactor core configurations and the shielding effects. An appreciable irradiation space in the reactors is reserved for isotope production. Fruitful international co-operation has been established in all these activities, on the basis of either bilateral or multilateral arrangements. The paper gives a critical analysis of the utilization of research reactors in a developing country such as Yugoslavia. The investments in and the operational costs of research reactors are compared with the benefits obtained in different areas of reactor application. The impact on the general scientific, technological and educational level in the country is also considered. In particular, an attempt is made ro

  6. Experience in utilizing research reactors in Yugoslavia

    International Nuclear Information System (INIS)

    Pop-Jordanov, J.; Raisic, N.; Copic, M.; Gabrovsek, Z.

    1972-01-01

    The nuclear institutes in Yugoslavia possess three research reactors. Since 1958, two heavy-water reactors have been in operation at the 'Boris Kidric' Institute, a zero-power reactor RB and a 6. 5-MW reactor RA. At the Jozef Stefan Institute, a 250-kW TRIGA Mark II reactor has been operating since 1966. All reactors are equipped with the necessary experimental facilities. The main activities based on these reactors are: (1) fundamental research in solid-state and nuclear physics; (2) R and D activities related to nuclear power program; and (3) radioisotope production. In fundamental physics, inelastic neutron scattering and diffraction phenomena are studied by means of the neutron beam tubes and applied to investigations of the structures of solids and liquids. Valuable results are also obtained in n - γ reaction studies. Experiments connected with the fuel -element development program, owing to the characteristics of the existing reactors, are limited to determination of the fuel element parameters, to studies on the purity of uranium, and to a small number of capsule irradiations. All three reactors are also used for the verification of different methods applied in the analysis of power reactors, particularly concerning neutron flux distributions, the optimization of reactor core configurations and the shielding effects. An appreciable irradiation space in the reactors is reserved for isotope production. Fruitful international co-operation has been established in all these activities, on the basis of either bilateral or multilateral arrangements. The paper gives a critical analysis of the utilization of research reactors in a developing country such as Yugoslavia. The investments in and the operational costs of research reactors are compared with the benefits obtained in different areas of reactor application. The impact on the general scientific, technological and educational level in the country is also considered. In particular, an attempt is made ro

  7. The pharmacokinetics and metabolism of lumiracoxib in chimeric humanized and murinized FRG mice.

    Science.gov (United States)

    Dickie, A P; Wilson, C E; Schreiter, K; Wehr, R; Wilson, E M; Bial, J; Scheer, N; Wilson, I D; Riley, R J

    2017-07-01

    The pharmacokinetics and metabolism of lumiracoxib were studied, after administration of single 10mg/kg oral doses to chimeric liver-humanized and murinized FRG mice. In the chimeric humanized mice, lumiracoxib reached peak observed concentrations in the blood of 1.10±0.08μg/mL at 0.25-0.5h post-dose with an AUC inf of 1.74±0.52μgh/mL and an effective half-life for the drug of 1.42±0.72h (n=3). In the case of the murinized animals peak observed concentrations in the blood were determined as 1.15±0.08μg/mL at 0.25h post-dose with an AUC inf of 1.94±0.22μgh/mL and an effective half-life of 1.28±0.02h (n=3). Analysis of blood indicated only the presence of unchanged lumiracoxib. Metabolic profiling of urine, bile and faecal extracts revealed a complex pattern of metabolites for both humanized and murinized animals with, in addition to unchanged parent drug, a variety of hydroxylated and conjugated metabolites detected. The profiles obtained in humanized mice were different compared to murinized animals with e.g., a higher proportion of the dose detected in the form of acyl glucuronide metabolites and much reduced amounts of taurine conjugates. Comparison of the metabolic profiles obtained from the present study with previously published data from C57bl/6J mice and humans, revealed a greater though not complete match between chimeric humanized mice and humans, such that the liver-humanized FRG model may represent a useful approach to assessing the biotransformation of such compounds in humans. Copyright © 2017 Elsevier Inc. All rights reserved.

  8. 2012 review of French research reactors

    International Nuclear Information System (INIS)

    Estrade, Jerome

    2013-01-01

    Proposed by the French Reactor Operators' Club (CER), the meeting and discussion forum for operators of French research reactors, this report first gives a brief presentation of these reactors and of their scope of application, and a summary of highlights in 2012 for each of them. Then, it proposes more detailed presentations and reviews of characteristics, activities, highlights, objectives and results for the different types of reactors: neutron beam reactors (Orphee, High flux reactor-Laue-Langevin Institute or HFR-ILL), technological irradiation reactors (Osiris and Phenix), training reactors (Isis and Azur), reactors for safety research purposes (Cabri and Phebus), reactors for neutronic studies (Caliban, Prospero, Eole, Minerve and Masurca), and new research reactors (the RES facility and the Jules Horowitz reactor or JHR)

  9. Supervisory system to monitor the neutron flux of the IPR-R1 TRIGA research reactor at CDTN

    International Nuclear Information System (INIS)

    Pinto, Antonio Juscelino; Mesquita, Amir Zacarias; Tello, Cledola Cassia Oliveira

    2009-01-01

    The IPR-R1 TRIGA Mark I nuclear research reactor at the Nuclear Technology Development Center - CDTN (Belo Horizonte) is a pool type reactor. It was designed for research, training and radioisotope production. The International Atomic Energy Agency- IAEA - recommends the use of friendly interfaces for monitoring and controlling the operational parameters of nuclear reactors. This paper reports the activities for implementing a supervisory system, using LabVIEW software, with the purpose to provide the IPR-R1 TRIGA research reactor with a modern, safe and reliable system to monitor the time evolution of the power of its core. The use of the LabVIEW will introduce modern techniques, based on electronic processor and visual interface in video monitor, substituting the mechanical strip chart recorders (ink-pen drive and paper) that monitor the current neutrons flux, which is proportional to the thermal power supplied by reactor core. The main objective of the system will be to follow the evolution of the neutronic flux originated in the Linear and Logarithmic channels. A great advantage of the supervisory software nowadays, in relation to computer programs currently used in the facility, is the existence of new resources such as the data transmission and graphical interfaces by net, grid lines display in the graphs, and resources for real time reactor core video recordings. The considered system could also in the future be optimized, not only for data acquisition, but also for the total control of IPR-R1 TRIGA reactor(author)

  10. Burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1

    Directory of Open Access Journals (Sweden)

    Muhammad Atta

    2011-01-01

    Full Text Available The burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1, reference operating core, has been carried out utilizing standard computer codes WIMS/D4, CITATION, and RELAP5/MOD3.4. Reactor codes WIMS/D4 and CITATION have been used for the calculations of neutronic parameters including peaking factors and power profiles at different burn-up considering a xenon free core and also the equilibrium xenon values. RELAP5/MOD3.4 code was utilized for the determination of peak fuel centerline, clad and coolant temperatures to ensure the safety of the reactor throughout the cycle. The calculations reveal that the reactor is safe and no nucleate boiling will commence at any part of the core throughout the cycle and that the safety margin increases with burnup as peaking factors decrease.

  11. Demolition of the FRJ-1 research reactor (MERLIN); Abbau des Reaktorblocks des Forschungsreaktors FRJ-1 (MERLIN)

    Energy Technology Data Exchange (ETDEWEB)

    Stahn, B.; Matela, K.; Zehbe, C. [Forschungszentrum Juelich GmbH (Germany); Poeppinghaus, J. [Gesellschaft fuer Nuklearservice, Essen (Germany); Cremer, J. [SNT Siempelkamp Nukleartechnik, Heidelberg (Germany)

    2003-06-01

    FRJ-2 (MERLIN), the swimming pool reactor cooled and moderated by light water, was built at the then Juelich Nuclear Research Establishment (KFA) between 1958 and 1962. In the period between 1964 and 1985, it was used for. The reactor was decommissioned in 1985. Since 1996, most of the demolition work has been carried out under the leadership of a project team. The complete secondary cooling system was removed by late 1998. After the cooling loops and experimental installations had been taken out, the reactor vessel internals were removed in 2000 after the water had been drained from the reactor vessel. After the competent authority had granted a license, demolition of the reactor block, the central part of the research reactor, was begun in October 2001. In a first step, the reactor operating floor and the reactor attachment structures were removed by the GNS/SNT consortium charged with overall planning and execution of the job. This phase gave rise to approx. The reactor block proper is dismantled in a number of steps. A variety of proven cutting techniques are used for this purpose. Demolition of the reactor block is to be completed in the first half of 2003. (orig.) [German] Der mit Leichtwasser gekuehlte und moderierte Schwimmbad-Forschungsreaktor FRJ-2 (MERLIN) wurde von 1958 bis 1962 fuer die damalige Kernforschungsanlage Juelich (KFA) errichtet. Von 1964 bis 1985 wurde er fuer Experimente mit zunaechst 5 MW und spaeter 10 MW thermischer Leistung bei einem maximalen thermischen Neutronenfluss von 1,1.10{sup 14} n/cm{sup 2}s genutzt. Im Jahr 1985 stellte der Reaktor seinen Betrieb ein. Die Brennelemente wurden aus der Anlage entfernt und in die USA und nach Grossbritannien verbracht. Seit 1996 erfolgen die wesentlichen Abbautaetigkeiten unter Leitung eines verantwortlichen Projektteams. Bis Ende 1998 wurde das komplette Sekundaerkuehlsystem entfernt. Dem Abbau der Kuehlkreislaeufe und Experimentiereinrichtungen folgte im Jahr 2000 der Ausbau der

  12. Monte Carlo simulation of core physics parameters of the Nigeria Research Reactor-1 (NIRR-1)

    Energy Technology Data Exchange (ETDEWEB)

    Jonah, S.A. [Reactor Engineering Section, Centre for Energy Research and Training, Ahmadu Bello University, Zaria, P.M.B. 1014 (Nigeria)], E-mail: jonahsa2001@yahoo.com; Liaw, J.R.; Matos, J.E. [RERTR Program, Nuclear Engineering Division, Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439 (United States)

    2007-12-15

    The Monte Carlo N-Particle (MCNP) code, version 4C (MCNP4C) and a set of neutron cross-section data were used to develop an accurate three-dimensional computational model of the Nigeria Research Reactor-1 (NIRR-1). The geometry of the reactor core was modeled as closely as possible including the details of all the fuel elements, reactivity regulators, the control rod, all irradiation channels, and Be reflectors. The following reactor core physics parameters were calculated for the present highly enriched uranium (HEU) core: clean cold core excess reactivity ({rho}{sub ex}), control rod (CR) and shim worth, shut down margin (SDM), neutron flux distributions in the irradiation channels, reactivity feedback coefficients and the kinetics parameters. The HEU input model was validated by experimental data from the final safety analyses report (SAR). The model predicted various key neutronics parameters fairly accurately and the calculated thermal neutron fluxes in the irradiation channels agree with the values obtained by foil activation method. Results indicate that the established Monte Carlo model is an accurate representation of the NIRR-1 HEU core and will be used to perform feasibility for conversion to low enriched uranium (LEU)

  13. The CEA research reactors

    International Nuclear Information System (INIS)

    Schwartz, J.P.

    1993-01-01

    Two main research reactors, specifically designed, PEGASE reactor and Laue-Langevin high flux reactor, are presented. The PEGASE reactor was designed at the end of the 50s for the study of the gas cooled reactor fuel element behaviour under irradiation; the HFR reactor, was designed in the late 60s to serve as a high yield and high level neutron source. Historical backgrounds, core and fuel characteristics and design, flux characteristics, etc., are presented. 5 figs

  14. Replacement research reactor for Australia

    International Nuclear Information System (INIS)

    Miller, Ross

    1998-01-01

    In 1992, the Australian Government commissioned a review into the need for a replacement research reactor. That review concluded that in about years, if certain conditions were met, the Government could make a decision in favour of a replacement reactor. A major milestone was achieved when, on 3 September 1997, the Australian Government announced the construction of a replacement research reactor at the site of Australia's existing research reactor HIFAR, subject to the satisfactory outcome of an environmental assessment process. The reactor will be have the dual purpose of providing a first class facility for neutron beam research as well as providing irradiation facilities for both medical isotope production and commercial irradiations. The project is scheduled for completion before the end of 2005. (author)

  15. Status of prompt gamma neutron activation analysis (PGAA) at TRR-1/M1 (Thai Research Reactor-1/Modified 1)

    Energy Technology Data Exchange (ETDEWEB)

    Asvavijnijkulchai, Chanchai; Dharmavanij, Wanchai; Siangsanan, Pariwat; Ratanathongchai, Wichian; Chongkum, Somporn [Physics Division, Office of Atomic Energy for Peace, Vibhavadi Rangsit Road, Chatuchak, Bangkok (Thailand)

    1999-08-01

    The first prompt gamma activation analysis (PGAA) was designed, constructed and installed at a 6 inch diameter neutron beam port of the Thai Research Reactor-1/Modified 1 (TRR-1/M1) since 1989. Beam characteristic were made by Gd foil irradiation, X-ray film exposing and densitometry scanning consequently. The thermal neutron flux at sample position was measured by Au foil activation, and was about 1 x 10{sup 7} n.cm{sup 2}.sec{sup -1} at 700 kW operating power. The experiments have been conducted successfully. In 1998, the PGAA facility has been developed for the reactor operating power at 1.2 MW. The new PGAA system, e.g., beam shutter, gamma collimator and biological shields have been designed to reduce the leakage of neutrons and gamma radiation to the acceptance levels in accordance with the International Commission on Radiation Protection Publication 60 (ICRP 60). The construction and installation will be completed in April 1999. (author)

  16. Status of prompt gamma neutron activation analysis (PGAA) at TRR-1/M1 (Thai Research Reactor-1/Modified 1)

    International Nuclear Information System (INIS)

    Asvavijnijkulchai, Chanchai; Dharmavanij, Wanchai; Siangsanan, Pariwat; Ratanathongchai, Wichian; Chongkum, Somporn

    1999-01-01

    The first prompt gamma activation analysis (PGAA) was designed, constructed and installed at a 6 inch diameter neutron beam port of the Thai Research Reactor-1/Modified 1 (TRR-1/M1) since 1989. Beam characteristic were made by Gd foil irradiation, X-ray film exposing and densitometry scanning consequently. The thermal neutron flux at sample position was measured by Au foil activation, and was about 1 x 10 7 n.cm 2 .sec -1 at 700 kW operating power. The experiments have been conducted successfully. In 1998, the PGAA facility has been developed for the reactor operating power at 1.2 MW. The new PGAA system, e.g., beam shutter, gamma collimator and biological shields have been designed to reduce the leakage of neutrons and gamma radiation to the acceptance levels in accordance with the International Commission on Radiation Protection Publication 60 (ICRP 60). The construction and installation will be completed in April 1999. (author)

  17. Preparation fo nuclear research reactors operators

    International Nuclear Information System (INIS)

    Roedel, G.

    1986-01-01

    The experience obtained with the training of operators of nuclear research reactors is presented. The main tool used in the experiments is the IPR-R1 reactor, a TRIGA MARK I type, owned by Nuclear Technology Development Centre (CDTN) of NUCLEBRAS. The structures of the Research Reactors Operators Training Course and of the Radiological Protection Course, as well as the Operators Qualifying and Requalifying Program, all of them prepared at CDTN are also presented. Mention is made of the application of similar experiments to other groups, such as students coming from Nuclear Sciences and Techniques Course of the Federal University of Minas Gerais. (Author) [pt

  18. Preparation of nuclear research reactors operators

    International Nuclear Information System (INIS)

    Roedel, G.

    1986-01-01

    The experience obtained with the training of operators of nuclear research reactors is presented. The main tool used in the experiments is the IPR-R1 reactor, a TRIGA MARK I type, owned by Nuclear Technology Development Centre (CDTN) of NUCLEBRAS. The structures of the Research Reactors Operators Training Course and of the Radiological Protection Course, as well as the Operators Qualifying and Requalifying Program, all of them prepared at CDTN, are also presented. Mention is made of the application of similar experiments to other groups, such as students coming from Nuclear Sciences and Techniques Course of the Federal University of Minas Gerais. (Author) [pt

  19. Decommissioning of the Neuherberg Research Reactor (FRN)

    International Nuclear Information System (INIS)

    Demmeler, M.; Rau, G.; Strube, D.

    1982-01-01

    The Neuherberg Research Reactor is of type TRIGA MARK III with 1 MW steady state power and pulsable up to 2000 MW. During more than ten years of operation 12000 MWh and 6000 reactor pulses had been performed. In spite of its good technical condition and of permanent safe operation without any failures, the decommissioning of the Neuherberg research reactor was decided by the GSF board of directors to save costs for maintaining and personnel. As the mode of decommissioning the safe enclosure was chosen which means that the fuel elements will be transferred back to the USA. All other radioactive reactor components will be enclosed in the reactor block. Procedures for licensing of the decommissioning, dismantling procedures and time tables are presented

  20. Research reactors: design, safety requirements and applications

    International Nuclear Information System (INIS)

    Hassan, Abobaker Mohammed Rahmtalla

    2014-09-01

    There are two types of reactors: research reactors or power reactors. The difference between the research reactor and energy reactor is that the research reactor has working temperature and fuel less than the power reactor. The research reactors cooling uses light or heavy water and also research reactors need reflector of graphite or beryllium to reduce the loss of neutrons from the reactor core. Research reactors are used for research training as well as testing of materials and the production of radioisotopes for medical uses and for industrial application. The difference is also that the research reactor smaller in terms of capacity than that of power plant. Research reactors produce radioactive isotopes are not used for energy production, the power plant generates electrical energy. In the world there are more than 284 reactor research in 56 countries, operates as source of neutron for scientific research. Among the incidents related to nuclear reactors leak radiation partial reactor which took place in three mile island nuclear near pennsylvania in 1979, due to result of the loss of control of the fission reaction, which led to the explosion emitting hug amounts of radiation. However, there was control of radiation inside the building, and so no occurred then, another accident that lead to radiation leakage similar in nuclear power plant Chernobyl in Russia in 1986, has led to deaths of 4000 people and exposing hundreds of thousands to radiation, and can continue to be effect of harmful radiation to affect future generations. (author)

  1. New research possibilities at the Budapest research reactor

    International Nuclear Information System (INIS)

    Hargitai, T.; Vidovszky, I.

    2001-01-01

    The Budapest Research Reactor is the first nuclear facility of Hungary. It was commissioned in 1959, reconstructed and upgraded in 1967 and 1986-92. The main purpose of the reactor is to serve neutron research. The reactor was extended by a liquid hydrogen type cold neutron source in 2000. The research possibilities are much improved by the CNS both in neutron scattering and neutron activation. (author)

  2. ISO-9001: An approach to accreditation for an MTR facility: SAFARI-1 research reactor

    International Nuclear Information System (INIS)

    Piani, C.S.B.; Du Bruyn, J.F.B.

    2000-01-01

    The SAFARI-1 Research Reactor obtained ISO-9001 accreditation via the South African Bureau of Standards in September 1998. In view of the commercial applications of the reactor, the value of acquisition of the accreditation was considered against the cost of implementation of the Quality System. The criteria identified in the ISO-9001 standard were appraised and a superstructure derived for management of the generation and implementation of a suitable Quality Management System (QMS) for the fairly unique application of a nuclear research reactor. A Quality Policy was established, which formed the basis of the QMS against which the various requirements and/or standards were identified. In addition, since it was considered advantageous to incorporate the management controls of Conventional and Radiological Safety as well as Plant Maintenance and Environmental Management (ISO 14001), these aspects were included in the QMS. (author)

  3. Jordan Research and Training Reactor (JRTR) Utilization Facilities

    International Nuclear Information System (INIS)

    Xoubi, N.

    2013-01-01

    Jordan Research and Training Reactor (JRTR) is a 5 MW light water open pool multipurpose reactor that serves as the focal point for Jordan National Nuclear Centre, and is designed to be utilized in three main areas: Education and training, nuclear research, and radioisotopes production and other commercial and industrial services. The reactor core is composed of 18 fuel assemblies, MTR plate type 19.75% enriched uranium silicide (U 3 Si 2 ) in aluminium matrix, and is reflected on all sides by beryllium and graphite. The reactor power is upgradable to 10 MW with a maximum thermal flux of 1.45×10 14 cm -2 s -1 , and is controlled by a Hafnium control absorber rod and B 4 C shutdown rod. The reactor is designed to include laboratories and classrooms that will support the establishment of a nuclear reactor school for educating and training students in disciplines like nuclear engineering, reactor physics, radiochemistry, nuclear technology, radiation protection, and other related scientific fields where classroom instruction and laboratory experiments will be related in a very practical and realistic manner to the actual operation of the reactor. JRTR is designed to support advanced nuclear research as well as commercial and industrial services, which can be preformed utilizing any of its 35 experimental facilities. (author)

  4. Research reactors spent fuel management in the Nuclear Research Institute Rez

    International Nuclear Information System (INIS)

    Rychecky, J.

    2001-01-01

    In Czech Republic 3 research and testing nuclear reactors are operated at present time, with the biggest one being the Nuclear Research Institute (NRI) reactor LVR-15, operated with maximum power 10 MW. This reactor serves as a radiation source for material testing, producing of ionizing radiation sources, theoretical studies, and, most recently, for boron neutron capture therapy. Another NRI reactor LR-0 is a reactor of zero power used mainly for the studies of WWER 1000 spent fuel criticality. For training of students the reactor called VRABEC (VR-1), operated also with very low power, serves since 1990 at the Faculty of Nuclear Engineering, of Czech Technical University. The similar testing type reactor (SR-0), already decommissioned, was also used since 1974 to 1989 in Skoda, Nuclear Machinery, Plzen. This contribution summarizes the present state of the spent fuel (SF) management of these nuclear reactors. As the SF management is different for very low or zero power reactors and power reactors, the first type will be only briefly discussed, and then the main attention will be devoted to SF management of the NRI experimental reactor LVR-15

  5. Research reactors spent fuel management in the Nuclear Research Institute Rez

    Energy Technology Data Exchange (ETDEWEB)

    Rychecky, J. [Nuclear Research Institute, 25068 Rez (Czech Republic)

    2001-07-01

    In Czech Republic 3 research and testing nuclear reactors are operated at present time, with the biggest one being the Nuclear Research Institute (NRI) reactor LVR-15, operated with maximum power 10 MW. This reactor serves as a radiation source for material testing, producing of ionizing radiation sources, theoretical studies, and, most recently, for boron neutron capture therapy. Another NRI reactor LR-0 is a reactor of zero power used mainly for the studies of WWER 1000 spent fuel criticality. For training of students the reactor called VRABEC (VR-1), operated also with very low power, serves since 1990 at the Faculty of Nuclear Engineering, of Czech Technical University. The similar testing type reactor (SR-0), already decommissioned, was also used since 1974 to 1989 in Skoda, Nuclear Machinery, Plzen. This contribution summarizes the present state of the spent fuel (SF) management of these nuclear reactors. As the SF management is different for very low or zero power reactors and power reactors, the first type will be only briefly discussed, and then the main attention will be devoted to SF management of the NRI experimental reactor LVR-15.

  6. Reactor core conversion studies of Ghana: Research Reactor-1 and proposal for addition of safety rod

    International Nuclear Information System (INIS)

    Odoi, H.C.

    2014-06-01

    The inclusion of an additional safety rod in conjunction with a core conversion study of Ghana Research Reactor-1 (GHARR-1) was carried out using neutronics, thermal hydraulics and burnup codes. The study is based on a recommendation by Integrated Safety Assessment for Research Reactors (INSARP) mission to incorporate a safety rod to the reactor safety system as well as the need to replace the reactor fuel with LEU. Conversion from one fuel type to another requires a complete re-evaluation of the safety analysis. Changes to the reactivity worth, shutdown margin, power density and material properties must be taken into account, and appropriate modifications made. Neutronics analysis including burnup was studied followed by thermal hydraulics analyses which comprise steady state and transients. Four computer codes were used for the analysis; MCNP, REBUS, PLTEP and PARET. The neutronics analysis revealed that the LEU core must be operated at 34 Kw in order to attain the flux of 1.0E12 n/cm 2 .s as the nominal flux of the HEU core. The auxiliary safety rod placed at a modified irradiation site gives a better worth than the cadmium capsules. For core excess reactivity of 4 mk, 348 fuel pins would be appropriate for the GHARR-1 LEU core. Results indicate that flux level of 1.0E12 n/cm 2 .s in the inner irradiation channel will not be compromised, if the power of the LEU core is increased to 34 kW. The GHARR-1 core using LEU-U0 2 -12.5% fuel can be operated for 23 shim cycles, with cycles length 2.5 years, for over 57 years at the 17 kW power level. All 23 LEU cycles meet the ∼ 4.0 mk excess reactivity required at the beginning of cycle . For comparison, the MNSR HEU reference core can also be operated for 23 shim cycles, but with a cycle length of 2.0 years for just over 46 years at 15.0kW power level. It is observed that the GHARR-1 core with LEU UO 2 fuel enriched to 12.5% and a power level of 34 kW can be operated ∼25% longer than the current HEU core operated at

  7. Seismic research on graphite reactor core

    International Nuclear Information System (INIS)

    Lai Shigang; Sun Libin; Zhang Zhengming

    2013-01-01

    Background: Reactors with graphite core structure include production reactor, water-cooled graphite reactor, gas-cooled reactor, high-temperature gas-cooled reactor and so on. Multi-body graphite core structure has nonlinear response under seismic excitation, which is different from the response of general civil structure, metal connection structure or bolted structure. Purpose: In order to provide references for the designing and construction of HTR-PM. This paper reviews the history of reactor seismic research evaluation from certain countries, and summarizes the research methods and research results. Methods: By comparing the methods adopted in different gas-cooled reactor cores, inspiration for our own HTR seismic research was achieved. Results and Conclusions: In this paper, the research ideas of graphite core seismic during the process of designing, constructing and operating HTR-10 are expounded. Also the project progress of HTR-PM and the research on side reflection with the theory of similarity is introduced. (authors)

  8. Utilization of nuclear research reactors

    International Nuclear Information System (INIS)

    1980-01-01

    Full text: Report on an IAEA interregional training course, Budapest, Hungary, 5-30 November 1979. The course was attended by 19 participants from 16 Member States. Among the 28 training courses which the International Atomic Energy Agency organized within its 1979 programme of technical assistance was the Interregional Training Course on the Utilization of Nuclear Research Reactors. This course was held at the Nuclear Training Reactor (a low-power pool-type reactor) of the Technical University, Budapest, Hungary, from 5 to 30 November 1979 and it was complemented by a one-week Study Tour to the Nuclear Research Centre in Rossendorf near Dresden, German Democratic Republic. The training course was very successful, with 19 participants attending from 16 Member States - Bangladesh, Bolivia, Czechoslovakia, Ecuador, Egypt, India, Iraq, Korean Democratic People's Republic, Morocco, Peru, Philippines, Spain, Thailand, Turkey, Vietnam and Yugoslavia. Selected invited lecturers were recruited from the USA and Finland, as well as local scientists from Hungarian institutions. During the past two decades or so, many research reactors have been put into operation around the world, and the demand for well qualified personnel to run and fully utilize these facilities has increased accordingly. Several developing countries have already acquired small- and medium-size research reactors mainly for isotope production, research in various fields, and training, while others are presently at different stages of planning and installation. Through different sources of information, such as requests to the IAEA for fellowship awards and experts, it became apparent that many research reactors and their associated facilities are not being utilized to their full potential in many of the developing countries. One reason for this is the lack of a sufficient number of trained professionals who are well acquainted with all the capabilities that a research reactor can offer, both in research and

  9. Extensive utilization of training reactor VR-1

    International Nuclear Information System (INIS)

    Matejka, Karel; Sklenka, Lubomir

    2003-01-01

    Full text: The training reactor VR-1 Vrabec ('Sparrow'), operated at the Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University in Prague, was started up on December 3, 1990. Particularly, it is designed and operated for training of students from Czech universities, preparing of experts for the Czech nuclear programme, as well as for certain research and development work, and for information programmes in the sphere of non-military nuclear energy use (public relation). The VR-1 training reactor is a pool-type light-water reactor based on enriched uranium with maximum thermal power 1kWth and short time period up to 5kW th . The moderator of neutrons is light demineralized water (H 2 O) that is also used as a reflector, a biological shielding, and a coolant. Heat is removed from the core with natural convection. The reactor core contains 14 to 18 fuel assemblies IRT-3M, depending on the geometric arrangement and kind of experiments to be performed in the reactor. The core is accommodated in a cylindrical stainless steel vessel - pool, which is filled with water. UR-70 control rods serve the reactor control and safe shutdown. Training of the VR-1 reactor provides students with experience in reactor and neutron physics, dosimetry, nuclear safety, and nuclear installation operation. Students from technical universities and from natural sciences universities come to the reactor for training. Approximately 200 university students are introduced to the reactor (lectures, experiments, experimental and diploma works, etc.) every year. About 12 different faculties from Czech universities use the reactor. International co-operation with European universities in Germany, Hungary, Austria, Slovakia, Holland and UK is frequent. Practical Course on Reactor Physics in Framework of European Nuclear Engineering Network has been newly introduced. Currently, students can try out more than 20 experimental exercises. Further training courses have been included

  10. Current tendencies and perspectives of development research reactors of Russia

    International Nuclear Information System (INIS)

    Gabaraev, B.A.; Kchmelschikov, V.V.

    2004-01-01

    Full text: During more than fifty years many Research Reactors were constructed under Russian projects, and that is a considerable contribution to the world reactor building. The designs of Research Reactors, constructed under Russian projects, appeared to be so successful, that permitted to raise capacity and widen the range of their application. The majority of Russian Research Reactors being middle-aged are far from having their designed resources exhausted and are kept on the intensive run still. In 2000 'Strategy of nuclear power development in Russia in the first half of XXI century' was elaborated and approved. The national nuclear power requirements and possible ways of its development determined in this document demanded to analyze the state of the research reactors base. The analysis results are presented in this report. The main conclusion consists in the following statement: on the one hand quantity and experimental potentialities of domestic Research Reactors are sufficient for the solution of reactor materials science tasks, and on the other hand the reconstruction and modernization appears to be the most preferable way of research reactors development for the near-term outlook. At present time the modernization and reconstruction works and works on extension of operational life of high-powered multipurpose MIR-M1, SM-3, IRV-1M, BOR-60, IVV-2M and others are conducted. There is support for the development of Research Reactors, intended for carrying out the fundamental investigations on the neutron beams. Toward this end the Government of Russia gives financial and professional support with a view to complete the reactor PIK construction in PINPh and the reactor IBR-2 modernization in JINR. In future prospect Research Reactors branch in Russia is to acquire the following trends: - limited number of existent scientific centers, based on the construction sites, with high flux materials testing research reactors, equipped with experimental facilities

  11. RAZORBACK - A Research Reactor Transient Analysis Code Version 1.0 - Volume 3: Verification and Validation Report.

    Energy Technology Data Exchange (ETDEWEB)

    Talley, Darren G.

    2017-04-01

    This report describes the work and results of the verification and validation (V&V) of the version 1.0 release of the Razorback code. Razorback is a computer code designed to simulate the operation of a research reactor (such as the Annular Core Research Reactor (ACRR)) by a coupled numerical solution of the point reactor kinetics equations, the energy conservation equation for fuel element heat transfer, the equation of motion for fuel element thermal expansion, and the mass, momentum, and energy conservation equations for the water cooling of the fuel elements. This V&V effort was intended to confirm that the code shows good agreement between simulation and actual ACRR operations.

  12. Nuclear Capacity Building through Research Reactors

    International Nuclear Information System (INIS)

    2017-01-01

    Four Instruments: •The IAEA has recently developed a specific scheme of services for Nuclear Capacity Building in support of the Member States cooperating research reactors (RR) willing to use RRs as a primary facility to develop nuclear competences as a supporting step to embark into a national nuclear programme. •The scheme is composed of four complementary instruments, each of them being targeted to specific objective and audience: Distance Training: Internet Reactor Laboratory (IRL); Basic Training: Regional Research Reactor Schools; Intermediate Training: East European Research Reactor Initiative (EERRI); Group Fellowship Course Advanced Training: International Centres based on Research Reactors (ICERR)

  13. IDAS-RR: an incident data base system for research reactors

    International Nuclear Information System (INIS)

    Matsumoto, Kiyoshi; Kohsaka, Atsuo; Kaminaga, Masanori; Murayama, Youji; Ohnishi, Nobuaki; Maniwa, Masaki.

    1990-03-01

    An Incident Data Base System for Research Reactors, IDAS-RR, has been developed. IDAS-RR has information about abnormal incidents (failures, transients, accidents, etc.) of research reactors in the world. Data reference, input, editing and other functions of IDAS-RR are menu driven. The routine processing and data base management functions are performed by the system software and hardware. PC-9801 equipment was selected as the hardware because of its portability and popularity. IDAS-RR provides effective reference information for the following activities. 1) Analysis of abnormal incident of research reactors, 2) Detail analysis of research reactor behavior in the abnormal incident for building the knowledge base of the reactor emergency diagnostic system for research reactor, 3) Planning counter-measure for emergency situation in the research reactor. This report is a user's manual of IDAS-RR. (author)

  14. Reversal of OFI and CHF in Research Reactors Operating at 1 to 50 Bar. Version 1.0

    Energy Technology Data Exchange (ETDEWEB)

    Kalimullah, M. [Argonne National Lab. (ANL), Argonne, IL (United States); Olson, A. P. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Feldman, E. E. [Argonne National Lab. (ANL), Argonne, IL (United States); Matos, J. E. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2014-02-28

    The conditions at which the critical heat flux (CHF) and the heat flux at the onset of Ledinegg flow instability (OFI) are equal, are determined for a coolant channel with uniform heat flux as a function of five independent parameters: the channel exit pressure (P), heated length (Lh) , heated diameter (Dh), inlet temperature (Tin), and mass flux (G). A diagram is made by plotting the mass flux and heat flux at the OFI-CHF intersection (reversal from CHF > OFI to CHF < OFI as G increases) as a function of P (1 to 50 bar), for 36 combinations of the remaining three parameters (Lh , Dh , Tin): Lh = 0.28, 0.61, 1.18 m; Dh = 3, 4, 6, 8 mm; Tin = 30, 50, 70 °C. The use of the diagram to scope whether a research reactor is OFI-limited (below the curve) or CHF-limited based on the five parameters of its coolant channel is described. Justification for application of the diagram to research reactors with axially non-uniform heat flux is provided. Due to its limitations (uncertainties not included), the diagram cannot replace the detailed thermal-hydraulic analysis required for a reactor safety analysis. In order to make the OFI-CHF intersection diagram, two world-class CHF prediction methods (the Hall-Mudawar correlation and the extended Groeneveld 2006 table) are compared for 216 combinations of the five independent parameters. The two widely used OFI correlations (the Saha- Zuber and the Whittle-Forgan with η = 32.5) are also compared for the same combinations of the five parameters. The extended Groeneveld table and the Whittle-Forgan OFI correlation are selected for use in making the diagram. Using the above five design parameters, a research reactor can be represented by a point on the reversal diagram, and the diagram can be used to scope, without a thermal-hydraulic calculation, whether the OFI will occur before the CHF, or the CHF will occur before the OFI when the reactor power is increased keeping the five parameters fixed.

  15. Proceedings of first SWCR-KURRI academic seminar on research reactors and related research topics

    International Nuclear Information System (INIS)

    Kimura, Itsuro; Cong, Zhebao

    1986-01-01

    These are the proceedings of an academic seminar on research reactors and related research topics held at the Southwest Centre for Reactor Engineering Research and Design in Chengdu, Sichuan, People's Republic of China in September 24-26 in 1985. Included are the chairmen's addresses and 10 papers presented at the seminar in English. The titles of these papers are: (1) Nuclear Safety and Safeguards, (2) General Review of Thorium Research in Japanese Universities, (3) Comprehensive Utilization and Economic Analysis of the High Flux Engineering Test Reactor, (4) Present States of Applied Health Physics in Japan, (5) Neutron Radiography with Kyoto University Reactor, (6) Topics of Experimental Works with Kyoto University Reactor, (7) Integral Check of Nuclear Data for Reactor Structural Materials, (8) The Reactor Core, Physical Experiments and the Operation Safety Regulation of the Zero Energy Thermal Reactor for PWR Nuclear Power Plant, (9) HFETR Core Physical Parameters at Power, (10) Physical Consideration for Loads of Operated Ten Cycles in HFETR. (author)

  16. Reliability database of IEA-R1 Brazilian research reactor: Applications to the improvement of installation safety

    International Nuclear Information System (INIS)

    Oliveira, P.S.P.; Tondin, J.B.M.; Martins, M.O.; Yovanovich, M.; Ricci Filho, W.

    2010-01-01

    In this paper the main features of the reliability database being developed at Ipen-Cnen/SP for IEA-R1 reactor are briefly described. Besides that, the process for collection and updating of data regarding operation, failure and maintenance of IEA-R1 reactor components is presented. These activities have been conducted by the reactor personnel under the supervision of specialists in Probabilistic Safety Analysis (PSA). The compilation of data and subsequent calculation are based on the procedures defined during an IAEA Coordinated Research Project which Brazil took part in the period from 2001 to 2004. In addition to component reliability data, the database stores data on accident initiating events and human errors. Furthermore, this work discusses the experience acquired through the development of the reliability database covering aspects like improvements in the reactor records as well as the application of the results to the optimization of operation and maintenance procedures and to the PSA carried out for IEA-R1 reactor. (author)

  17. Status of Dalat research reactor and progress of new reactor plan in Vietnam

    International Nuclear Information System (INIS)

    Dien, Nguyen Nhi; Vien, Luong Ba

    2005-01-01

    The Dalat Nuclear Research Reactor (DNRR) is a 500-kW pool-type reactor loaded with the Soviet WWR-M2 Fuel Assemblies (FA), moderated and cooled by light water. The reactor was reconstructed from the USA 250-kW TRIGA Mark-II reactor built in early 1960s. The first criticality of the renovated reactor was achieved on 1 st November 1983, and then on 20 March 1984 the reactor was officially inaugurated and its activities restarted. During the last twenty years, the DNRR has played an important role as a large national research facility to implement researches and applications, and its utilization has been broadened in various fields of human life. However, due to the limitation of the neutron flux and power level, the out-of date design of the experimental facilities and the ageing of the reactor facilities, it cannot meet the increasing user's demands even in the existing utilization areas. In addition, the utilization demands of the Research Reactor (RR) will be increased along with the development of the nation's economy growth. In this aspect, it is necessary to have in Vietnam a new high performance multipurpose RR with a sufficient neutron flux and power level. According to the last draft of a national strategy for atomic energy development submitted to the Government for consideration and approval, it is expected that a new high power RR would be put into operation before 2020. The operation and utilization status of the DNRR is presented and some preliminary results of the national research project on new reactor plan for Vietnam are discussed in this paper

  18. Gas-cooled reactors

    International Nuclear Information System (INIS)

    Schulten, R.; Trauger, D.B.

    1976-01-01

    Experience to date with operation of high-temperature gas-cooled reactors has been quite favorable. Despite problems in completion of construction and startup, three high-temperature gas-cooled reactor (HTGR) units have operated well. The Windscale Advanced Gas-Cooled Reactor (AGR) in the United Kingdom has had an excellent operating history, and initial operation of commercial AGRs shows them to be satisfactory. The latter reactors provide direct experience in scale-up from the Windscale experiment to fullscale commercial units. The Colorado Fort St. Vrain 330-MWe prototype helium-cooled HTGR is now in the approach-to-power phase while the 300-MWe Pebble Bed THTR prototype in the Federal Republic of Germany is scheduled for completion of construction by late 1978. THTR will be the first nuclear power plant which uses a dry cooling tower. Fuel reprocessing and refabrication have been developed in the laboratory and are now entering a pilot-plant scale development. Several commercial HTGR power station orders were placed in the U.S. prior to 1975 with similar plans for stations in the FRG. However, the combined effects of inflation, reduced electric power demand, regulatory uncertainties, and pricing problems led to cancellation of the 12 reactors which were in various stages of planning, design, and licensing

  19. The heavy water accountancy for research reactors in JAERI

    International Nuclear Information System (INIS)

    Yoshijima, Tetsuo; Tanaka, Sumitoshi; Nemoto, Denjirou

    1998-11-01

    The three research reactors have been operated by the Department of Research Reactor and used about 41 tons heavy water as coolant, moderator and reflector of research reactors. The JRR-2 is a tank type research reactor of 10MW in thermal power and its is used as moderator, coolant and reflector about 16 tons heavy water. The JRR-3M is a light water cooled and moderated pool type research reactor with a thermal power of 20MW and its is used as reflector about 7.3 tons heavy water. In the JRR-4, which is a light water cooled swimming pool type research reactor with the maximum thermal power of 3.5MW, about 1 ton heavy water is used to supply fully thermalized neutrons with a neutron beam experiment of facility. The heavy water was imported from U.S.A., CANADA and Norway. Parts of heavy water is internationally controlled materials, therefore management of heavy water is necessary for materials accountancy. This report described the change of heavy water inventories in each research reactors, law and regulations for accounting of heavy water in JAERI. (author)

  20. Future plans on the Kyoto University Research Reactor (KUR)

    International Nuclear Information System (INIS)

    Shibata, Seiichi

    2000-01-01

    The Research Reactor Institute (RRI), Kyoto University, for aiming at performing the 'Experiments using a reactor and its related research', was established in Showa 38 (1963) as a cooperative research institute for universities and so on in allover Japan. Operation using KUR of one of main facilities in RRI was started by 1 MW of its rated output in 1964, and converted to 5 MW in 1968, after which through development , addition and modification of various research apparatus it has been proposed to the cooperative application researches with universities and so on in allover Japan, hitherto. Among these periods, its research organization is improved to six departments containing twenty divisions and two attached research facilities to progress some investigations on future plans at RRI for response to new researching trends. Here were described on present state of research on use of low concentrated uranium fuels at research reactor, and future plans on neutron factory and hybrid reactor. The former aims at establishment of a new research facility capable of alternating to KUR for future academic research on research reactor containing high quality and high degree application of neutron field and safety management and feature upgrading of nuclear energy. And, the latter aims at development on an accelerator drive uncritical reactor combined an accelerator neutron source and an uncritical reactor. (G.K.)

  1. IAEA activities on research reactor safety

    International Nuclear Information System (INIS)

    Alcala-Ruiz, F.

    1995-01-01

    Since its inception in 1957, the International Atomic Energy Agency (IAEA) has included activities in its programme to address aspects of research reactors such as safety, utilization and fuel cycle considerations. These activities were based on statutory functions and responsibilities, and on the current situation of research reactors in operation around the world; they responded to IAEA Member States' general or specific demands. At present, the IAEA activities on research reactors cover the above aspects and respond to specific and current issues, amongst which safety-related are of major concern to Member States. The present IAEA Research Reactor Safety Programme (RRSP) is a response to the current situation of about 300 research reactors in operation in 59 countries around the world. (orig.)

  2. 25th birthday of the first criticality of IPR-R1 nuclear research reactor

    International Nuclear Information System (INIS)

    Tofani, P.C.; Stasiulevicius, R.; Roedel, G.

    1988-01-01

    The historical evolution of IPR-R1 research reactor of Instituto de Pesquisas Radioativas-Nuclebras, since the data of its first criticality, is presented. The modifications and the main activities carried out, are presented. (M.C.K.) [pt

  3. Safety status of Russian research reactors

    International Nuclear Information System (INIS)

    Morozov, S.I.

    2001-01-01

    Gosatomnadzor of Russia is conducting the safety regulation and inspection activity related to nuclear and radiation safety at nuclear research facilities, including research reactors, critical assemblies and sub-critical assemblies. It implies implementing three major activities: 1) establishing the laws and safety standards in the field of research reactors nuclear and radiation safety; 2) research reactors licensing; and 3) inspections (or license conditions tracking and inspection). The database on nuclear research facilities has recently been updated based on the actual status of all facilities. It turned out that many facilities have been shutdown, whether temporary or permanently, waiting for the final decision on their decommissioning. Compared to previous years the situation has been inevitably changing. Now we have 99 nuclear research facilities in total under Gosatomnadzor of Russia supervision (compared to 113 in previous years). Their distribution by types and operating organizations is presented. The licensing and conduct of inspection processes are briefly outlined with emphasis being made on specific issues related to major incidents that happened in 2000, spent fuel management, occupational exposure, effluents and emissions, emergency preparedness and physical protection. Finally, a summary of problems at current Russian research facilities is outlined. (author)

  4. Application of NDT and ISI to research reactor in the Czech Republic

    International Nuclear Information System (INIS)

    Peterka, F.

    2001-01-01

    Full text: The objectives of the proposed research project are: (a) to review the present status of ISI and NDT to VR-1 and LVR-15 research reactors. (b) to be involved in the development of the ISI programme for VR-1 and LVR-15 research reactors and medium and high power research reactors of WWER type. Dr. Peterka briefly described the activities on the VR-1 and LWR-15 reactors and presented an example of a procedure to apply liquid penetrant testing. (author)

  5. Design of a multipurpose research reactor

    International Nuclear Information System (INIS)

    Sanchez Rios, A.A.

    1990-01-01

    The availability of a research reactor is essential in any endeavor to improve the execution of a nuclear programme, since it is a very versatile tool which can make a decisive contribution to a country's scientific and technological development. Because of their design, however, many existing research reactors are poorly adapted to certain uses. In some nuclear research centres, especially in the advanced countries, changes have been made in the original designs or new research prototypes have been designed for specific purposes. These modifications have proven very costly and therefore beyond the reach of developing countries. For this reason, what the research institutes in such countries need is a single sufficiently versatile nuclear plant capable of meeting the requirements of a nuclear research programme at a reasonable cost. This is precisely what a multipurpose reactor does. The Mexican National Nuclear Research Institute (ININ) plans to design and build a multipurpose research reactor capable at the same time of being used for the development of reactor design skills and for testing nuclear materials and fuels, for radioisotopes production, for nuclear power studies and basic scientific research, for specialized training, and so on. For this design work on the ININ Multipurpose Research Reactor, collaborative relations have been established with various international organizations possessing experience in nuclear reactor design: Atomehnergoeksport of the USSR: Atomic Energy of Canada Limited (AECL); General Atomics (GA) of the USA; and Japan Atomic Energy Research Institute

  6. Improving nuclear safety at international research reactors: The Integrated Research Reactor Safety Enhancement Program (IRRSEP)

    International Nuclear Information System (INIS)

    Huizenga, David; Newton, Douglas; Connery, Joyce

    2002-01-01

    Nuclear energy continues to play a major role in the world's energy economy. Research and test reactors are an important component of a nation's nuclear power infrastructure as they provide training, experiments and operating experience vital to developing and sustaining the industry. Indeed, nations with aspirations for nuclear power development usually begin their programs with a research reactor program. Research reactors also are vital to international science and technology development. It is important to keep them safe from both accident and sabotage, not only because of our obligation to prevent human and environmental consequence but also to prevent corresponding damage to science and industry. For example, an incident at a research reactor could cause a political and public backlash that would do irreparable harm to national nuclear programs. Following the accidents at Three Mile Island and Chernobyl, considerable efforts and resources were committed to improving the safety posture of the world's nuclear power plants. Unsafe operation of research reactors will have an amplifying effect throughout a country or region's entire nuclear programs due to political, economic and nuclear infrastructure consequences. (author)

  7. IAEA programme on research reactor safety

    International Nuclear Information System (INIS)

    Alcala, F.; Di Meglio, A.F.

    1995-01-01

    This paper describes the IAEA programme on research reactor safety and includes the safety related areas of conversions to the use of low enriched uranium (LEU) fuel. The program is based on the IAEA statutory responsibilities as they apply to the requirements of over 320 research reactors operating around the world. The programme covers four major areas: (a) the development of safety documents; (b) safety missions to research reactor facilities; (c) support of research programmes on research reactor safety; (d) support of Technical Cooperation projects on research reactor safety issues. The demand for these activities by the IAEA member states has increased substantially in recent years especially in developing countries with increasing emphasis being placed on LEU conversion matters. In response to this demand, the IAEA has undertaken an extensive programme for each of the four areas above. (author)

  8. IAEA activities in the field of research reactors safety

    International Nuclear Information System (INIS)

    Ciuculescu, C.; Boado Magan, H.J.

    2004-01-01

    IAEA activities in the field of research reactor safety are included in the programme of the Division of Nuclear Installations Safety. Following the objectives of the Division, the results of the IAEA missions and the recommendations from International Advisory Groups, the IAEA has conducted in recent years a certain number of activities aiming to enhance the safety of research reactors. The following activities will be presented: (a) the new Requirements for the Safety of Research Reactors, main features and differences with previous standards (SS-35-S1 and SS-35-S2) and the grading approach for implementation; (b) new documents being developed (safety guides, safety reports and TECDOC's); (c) activities related to the Incident Reporting System for Research Reactor (IRSRR); (d) the new features implemented for the INSARR missions; (e) the Code of Conduct on the Safety of Research Reactors adopted by the Board of Governors on 8 March 2004, following the General Conference Resolution GC(45)/RES/10; and (f) the survey on the safety of research reactors published on the IAEA website on February 2003 and the results obtained. (author)

  9. European Research Reactor Conference (RRFM) 2015: Conference Proceedings

    International Nuclear Information System (INIS)

    2015-01-01

    In 2015 the European Research Reactor Conference, RRFM, took place in Bucharest, Romania. The conference programme resolved around a series of plenary sessions dedicated to the latest global developments with regards to research reactor technology and management. Parallel sessions focused on all areas of the fuel cycle of research reactors, their utilisation, operation and management as well as new research reactor projects and Innovative methods in reactor physics and thermo-hydraulics. The European Research Reactor Conference also gave special attention to safety and security of research reactors

  10. European Research Reactor Conference (RRFM) 2016: Conference Proceedings

    International Nuclear Information System (INIS)

    2016-01-01

    The 2016 European Research Reactor Conference, RRFM, took place in Berlin, Germany. The conference programme resolved around a series of plenary sessions dedicated to the latest global developments with regards to research reactor technology and management. Parallel sessions focused on all areas of the fuel cycle of research reactors, their utilisation, operation and management as well as new research reactor projects and Innovative methods in reactor physics and thermo-hydraulics. The European Research Reactor Conference also gave special attention to safety and security of research reactors.

  11. Burnable poisons in the light water reactor design, microburnup experiments and calculations. Part of a coordinated programme on burnup calculations and experiments for thermal reactors

    International Nuclear Information System (INIS)

    Penndorf, K.

    1976-04-01

    Investigations on Research Agreement N 1519/CF (1.8.1974 - 31.7.1975) entitled ''Burnable poisons in light water reactor design, microburnup experiments and calculations'' were carried out in the frame of the IAEA's coordinated research programme on ''Burn-up calculation and experiments for thermal reactors''. The theoretical and experimental work on application of solid burnable poison used for reduction of the amount of boric acid necessary to control of PWR or to lower the number of control rods needed in a BWR. Solid burnable poisons are needed in present PWR designs for the reduction of the boron acid concentration in order to prevent positive coefficients of reactivity. The special operational conditions of a ship reactor lead to the application of this kind of poison for compensation of almost all burnup reactivity. This strengthens the necessity of a very accurate and many dimensional calculations because an appropriate binding of reactivity has to be kept over the whole cycle time. Several burnup experiments had been run in the 15 MW material test reactor FRG-II. The following devices have been irradiated: poison pins within and without PWR fuel pin lattice segments and fuel pins containing pellets with a poison core. Measurements of reactivity, fluence, fission product concentration have been performed. Methods applied were γ-scanning and neutron pulse, radiography and transmission measurement techniques. Evaluation of the experiments was done by one and two dimensional Ssub(N) transport burnup calculations. In parallel a collision probability transport burnup code for current PWR design work is being developed, the main feature of which is economy in manpower and computer time

  12. The role of research reactor and its future

    International Nuclear Information System (INIS)

    Nakagome, Yoshihiro

    2005-01-01

    About a half century passed since the start of operation of research reactors. Many research reactors were stopped their operation or decommissioned. With the practical use of nuclear energy, the meaning of research reactor has been buried in oblivion in the developed countries. Furthermore, under the nuclear weapons nonproliferation policy, the use of high enriched uranium fuel in research reactors is obliged to change to the use of low enriched uranium fuel. In such severe situation, this paper refers to the role of the research reactor once more through the operation experience of university-owned research reactor KUR (Kyoto University Reactor, Japan) and describes that research reactor is indispensable for the preparation to the second coming nuclear age. (author)

  13. The market for research reactors

    International Nuclear Information System (INIS)

    Roegler, H.J.

    1986-01-01

    The assay deals with some basic questions if there is an international market for research reactors at all, which influencing factors affect this market, and if research reactors have any effects on the future market for nuclear engineering. (UA) [de

  14. The GKSS cold neutron source

    International Nuclear Information System (INIS)

    Knop, W.; Wedderien, T.; Krull, W.

    1995-01-01

    The FRG-1 research reactor, in operation since 1958 at 5 MW power, is upgraded and refurbished many times to follow the changing demands on safe operation and the today needs for scientific research. This requires during the lifetime of the reactor many measures to follow these demands. Within the last years many additional activities have been made to overcome the ageing of the experiments, to change the experimental facilities and to increase the neutron flux and adapt the neutron spectrum to ensure good scientific utilization of the research reactor for the next 15 to 20 years. (orig./HP)

  15. Present status of research reactor decommissioning programme in Indonesia

    International Nuclear Information System (INIS)

    Suripto, A.; Mulyanto, N.

    2002-01-01

    At present Indonesia has 3 research reactors, namely the 30 MW MTR-type multipurpose reactor at Serpong Site, two TRIGA-type research reactors, the first one being 1 MW located at Bandung Site and the second one a small reactor of 100 kW at Yogyakarta Site. The TRIGA Reactor at the Bandung Site reached its first criticality at 250 kW in 1964, and then was operated at 1000 kW since 1971. In October 2000 the reactor power was successfully upgraded to 2 MW. This reactor has already been operated for 38 years. There is not yet any decision for the decommissioning of this reactor. However it will surely be an object for the near future decommissioning programme and hence anticipation for the above situation becomes necessary. The regulation on decommissioning of research reactor is already issued by the independent regulatory body (BAPETEN) according to which the decommissioning permit has to be applied by the BATAN. For Indonesia, an early decommissioning strategy for research reactor dictates a restricted re-use of the site for other nuclear installation. This is based on high land price, limited availability of radwaste repository site, and other cost analysis. Spent graphite reflector from the Bandung TRIGA reactor is recommended for a direct disposal after conditioning, without any volume reduction treatment. Development of human resources, technological capability as well as information flow from and exchange with advanced countries are important factors for the future development of research reactor decommissioning programme in Indonesia. (author)

  16. Power transients of Ghana research reactor-1 using PARET/ANL thermal hydraulic code

    International Nuclear Information System (INIS)

    Ampomah-Amoaka, E.; Akaho, E.H.K.; Anim-Sampong, S.; Nyarko, B.J.B.

    2010-01-01

    PARET/ANL(Version 7.3 of 2007) thermal-hydraulic code was used to perform transient analysis of the Ghana Research Reactor-1.The reactivities inserted were 2.1mk and 4mk.The peak power of 5.81kW was obtained for 2.1 mk insertion whereas the peak power for 4mk insertion of reactivity was 92.32kW.These results compare closely with experiments and theoretical studies conducted previously.

  17. Measures aimed at enhancing safe operation of the Nigeria Research Reactor-1 (NIRR-1)

    International Nuclear Information System (INIS)

    Balogun, G.I.; Jonah, S.A.; Umar, I.M.

    2005-01-01

    Safety culture has been defined as 'that assembly of characteristics and attitudes in organizations and individuals which establishes that as an overriding priority, nuclear plant safety issues receive the attention warranted by their significance'. This paper briefly highlights efforts being made at the Centre for Energy Research and Training (CERT) towards realizing this broad objective as far as possible. To this end CERT realizes the need for instituted safety measures to reflect significant, site-specific peculiar characteristics of any generic reactor types. Consequently, standard procedures for pre-startup, startup and shutdown of NIRR-1 (a miniature neutron source reactor - MNSR) have been reviewed to reflect our local conditions and peculiarities. The review has revealed the need to incorporate important steps that impact on overall safety of the facility. For instance an interlocking system is being considered between NIRR-1 startup on the one hand and mandatory pre-startup measures on the other. Also a procedure has been put in place that would facilitate rapid response in the event of a rod-stuck-at-full-withdrawal incident. Furthermore, a program of automation of important analysis and design calculations of MNSRs is going on. Emphases are also placed, and deliberate efforts are being made, to ensure that a working atmosphere prevails that would foster the correct attitudinal approach to matters of reactor safety. A regime of constant dialogue and discussions amongst operating personnel has been factored into the overall operational program. (author)

  18. Concept and realization of quality control in FRG

    International Nuclear Information System (INIS)

    Odoj, R.; Martens, B.R.; Warnecke, E.

    1987-01-01

    Radioactive waste packages to be disposed of in FRG must meet the waste acceptance requirement of the repository. Compliance with the waste acceptance requirements must be checked by a quality control. Fourteen relevant properties are necessary to be assayed. The quality control group supports the observance of the waste acceptance requirements by the following measures: -checking of the documentation of the waste producers; -random tests at conditioned radioactive wastes; -qualification of conditioning processes; -control of the qualified processes by inspections. The relevant properties of disposal are termed and their checking by non-destructive and if necessary destructive test methods will be explained as well as quality control by process control, i.e. qualification of conditioning processes by instrumentation and documentation. The mobile and stationary testing facilities will be outlined

  19. Safety in the utilization and modification of research reactors

    International Nuclear Information System (INIS)

    1994-01-01

    This Safety Guide presents guidelines, approved by international consensus, for the safe utilization and modification of research reactors. While the Guide is most applicable to existing reactors, it is also recommended for use by organizations planning to put a new reactor into operation. 1 fig

  20. Guidelines for the review research reactor safety. Reference document for IAEA Integrated Safety Assessment of Research Reactors (INSARR)

    International Nuclear Information System (INIS)

    1997-01-01

    In 1992, the IAEA published new safety standards for research reactors as part of the set of publications considered by its Research Reactor Safety Programme (RRSP). This set also includes publications giving guidance for all safety aspects related to the lifetime of a research reactor. In addition, the IAEA has also revised the Safety Standards for radiation protection. Consequently, it was considered advisable to revise the Integrated Safety Assessment of Research Reactors (INSARR) procedures to incorporate the new requirements and guidance as well as to extend the scope of the safety reviews to currently operating research reactors. The present report is the result of this revision. The purpose of this report is to give guidance on the preparation, execution, reporting and follow-up of safety review mission to research reactors as conducted by the IAEA under its INSARR missions safety service. However, it will also be of assistance to operators and regulators in conducting: (a) ad hoc safety assessments of research reactors to address individual issues such as ageing or safety culture; and (b) other types of safety reviews such as internal and peer reviews and regulatory inspections

  1. Operational parameters study of IPR-R1 TRIGA research reactor using virtual instruments

    International Nuclear Information System (INIS)

    Pinto, Antonio Juscelino; Mesquita, Amir Zacarias; Lameiras, Fernando Soares

    2013-01-01

    The instrumentation of nuclear reactors is designed with the principle of reliability, redundancy and diversification of control systems. Reliable monitoring of the parameters involved in the chain reaction is of great importance regarding efficiency and operational safety of the installation. The main goal of the simulation system in this proposed paper is to provide the study and improvement in understanding how these operational variables are interrelated and their behavior especially those related to neutronic and thermohydraulics. The work will be developed using the software LabVIEW ® (Laboratory Virtual Instruments Engineering Workbench). The program will enable the study of the variables involved in the operation of the installation throughout its operating range, for instance, a few mW up to 250 kW. The IPR-R1 TRIGA is a research nuclear reactor placed in open pool and cooled by light water with natural circulation. It is located at the Nuclear Technology Development Center (CDTN), in Belo Horizonte Brazil. The developing system employs the modern concept of virtual instruments (VIs), using microprocessors and visual interface on video monitors. LabVIEW ® breaks the paradigm of text-based programming language, for programming based on icons. The system will enable the use of this reactor in training and personnel training in the nuclear field. The work follows the recommendations of the International Atomic Energy Agency (IAEA), which has encouraged its members to develop strategic plans in order to use their research reactors. (author)

  2. Operational parameters study of IPR-R1 TRIGA research reactor using virtual instruments

    Energy Technology Data Exchange (ETDEWEB)

    Pinto, Antonio Juscelino; Mesquita, Amir Zacarias; Lameiras, Fernando Soares, E-mail: ajp@cdtn.br, E-mail: amir@cdtn.br, E-mail: fsl@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2013-07-01

    The instrumentation of nuclear reactors is designed with the principle of reliability, redundancy and diversification of control systems. Reliable monitoring of the parameters involved in the chain reaction is of great importance regarding efficiency and operational safety of the installation. The main goal of the simulation system in this proposed paper is to provide the study and improvement in understanding how these operational variables are interrelated and their behavior especially those related to neutronic and thermohydraulics. The work will be developed using the software LabVIEW ® (Laboratory Virtual Instruments Engineering Workbench). The program will enable the study of the variables involved in the operation of the installation throughout its operating range, for instance, a few mW up to 250 kW. The IPR-R1 TRIGA is a research nuclear reactor placed in open pool and cooled by light water with natural circulation. It is located at the Nuclear Technology Development Center (CDTN), in Belo Horizonte Brazil. The developing system employs the modern concept of virtual instruments (VIs), using microprocessors and visual interface on video monitors. LabVIEW ® breaks the paradigm of text-based programming language, for programming based on icons. The system will enable the use of this reactor in training and personnel training in the nuclear field. The work follows the recommendations of the International Atomic Energy Agency (IAEA), which has encouraged its members to develop strategic plans in order to use their research reactors. (author)

  3. Current trends in and prospects for development of Russian research reactors

    International Nuclear Information System (INIS)

    Arkhangelsky, N.V.; Cherepnin, Yu.S.; Gabaraev, B.A.; Khmelshchikov, V.V.; Tretiyakov, I.T.

    2004-01-01

    Over more than fifty years, many research reactors were built to Russian designs both at home and abroad, which is a considerable contribution to the world reactor engineering. Russian research reactors proved to be successful to an extent that it was found possible to raise capacity and to extend the range of their application. Though having a fairly long operating record, the majority of Russian research reactors are far from the end of their service life and are still in active use. In 2000, the 'Strategy of nuclear power development in Russia in the first half of the 21 st century' was elaborated and approved. The requirements of national nuclear power and the possible ways of its development identified in this document called for assessing the existing research capabilities. The findings of such assessment are presented in this report. The main conclusion lies in the following. On the one hand, the number and experimental capabilities of domestic research reactors are sufficient for coping with the objectives of research in, and on the other hand, retrofitting and upgrading appear to be the most expedient way of managing the operation of research reactors in the near term. Activities are under way to upgrade and extend the service life of multipurpose reactors, such as MIR-M1, SM-3, IRV-1M, BOR-60, IVV-2M, and others. The Federal Agency of the Russian Federation for Atomic Energy (Rosatom) supports the development of reactors intended for fundamental research with the use of neutron beams. To this end, Rosatom renders financial and professional support with a view to complete the PIK reactor construction at PIYaF and the IBR-2 reactor upgrades at JINR. In a longer term, the development of research reactors in Russia is expected to have the following pattern: - a small number of high-flux testing reactors with up-to-date experimental facilities located on the sites of the existing research centers; - PIK reactor, catering to domestic and foreign needs for beam

  4. Design and construction of multi research reactor

    International Nuclear Information System (INIS)

    1985-05-01

    This is the report about design and construction of multi research reactor, which introduces the purpose and necessity of the project, business contents, plan of progress of project and budget for the project. There are three appendixes about status of research reactor in other country, a characteristic of research reactor, three charts about evaluation, process and budget for the multi research reactor and three drawings for the project.

  5. Replacement Nuclear Research Reactor: Draft Environmental Impact Statement. Vol. 1. Main report

    International Nuclear Information System (INIS)

    1998-07-01

    The Draft Environmental Impact Statement (EIS) for the replacement of the Australian Research reactor has been released. An important objective of the EIS process is to ensure that all relevant information has been collected and assessed so that the Commonwealth Government can make an informed decision on the proposal. The environmental assessment of the proposal to construct and operate a replacement reactor described in the Draft EIS has shown that the scale of environmental impacts that would occur would be acceptable, provided that the management measures and commitments made by ANSTO are adopted. Furthermore, construction and operation of the proposed replacement reactor would result in a range of benefits in health care, the national interest, scientific achievement and industrial capability. It would also result in a range of benefits derived from increased employment and economic activity. None of the alternatives to the replacement research reactor considered in the Draft EIS can meet all of the objectives of the proposal. The risk from normal operations or accidents has been shown to be well within national and internationally accepted risk parameters. The dose due to reactor operations would continue to be small and within regulatory limits. For the replacement reactor, the principle of 'As Low As Reasonably Achievable' would form an integral part of the design and licensing process to ensure that doses to operators are minimized. Costs associated with the proposal are $286 million (in 1997 dollars) for design and construction. The annual operating and maintenance costs are estimated to be $12 million per year, of which a significant proportion will be covered by commercial activities. The costs include management of the spent fuel from the replacement reactor as well as the environmental management costs of waste management, safety and environmental monitoring. Decommissioning costs for the replacement reactor would arise at the end of its lifetime

  6. Nuclear safety requirements for operation licensing of Egyptian research reactors

    International Nuclear Information System (INIS)

    Ahmed, E.E.M.; Rahman, F.A.

    2000-01-01

    From the view of responsibility for health and nuclear safety, this work creates a framework for the application of nuclear regulatory rules to ensure safe operation for the sake of obtaining or maintaining operation licensing for nuclear research reactors. It has been performed according to the recommendations of the IAEA for research reactor safety regulations which clearly states that the scope of the application should include all research reactors being designed, constructed, commissioned, operated, modified or decommissioned. From that concept, the present work establishes a model structure and a computer logic program for a regulatory licensing system (RLS code). It applies both the regulatory inspection and enforcement regulatory rules on the different licensing process stages. The present established RLS code is then applied to the Egyptian Research Reactors, namely; the first ET-RR-1, which was constructed and still operating since 1961, and the second MPR research reactor (ET-RR-2) which is now in the preliminary operation stage. The results showed that for the ET-RR-1 reactor, all operational activities, including maintenance, in-service inspection, renewal, modification and experiments should meet the appropriate regulatory compliance action program. Also, the results showed that for the new MPR research reactor (ET-RR-2), all commissioning and operational stages should also meet the regulatory inspection and enforcement action program of the operational licensing safety requirements. (author)

  7. Research reactor back-end options - decommissioning: a necessary consideration

    International Nuclear Information System (INIS)

    England, M.R.; Parry, D.R.; Smith, C.

    1998-01-01

    Decommissioning is a challenge, which all radioactive site licensees eventually need to face and research reactors are no exception. BNFL has completed numerous major decommissioning projects at its own operational sites and has undertaken similar works at customers' sites including the decommissioning of the Universities Research Reactor (URR), Risley and the ICI TRIGA 1-Mk I Reactor at Billingham. Based on the execution of such projects BNFL has gained an understanding of the variety of customer requirements and the effectiveness of specific decommissioning techniques for research reactors. This paper addresses factors to be considered when reviewing the way forward following shut down and how these affect the final decisions for fuel management and the extent of decommissioning. Case studies are described from BNFL's recent experience decommissioning both the URR and ICI TRIGA reactors. (author)

  8. RA research reactor, Part 1, Operation and maintenance of the RA nuclear reactor for 1985

    International Nuclear Information System (INIS)

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

    1985-01-01

    According to the plan, RA reactor was to be in operation in mid September 1985. But, since the building of the emergency cooling system, nor the reconstruction of the existing special ventilation system were not finished until the end of August reactor was not operated during 1985. During the previous four years reactor operation was limited by the temporary operating license issued by the Committee of Serbian ministry for health and social care, which was cancelled in August 1984. The reason was the non existing emergency cooling system and lack of appropriate filters in the special ventilation system. This temporary license has limited the reactor power to 2 MW from 1981-1984. Control and maintenance of the reactor instrumentation and tools was done regularly but dependent on the availability of the spare parts. In order to enable future reliable operation of the RA reactor, according to new licensing regulations, during 1984, three major tasks have started: building of the new emergency system, reconstruction of the existing ventilation system, and renewal of the reactor instrumentation. IAEA has approved the amount of 1,300,000 US dollars for the renewal of the instrumentation [sr

  9. Nuclear reactor safety research in Idaho

    International Nuclear Information System (INIS)

    Zeile, H.J.

    1983-01-01

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

  10. The future role of research reactors

    International Nuclear Information System (INIS)

    Glaeser, W.

    2001-01-01

    The decline of neutron source capacity in the next decades urges for the planning and construction of new neutron sources for basic and applied research with neutrons. Modern safety precautions of research reactors make them competitive with other ways of neutron production using non-chain reactions for many applications. Research reactors consequently optimized offer a very broad range of possible applications in basic and applied research. Research reactors at universities also in the future have to play an important role in education and training in basic and applied nuclear science. (orig.)

  11. Back-end of the research reactor fuel cycle

    International Nuclear Information System (INIS)

    Gruber, Gehard J.

    1996-01-01

    This paper outlines the status of topics and issues related to: (1) Research Reactor Spent Nuclear Fuel Return to the U.S., including policy, shipments and ports of entry, management sites, fees, storage technologies, contracts, actual shipment, and legal process, (2) UKAEA: MTR Spent Nuclear Fuel Reprocessing, (3) COGEMA: MTR Spent Nuclear Fuel Reprocessing, and (4) Intermediate Storage + Direct Disposal for Research Reactors. (author)

  12. Progress report concerning safety research for nuclear reactor facilities

    International Nuclear Information System (INIS)

    1978-01-01

    Examination and evaluation of safety research results for nuclear reactor facilities have been performed, as more than a year has elapsed since the plan had been initiated in April, 1976, by the special sub-committee for the safety of nuclear reactor facilities. The research is carried out by being divided roughly into 7 items, and seems to be steadily proceeding, though it does not yet reach the target. The above 7 items include researches for (1) criticality accident, (2) loss of coolant accident, (3) safety for light water reactor fuel, (4) construction safety for reactor facilities, (5) reduction of release of radioactive material, (6) safety evaluation based on the probability theory for reactor facilities, and (7) aseismatic measures for reactor facilities. With discussions on the progress and the results of the research this time, research on the behaviour on fuel in abnormal transients including in-core and out-core experiments has been added to the third item, deleting the power-cooling mismatch experiment in Nuclear Safety Research Reactor of JAERI. Also it has been decided to add two research to the seventh item, namely measured data collection, classification and analysis, and probability assessment of failures due to an earthquake. For these 7 items, the report describes the concrete contents of research to be performed in fiscal years of 1977 and 1978, by discussing on most rational and suitable contents conceivable at present. (Wakatsuki, Y.)

  13. The utility of different reactor types for the research

    International Nuclear Information System (INIS)

    Stiennon, G.

    1983-01-01

    The report presents a general view of the use of the different belgian research reactor i.e. venus reactor, BR-1 reactor, BR-2 reactor and BR-3 reactor. Particular attention is given to the programmes which is in the interest of international collaboration. In order to reach an efficient utilization of such reactors they require a specialized personnel groups to deal with the irradiation devices and radioactive materials and post irradiation examinations, creating a complete material testing station. (A.J.)

  14. IRSN research programs concerning reactor safety

    International Nuclear Information System (INIS)

    Bardelay, J.

    2005-01-01

    This paper is made up of 3 parts. The first part briefly presents the missions of IRSN (French research institute on nuclear safety), the second part reviews the research works currently led by IRSN in the following fields : -) the assessment of safety computer codes, -) thermohydraulics, -) reactor ageing, -) reactivity accidents, -) loss of coolant, -) reactor pool dewatering, -) core meltdown, -) vapor explosion, and -) fission product release. In the third part, IRSN is shown to give a major importance to experimental programs led on research or test reactors for collecting valid data because of the complexity of the physical processes that are involved. IRSN plans to develop a research program concerning the safety of high or very high temperature reactors. (A.C.)

  15. The replacement research reactor

    International Nuclear Information System (INIS)

    Cameron, R.

    1999-01-01

    As a consequences of the government decision in September 1997. ANSTO established a replacement research reactor project to manage the procurement of the replacement reactor through the necessary approval, tendering and contract management stages This paper provides an update of the status of the project including the completion of the Environmental Impact Statement. Prequalification and Public Works Committee processes. The aims of the project, management organisation, reactor type and expected capabilities are also described

  16. New research reactor proposed for Australia

    International Nuclear Information System (INIS)

    Anon.

    1992-01-01

    A new research reactor has been proposed for construction within the next ten years, to replace the HIFAR reactor which operating capabilities have been over taken by later designs. This paper outlines the main research applications of the new reactor design and briefly examines issues related to its cost, economic benefits, safety and location

  17. Reactor protection systems for the Replacement Research Reactor, ANSTO

    International Nuclear Information System (INIS)

    Morris, C.R.

    2003-01-01

    The 20-MW Replacement Research Reactor Project which is currently under construction at ANSTO will have a combination of a state of the art triplicated computer based reactor protection system, and a fully independent, and diverse, triplicated analogue reactor protection system, that has been in use in the nuclear industry, for many decades. The First Reactor Protection System (FRPS) consists of a Triconex triplicated modular redundant system that has recently been approved by the USNRC for use in the USA?s power reactor program. The Second Reactor Protection System is a hardwired analogue system supplied by Foxboro, the Spec 200 system, which is also Class1E qualified. The FRPS is used to drop the control rods when its safety parameter setpoints have been reached. The SRPS is used to drain the reflector tank and since this operation would result in a reactor poison out due to the time it would take to refill the tank the FRPS trip setpoints are more limiting. The FRPS and SRPS have limited hardwired indications on the control panels in the main control room (MCR) and emergency control centre (ECC), however all FRPS and SRPS parameters are capable of being displayed on the reactor control and monitoring system (RCMS) video display units. The RCMS is a Foxboro Series I/A control system which is used for plant control and monitoring and as a protection system for the cold neutron source. This paper will provide technical information on both systems, their trip logics, their interconnections with each other, and their integration into the reactor control and monitoring system and control panels. (author)

  18. Research reactor modernization and refurbishment

    International Nuclear Information System (INIS)

    2009-08-01

    Many recent, high profile research reactor unplanned shutdowns can be directly linked to different challenges which have evolved over time. The concept of ageing management is certainly nothing new to nuclear facilities, however, these events are highlighting the direct impact unplanned shutdowns at research reactors have on various stakeholders who depend on research reactor goods and services. Provided the demand for these goods and services remains strong, large capital projects are anticipated to continue in order to sustain future operation of many research reactors. It is within this context that the IAEA organized a Technical Workshop to launch a broader Agency activity on research reactor modernization and refurbishment (M and R). The workshop was hosted by the operating organization of the HOR Research Reactor in Delft, the Netherlands, in October 2006. Forty participants from twenty-three countries participated in the meeting: with representation from Africa, Asia Pacific, Eastern Europe, North America, South America and Western Europe. The specific objectives of this workshop were to present facility reports on completed, existing and planned M and R projects, including the project objectives, scope and main characteristics; and to specifically report on: - the project impact (planned or actual) on the primary and key supporting motivation for the M and R project; - the project impact (planned or actual) on the design basis, safety, and/or regulatory-related reports; - the project impact (planned or actual) on facility utilization; - significant lessons learned during or following the completion of M and R work. Contributions from this workshop were reviewed by experts during a consultancy meeting held in Vienna in December 2007. The experts selected final contributions for inclusion in this report. Requests were also distributed to some authors for additional detail as well as new authors for known projects not submitted during the initial 2006 workshop

  19. Consideration of BORAX-type reactivity accidents applied to research reactors

    International Nuclear Information System (INIS)

    Couturier, Jean; Meignen, Renaud; Bourgois, Thierry; Biaut, Guillaume; Mireau, Jean-Pierre; Natta, Marc

    2011-01-01

    Most of the research reactors discussed in this document are pool-type reactors in which the reactor vessel and some of the reactor coolant systems are located in a pool of water. These reactors generally use fuel in plate assemblies formed by a compact layer of uranium (or U 3 Si 2 ) and aluminium particles, sandwiched between two thin layers of aluminium serving as cladding. The fuel melting process begins at 660 deg. C when the aluminium melts, while the uranium (or U 3 Si 2 ) particles may remain solid. The accident that occurred in the American SL-1 reactor in 1961, together with tests carried out in the United States as of 1954 in the BORAX-1 reactor and then, in 1962, in the SPERT-1 reactor, showed that a sudden substantial addition of reactivity in this type of reactor could lead to explosive mechanisms caused by degradation, or even fast meltdown, of part of the reactor core. This is what is known as a 'BORAX-type' accident. The aim of this document is first to briefly recall the circumstances of the SL-1 reactor accident, the lessons learned, how this operational feedback has been factored into the design of various research reactors around the world and, second, to describe the approach taken by France with regard to this type of accident and how, led by IRSN, this approach has evolved in the last decade. (authors)

  20. Light water reactor safety research project

    International Nuclear Information System (INIS)

    Markoczy, G.; Aksan, S.N.; Behringer, K.; Prodan, M.; Stierli, F.; Ullrich, G.

    1980-07-01

    The research and development activities for the safety of Light Water Power Reactors carried out 1979 at the Swiss Federal Institute for Reactor Research are described. Considerations concerning the necessity, objectives and size of the Safety Research Project are presented, followed by a detailed discussion of the activities in the five tasks of the program, covering fracture mechanics and nondestructive testing, thermal-hydraulics, reactor noise analysis and pressure vessel steel surveillance. (Auth.)

  1. Utilization of MVP for research on fast reactor

    International Nuclear Information System (INIS)

    Yokoyama, Kenji

    2001-01-01

    Utilization of the continuous energy Monte-Carlo code, MVP, for research on fast reactor in Power Reactor and Nuclear Fuel Development Corporation(PNC) is described. In this report, three types of utilization are reviewed; (1) a comparison of the eigenvalues calculated by MVP with the results by the deterministic methods, (2) an improvement of U-238 reaction rate evaluation in JUPITER experimental Analysis and (3) an evaluation of heterogeneity effects for Am reaction rates of the moderated subassemblies. Since the results of MVP can be used as references, MVP is very useful code in research on fast reactor. It is one of indispensable tools in order to verify the models in the deterministic methods. Furthermore, it can be used so as to investigate the new concept reactors, such as a reactor aiming to transmute minor actinides(MA). On the other hand, a problem of the variance reduction remains. Especially, a small reactivity cannot be estimated by MVP because of large variances. The development of a Monte-Carlo method for a small reactivity calculation will promote the utilization of MVP for research on fast reactor. (author)

  2. Cooling Performance of Natural Circulation for a Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Suki; Chun, J. H.; Yum, S. B. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    This paper deals with the core cooling performance by natural circulation during normal operation and a flow channel blockage event in an open tank-in-pool type research reactor. The cooling performance is predicted by using the RELAP5/ MOD3.3 code. The core decay heat is usually removed by natural circulation to the reactor pool water in open tank-in-pool type research reactors with the thermal power less than several megawatts. Therefore, these reactors have generally no active core cooling system against a loss of normal forced flow. In reactors with the thermal power less than around one megawatt, the reactor core can be cooled down by natural circulation even during normal full power operation. The cooling performance of natural circulation in an open tank-in-pool type research reactor has been investigated during the normal natural circulation and a flow channel blockage event. It is found that the maximum powers without void generation at the hot channel are around 1.16 MW and 820 kW, respectively, for the normal natural circulation and the flow channel blockage event.

  3. Reactor materials research as an effective instrument of nuclear reactor perfection

    International Nuclear Information System (INIS)

    Baryshnikov, M.

    2006-01-01

    The work is devoted to reactor materiology, as to the practical tool of nuclear reactor development. The work is illustrated with concrete examples from activity experience of the appropriate division of the Russian Research Centre Kurchatov Institute - Institute of Reactor Materials Research and Radiation Nanotechnologies. Besides the description of some modern potentials of the mentioned institute is given. (author)

  4. 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

  5. Utilization of research reactors

    International Nuclear Information System (INIS)

    1962-01-01

    About 200 research reactors are now in operation in different parts of the world, and at least 70 such facilities, which are in advanced stages of planning and construction, should be critical within the next two or three years. In the process of this development a multitude of problems are being encountered in formulating and carrying out programs for the proper utilization of these facilities, especially in countries which have just begun or are starting their atomic energy work. An opportunity for scientific personnel from different Member States to discuss research reactor problems was given at an international symposium on the Programing and Utilization of Research Reactors organized by the Agency almost immediately after the General Conference session. Two hundred scientists from 35 countries, as well as from the European Nuclear Energy Agency and EURATOM, attended the meeting which was held in Vienna from 16 to 21 October 1961

  6. United States Domestic Research Reactor Infrastructure TRIGA Reactor Fuel Support

    International Nuclear Information System (INIS)

    Morrell, Douglas

    2011-01-01

    The United State Domestic Research Reactor Infrastructure Program at the Idaho National Laboratory manages and provides project management, technical, quality engineering, quality inspection and nuclear material support for the United States Department of Energy sponsored University Reactor Fuels Program. This program provides fresh, unirradiated nuclear fuel to Domestic University Research Reactor Facilities and is responsible for the return of the DOE-owned, irradiated nuclear fuel over the life of the program. This presentation will introduce the program management team, the universities supported by the program, the status of the program and focus on the return process of irradiated nuclear fuel for long term storage at DOE managed receipt facilities. It will include lessons learned from research reactor facilities that have successfully shipped spent fuel elements to DOE receipt facilities.

  7. Annual report of Department of Research Reactor, 1997. April 1, 1997 - March 31, 1998

    International Nuclear Information System (INIS)

    1999-01-01

    The Department of Research Reactor is responsible for the operation, maintenance, utilization and related R and D works of the research reactors including JRR-2, JRR-3M (new JRR-3) and JRR-4. This report describes the activities of the department in fiscal year of 1997 and it also includes some of the technical topics on the works mentioned above. As for the research reactors, we carried out the operation, maintenance, the utilization of irradiation and neutron beam experiments, technical management including fuels and water chemistry, radiation monitoring as related R and D works. The international cooperations between the developing countries and the department were also made concerning the operation, utilization and safety analysis for nuclear facilities. (author)

  8. Annual report of department of research reactor, 1995 (April 1, 1995 - March 31, 1996)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-12-01

    The Department of Research Reactor is responsible for the operation, maintenance, utilization and related R and D works of the research reactors including JRR-2, JRR-3M (new JRR-3) and JRR-4. This report describes the activities of our department in fiscal year of 1995 and it also includes some of the technical topics on the works mentioned above. As for the research reactors, we carried out the operation, maintenance, irradiation utilization, neutron beam experiments, technical management including fuels and water chemistry, radiation monitoring as related R and D works. The international cooperations between the developing countries and our department were also made concerning the operation, utilization and safety analysis for nuclear facilities. (author)

  9. Annual report of Department of Research Reactor, 1996. April 1, 1996 - March 31, 1997

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-03-01

    The Department of Research Reactor is responsible for the operation, maintenance, utilization and related R and D works of the research reactors including JRR-2, JRR-3M (new JRR-3) and JRR-4. This report describes the activities of our department in fiscal year of 1996 and it also includes some of the technical topics on the works mentioned above. As for the research reactors, we carried out the operation, maintenance, irradiation utilization, neutron beam experiments, technical management including fuels and water chemistry, radiation monitoring as related R and D works. The international cooperations between the developing countries and our department were also made concerning the operation, utilization and safety analysis for nuclear facilities. (author)

  10. Current status and prospects of research reactors

    International Nuclear Information System (INIS)

    Gabaraev, A.B.; Cherepnin, Yu.S.; Tretyakov, I.T.; Khmelshikov, V.V.; Dollezhal, N.A.

    2009-01-01

    Full text: The first nuclear research reactors (RR) appeared in the 1940s. Their initial purpose was to provide knowledge of the main processes associated with neutron-induced nuclear reactions. Later, the rang of problems addressed expanded substantially. Besides fundamental research in the properties of matter, such reactors are successfully used for dealing with problems in the fields of materials science, nuclear engineering, medicine, isotope production, education, etc. Over the whole period of RR fleet growth, more than six hundred nuclear research facilities were built in 70 countries of the world. As of the end of 2008, the number of Russian research reactors in service was about 20% of the globally operating RR fleet. This paper discusses the current status of the world's RR fleet and describes the capabilities of the experimental reactor facilities existing in Russia. In the 21st century, research reactors will remain in demand to solve scientific and technological problems for innovative development of society. The emerging renaissance of nuclear power, the expanding RR uses for production of isotopes and other applications, the increase in the number of countries willing to use nuclear technologies in energy production, industry and science - all contribute to a rebirth of interest in research reactors. One of the ways to improve the experimental capabilities lies in radical upgrading of the reactor facilities with qualitative changes in the main neutronic characteristics of the core. The associated design approaches are illustrated with the example of the IBR-2M reactor at the JNRI in Dubna. The imperative need restricting the spread of nuclear threat leads us to give up using highly enriched uranium in most research reactors. Development of RR fuel with reduced enrichment in uranium has been one of the priority objectives of NIKIET for many years. This paper presents the latest results obtained along these lines, as applied to pool-type research

  11. Rationalization and future planning for AECL's research reactor capability

    International Nuclear Information System (INIS)

    Slater, J.B.

    1990-01-01

    AECL's research reactor capability has played a crucial role in the development of Canada's nuclear program. All essential concepts for the CANDU reactors were developed and tested in the NRX and NRU reactors, and in parallel, important contributions to basic physics were made. The technical feasibility of advanced fuel cycles and of the organic-cooled option for CANDU reactors were also demonstrated in the two reactors and the WR-1 reactor. In addition, an important and growing radio-isotope production industry was established and marketed on a world-wide basis. In 1984, however, it was recognized that a review and rationalization of the research reactor capability was required. The commercial success of the CANDU reactor system had reduced the scope and size of the required development program. Limited research and development funding and competition from other research facilities and programs, required that the scope be reduced to a support basis essential to maintain strategic capability. Currently, AECL, is part-way through this rationalization program and completion should be attained during 1992/93 when the MAPLE reactor is operational and decisions on NRX decommissioning will be made. A companion paper describes some of the unique operational and maintenance problems which have resulted from this program and the solutions which have been developed. Future planning must recognize the age of the NRU reactor (currently 32 years) and the need to plan for eventual replacement. Strategy is being developed and supporting studies include a full technical assessment of the NRU reactor and the required age-related upgrading program, evaluation of the performance characteristics and costs of potential future replacement reactors, particularly the advanced MAPLE concept, and opportunities for international co-operation in developing mutually supportive research programs

  12. Preventive and Predictive Maintenance, Warehousing of Spares, Periodic Testing and In-Service Inspection Activities at the Nigerian Research Reactor-1 Facility

    Energy Technology Data Exchange (ETDEWEB)

    Yusuf, I.; Mati, A. A.; Dewu, B. B.M., [Centre for Energy Research and Training, Ahmadu Bello University, Zaria (Nigeria)

    2014-08-15

    The Nigerian Research Reactor–1, or NIRR-1, is sited at Centre for Energy Research and Training, Ahmadu Bello University, Zaria, Nigeria. Activities on preventive or routine maintenance have been institutionalized since the commissioning of the reactor in February 2004. This has grossly reduced the rates of corrective maintenance activities and helped the reactor management a great deal in predicting failure rates of reactor components and other auxiliary units. Routine maintenance of systems and components are being carried out on a weekly, quarterly and annual basis based on manufacturer’s recommendations, which have been reviewed and improved over the years. The paper presents the implementation of maintenance activities in NIRR-1 from its initial criticality in 2004 till today and the new scheme for periodic testing and in-service-inspection developed after an IAEA Integrated Safety Assessment of Research Reactors mission. The measures put in place are envisaged to reduce the negative impact of ageing on NIRR-1 and its auxiliary systems. (author)

  13. In-service inspection of pool type research reactors

    International Nuclear Information System (INIS)

    Rajamani, K.

    2002-01-01

    In the case of Apsara Reactor, it has been proposed to carry out major modifications in the near future. It is planned to modify the core suitably with a heavy water reflector tank to demonstrate the Multiple Purpose Research Reactor concept. The core structure will be a stationary one and will be located at the 'B' position of the pool. The modified reactor will be operated at 1 MW power level. Suitable methodologies are evolved for carrying out a planned ISI for this modified reactor

  14. Research reactor safety - an overview of crucial aspects

    International Nuclear Information System (INIS)

    Laverie, M.

    1998-01-01

    Chronology of the commissioning orders of the French research reactors illustrates the importance of the time factor. When looking at older reactors, one must, on one hand, demonstrate, not only the absence of risks tied to the reactor's ageing, but, on the other hand, adapt the reactor's original technical designs to today's safety practices and standards. The evolution of reactor safety requirements over the last twenty years sometimes makes this adaptation difficult. The design of the next research reactors, after a one to two decades pause in construction, will require to set up new safety assessment bases that will have to take into account the nuclear power plant safety evolution. As a general statement, research reactor safety approaches will require the incorporation of specific design rules for research reactors: experience feedback for one of a kind design, frequent modifications required by research programmes, special operational requirements with operators/researchers interfaces. (author)

  15. Meeting on reactor safety research

    International Nuclear Information System (INIS)

    1982-09-01

    The meeting 'Reactor Safety Research' organized for the second time by the GRS by order of the BMFT gave a review of research activities on the safety of light water reactors in the Federal Repulbic of Germany, international co-operation in this field and latest results of this research institution. The central fields of interest were subjects of man/machine-interaction, operational reliability accident sequences, and risk. (orig.) [de

  16. Education and research at the VR-1 Vrabec training reactor facility

    International Nuclear Information System (INIS)

    Matejka, K.

    1993-01-01

    The results of 12 years' efforts devoted to the construction of the VR-1 ''Vrabec'' training reactor at the Faculty of Nuclear Science and Physical Engineering, Czech Technical University in Prague and to establishing the training reactor department, as well as the contribution of the training reactor facility to the teaching and scientific activities of the Faculty are presented in a comprehensive manner. The thesis is divided into 2 parts: (i) preconditions, reactor construction and commissioning, and constituting the reactor department, and (ii) basic and comprehensive information concerning the current utilization of the reactor for the benefit of students from various university level institutions. The prospects of scientific activities of the department are also outlined. Attention is paid to selected nuclear safety aspects of the reactor during operation and teaching of students, as well as to its innovated digital control system whose implementation is planned. The results achieved are compared with the initial goals and with similar experience abroad. (P.A.)

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

    International Nuclear Information System (INIS)

    1988-01-01

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

  18. Summary of IEA-R1 research a reactor licensing related to its power increase from 2 to 10 MW

    International Nuclear Information System (INIS)

    1989-04-01

    This work is a summary of IEA-R1 research reactor licensing related to its power increase from 2 to 10 MW. It reports also safety requirements, fuel elements, and reactor control modifications inherent to power increase. (A.C.A.S.)

  19. Replacement Nuclear Research Reactor: Draft Environmental Impact Statement. Vol. 1. Main report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-07-01

    The Draft Environmental Impact Statement (EIS) for the replacement of the Australian Research reactor has been released. An important objective of the EIS process is to ensure that all relevant information has been collected and assessed so that the Commonwealth Government can make an informed decision on the proposal. The environmental assessment of the proposal to construct and operate a replacement reactor described in the Draft EIS has shown that the scale of environmental impacts that would occur would be acceptable, provided that the management measures and commitments made by ANSTO are adopted. Furthermore, construction and operation of the proposed replacement reactor would result in a range of benefits in health care, the national interest, scientific achievement and industrial capability. It would also result in a range of benefits derived from increased employment and economic activity. None of the alternatives to the replacement research reactor considered in the Draft EIS can meet all of the objectives of the proposal. The risk from normal operations or accidents has been shown to be well within national and internationally accepted risk parameters. The dose due to reactor operations would continue to be small and within regulatory limits. For the replacement reactor, the principle of `As Low As Reasonably Achievable` would form an integral part of the design and licensing process to ensure that doses to operators are minimized. Costs associated with the proposal are $286 million (in 1997 dollars) for design and construction. The annual operating and maintenance costs are estimated to be $12 million per year, of which a significant proportion will be covered by commercial activities. The costs include management of the spent fuel from the replacement reactor as well as the environmental management costs of waste management, safety and environmental monitoring. Decommissioning costs for the replacement reactor would arise at the end of its lifetime

  20. Current status of the world's research reactors

    International Nuclear Information System (INIS)

    Dodd, B.

    1999-01-01

    Data from the IAEA's Research Reactor Database (RRDB) provides information with respect to the status of the world's research reactors. Some summary data are given. Recent initiatives by the IAEA regarding communications and information flow with respect to research reactors are discussed. Future plans and perspectives are also introduced. (author)

  1. Research reactor records in the INIS database

    International Nuclear Information System (INIS)

    Marinkovic, N.

    2001-01-01

    This report presents a statistical analysis of more than 13,000 records of publications concerned with research and technology in the field of research and experimental reactors which are included in the INIS Bibliographic Database for the period from 1970 to 2001. The main objectives of this bibliometric study were: to make an inventory of research reactor related records in the INIS Database; to provide statistics and scientific indicators for the INIS users, namely science managers, researchers, engineers, operators, scientific editors and publishers, decision-makers in the field of research reactors related subjects; to extract other useful information from the INIS Bibliographic Database about articles published in research reactors research and technology. (author)

  2. Effective utilization and management of research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Muranaka, R [International Atomic Energy Agency, Vienna (Austria). Div. of Research and Isotopes

    1984-06-01

    The problem of utilizing a research reactor effectively is closely related to its management and therefore should not be considered separately. Too often, attention has been focused on specific techniques and methods rather than on the overall programme of utilization, with the result that skills and equipment have been acquired without any active continuing programme of applications and services. The seminar reported here provided a forum for reactor managers, users, and operators to discuss their experience. At the invitation of the Government of Malaysia, it was held at the Asia Pacific Development Centre, Kuala Lumpur, from 7 to 11 November 1983. It was attended by about 50 participants from 19 Member States; it is hoped that a report on the seminar, including papers presented, can be published and thus reach a wider audience. Thirty-one lectures and contributions were presented at a total of seven sessions: Research reactor management; Radiation exposure and safety; Research reactor utilization (two sessions); PUSPATI Research Reactor Project Development; Core conversion to low-enriched uranium, and safeguards; Research reactor technology. In addition, a panel discussed the causes and resolutions of the under-utilization of research reactors.

  3. Effective utilization and management of research reactors

    International Nuclear Information System (INIS)

    Muranaka, R.

    1984-01-01

    The problem of utilizing a research reactor effectively is closely related to its management and therefore should not be considered separately. Too often, attention has been focused on specific techniques and methods rather than on the overall programme of utilization, with the result that skills and equipment have been acquired without any active continuing programme of applications and services. The seminar reported here provided a forum for reactor managers, users, and operators to discuss their experience. At the invitation of the Government of Malaysia, it was held at the Asia Pacific Development Centre, Kuala Lumpur, from 7 to 11 November 1983. It was attended by about 50 participants from 19 Member States; it is hoped that a report on the seminar, including papers presented, can be published and thus reach a wider audience. Thirty-one lectures and contributions were presented at a total of seven sessions: Research reactor management; Radiation exposure and safety; Research reactor utilization (two sessions); PUSPATI Research Reactor Project Development; Core conversion to low-enriched uranium, and safeguards; Research reactor technology. In addition, a panel discussed the causes and resolutions of the under-utilization of research reactors

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  5. IGORR-1: Proceedings of the first meeting of the international group on research reactors

    International Nuclear Information System (INIS)

    West, C.D.

    1990-05-01

    Many organizations, in several countries, are planning or implementing new or upgraded research reactor projects, but there has been no organized forum devoted entirely to discussion and exchange of information in this field. Over the past year or so, informal discussions resulted in widespread agreement that such a forum would serve a useful purpose. Accordingly, a proposal to form a group was submitted to the leading organizations known to be involved in projects to build or upgrade reactor facilities. Essentially all agreed to join in the formation of the International Group on Research Reactors (IGORR) and nominated a senior staff member to serve on its international organizing committee. The first IGORR meeting took place on February 28--March 2, 1990. It was very successful and well attended; some 52 scientists and engineers from 25 organizations in 10 countries participated in 2-1/2 days of open and informative presentations and discussions. Two workshop sessions offered opportunities for more detailed interaction among participants and resulted in identification of common R ampersand D needs, sources of data, and planned new facilities. Individual papers have been cataloged separately

  6. Present status and future prospect of research reactors

    International Nuclear Information System (INIS)

    Takemi, Hirokatsu

    1996-01-01

    The present status of research reactors more than MW class reactor in JAERI and the Kyoto University and the small reactors in the Musashi Institute of Technology, the Rikkyo University, the Tokyo University, the Kinki University and other countries are explained in the paper. The present status of researches are reported by the topics in each field. The future researches of the beam reactor and the irradiation reactor are reviewed. On various kinds of use of research reactor and demands of neutron field of a high order, new type research reactors under investigation are explained. Recently, the reactors are used in many fields such as the basic science: the basic physics, the material science, the nuclear physics, and the nuclear chemistry and the applied science; the earth and environmental science, the biology and the medical science. (S.Y.)

  7. Comparison among creep rupture strength extrapolation methods with application to data for AISI 316 SS from Italy, France, U.K. and F.R.G

    International Nuclear Information System (INIS)

    Brunori, G.; Cappellato, S.; Vacchiano, S.; Guglielmi, F.

    1982-01-01

    Inside Activity 3 ''Materials'' of WGCS, the member states UK and FRG have developed a work regarding extrapolation methods for creep data. This work has been done by comparising extrapolation methods in use in their countries by applying them to creep rupture strength data on AISI 316 SS obtained in UK and FRG. This work has been issued on April 1978 and the Community has dealed it to all Activity 3 Members. Italy, in the figure of NIRA S.p.A., has received, from the European Community a contract to extend the work to Italian and French data, using extrapolation methods currently in use in Italy. The work should deal with the following points: - Collect of Italian experimental data; - Chemical analysis on Italian Specimen; - Comparison among Italian experimental data with French, FRG and UK data; - Description of extrapolation methods in use in Italy; - Application of these extrapolation methods to Italian, French, British and Germany data; - Extensions of a Final Report

  8. Moving ring reactor 'Karin-1'

    International Nuclear Information System (INIS)

    1983-12-01

    The conceptual design of a moving ring reactor ''Karin-1'' has been carried out to advance fusion system design, to clarify the research and development problems, and to decide their priority. In order to attain these objectives, a D-T reactor with tritium breeding blanket is designed, a commercial reactor with net power output of 500 MWe is designed, the compatibility of plasma physics with fusion engineering is demonstrated, and some other guideline is indicated. A moving ring reactor is composed mainly of three parts. In the first formation section, a plasma ring is formed and heated up to ignition temperature. The plasma ring of compact torus is transported from the formation section through the next burning section to generate fusion power. Then the plasma ring moves into the last recovery section, and the energy and particles of the plasma ring are recovered. The outline of a moving ring reactor ''Karin-1'' is described. As a candidate material for the first wall, SiC was adopted to reduce the MHD effect and to minimize the interaction with neutrons and charged particles. The thin metal lining was applied to the SiC surface to solve the problem of the compatibility with lithium blanket. Plasma physics, the engineering aspect and the items of research and development are described. (Kako, I.)

  9. Proceedings of the sixth Asian symposium on research reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-08-01

    The symposium consisted of 16 sessions with 58 submitted papers. Major fields were: (1) status and future plan of research and testing reactors, (2) operating experiences, (3) design and modification of the facility, and reactor fuels, (4) irradiation studies, (5) irradiation facilities, (6) reactor characteristics and instrumentation, and (7) neutron beam utilization. Panel discussion on the 'New Trends on Application of Research and Test Reactors' was also held at the last of the symposium. About 180 people participated from China, Korea, Indonesia, Thailand, Bangladesh, Vietnam, Chinese Taipei, Belgium, France, USA, Japan and IAEA. The 58 of the presented papers are indexed individually. (J.P.N.)

  10. Proceedings of the sixth Asian symposium on research reactors

    International Nuclear Information System (INIS)

    1999-08-01

    The symposium consisted of 16 sessions with 58 submitted papers. Major fields were: 1) status and future plan of research and testing reactors, 2) operating experiences, 3) design and modification of the facility, and reactor fuels, 4) irradiation studies, 5) irradiation facilities, 6) reactor characteristics and instrumentation, and 7) neutron beam utilization. Panel discussion on the 'New Trends on Application of Research and Test Reactors' was also held at the last of the symposium. About 180 people participated from China, Korea, Indonesia, Thailand, Bangladesh, Vietnam, Chinese Taipei, Belgium, France, USA, Japan and IAEA. The 58 of the presented papers are indexed individually. (J.P.N.)

  11. Research reactor safety - an overview of crucial aspects

    Energy Technology Data Exchange (ETDEWEB)

    Laverie, M. [Atomic Energy Commission, Saclay, F-91191 Gif sur Yvette (France)

    1998-07-01

    Chronology of the commissioning orders of the French research reactors illustrates the importance of the time factor. When looking at older reactors, one must, on one hand, demonstrate, not only the absence of risks tied to the reactor's ageing, but, on the other hand, adapt the reactor's original technical designs to today's safety practices and standards. The evolution of reactor safety requirements over the last twenty years sometimes makes this adaptation difficult. The design of the next research reactors, after a one to two decades pause in construction, will require to set up new safety assessment bases that will have to take into account the nuclear power plant safety evolution. As a general statement, research reactor safety approaches will require the incorporation of specific design rules for research reactors: experience feedback for one of a kind design, frequent modifications required by research programmes, special operational requirements with operators/researchers interfaces. (author)

  12. New research reactor for Australia

    International Nuclear Information System (INIS)

    Miller, R.

    1992-01-01

    HIFAR, Australia's major research reactor was commissioned in 1958 to test materials for an envisaged indigenous nuclear power industry. HIFAR is a Dido type reactor which is operated at 10 MW. With the decision in the early 1970's not to proceed to nuclear power, HIFAR was adapted to other uses and has served Australia well as a base for national nuclear competence; as a national facility for neutron scattering/beam research; as a source of radioisotopes for medical diagnosis and treatment; and as a source of export revenue from the neutron transmutation doping of silicon for the semiconductor industry. However, all of HIFAR's capabilities are becoming less than optimum by world and regional standards. Neutron beam facilities have been overtaken on the world scene by research reactors with increased neutron fluxes, cold sources, and improved beams and neutron guides. Radioisotope production capabilities, while adequate to meet Australia's needs, cannot be easily expanded to tap the growing world market in radiopharmaceuticals. Similarly, neutron transmutation doped silicon production, and export income from it, is limited at a time when the world market for this material is expanding. ANSTO has therefore embarked on a program to replace HIFAR with a new multi-purpose national facility for nuclear research and technology in the form of a reactor: a) for neutron beam research, - with a peak thermal flux of the order of three times higher than that from HIFAR, - with a cold neutron source, guides and beam hall, b) that has radioisotope production facilities that are as good as, or better than, those in HIFAR, c) that maximizes the potential for commercial irradiations to offset facility operating costs, d) that maximizes flexibility to accommodate variations in user requirements during the life of the facility. ANSTO's case for the new research reactor received significant support earlier this month with the tabling in Parliament of a report by the Australian Science

  13. International benchmark study of advanced thermal hydraulic safety analysis codes against measurements on IEA-R1 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hainoun, A., E-mail: pscientific2@aec.org.sy [Atomic Energy Commission of Syria (AECS), Nuclear Engineering Department, P.O. Box 6091, Damascus (Syrian Arab Republic); Doval, A. [Nuclear Engineering Department, Av. Cmdt. Luis Piedrabuena 4950, C.P. 8400 S.C de Bariloche, Rio Negro (Argentina); Umbehaun, P. [Centro de Engenharia Nuclear – CEN, IPEN-CNEN/SP, Av. Lineu Prestes 2242-Cidade Universitaria, CEP-05508-000 São Paulo, SP (Brazil); Chatzidakis, S. [School of Nuclear Engineering, Purdue University, West Lafayette, IN 47907 (United States); Ghazi, N. [Atomic Energy Commission of Syria (AECS), Nuclear Engineering Department, P.O. Box 6091, Damascus (Syrian Arab Republic); Park, S. [Research Reactor Design and Engineering Division, Basic Science Project Operation Dept., Korea Atomic Energy Research Institute (Korea, Republic of); Mladin, M. [Institute for Nuclear Research, Campului Street No. 1, P.O. Box 78, 115400 Mioveni, Arges (Romania); Shokr, A. [Division of Nuclear Installation Safety, Research Reactor Safety Section, International Atomic Energy Agency, A-1400 Vienna (Austria)

    2014-12-15

    Highlights: • A set of advanced system thermal hydraulic codes are benchmarked against IFA of IEA-R1. • Comparative safety analysis of IEA-R1 reactor during LOFA by 7 working teams. • This work covers both experimental and calculation effort and presents new out findings on TH of RR that have not been reported before. • LOFA results discrepancies from 7% to 20% for coolant and peak clad temperatures are predicted conservatively. - Abstract: In the framework of the IAEA Coordination Research Project on “Innovative methods in research reactor analysis: Benchmark against experimental data on neutronics and thermal hydraulic computational methods and tools for operation and safety analysis of research reactors” the Brazilian research reactor IEA-R1 has been selected as reference facility to perform benchmark calculations for a set of thermal hydraulic codes being widely used by international teams in the field of research reactor (RR) deterministic safety analysis. The goal of the conducted benchmark is to demonstrate the application of innovative reactor analysis tools in the research reactor community, validation of the applied codes and application of the validated codes to perform comprehensive safety analysis of RR. The IEA-R1 is equipped with an Instrumented Fuel Assembly (IFA) which provided measurements for normal operation and loss of flow transient. The measurements comprised coolant and cladding temperatures, reactor power and flow rate. Temperatures are measured at three different radial and axial positions of IFA summing up to 12 measuring points in addition to the coolant inlet and outlet temperatures. The considered benchmark deals with the loss of reactor flow and the subsequent flow reversal from downward forced to upward natural circulation and presents therefore relevant phenomena for the RR safety analysis. The benchmark calculations were performed independently by the participating teams using different thermal hydraulic and safety

  14. Reference equilibrium core with central flux irradiation facility for Pakistan research reactor-1

    International Nuclear Information System (INIS)

    Israr, M.; Shami, Qamar-ud-din; Pervez, S.

    1997-11-01

    In order to assess various core parameters a reference equilibrium core with Low Enriched Uranium (LEU) fuel for Pakistan Research Reactor (PARR-1) was assembled. Due to increased volume of reference core, the average neutron flux reduced as compared to the first higher power operation. To get a higher neutron flux an irradiation facility was created in centre of the reference equilibrium core where the advantage of the neutron flux peaking was taken. Various low power experiments were performed in order to evaluate control rods worth and neutron flux mapping inside the core. The neutron flux inside the central irradiation facility almost doubled. With this arrangement reactor operation time was cut down from 72 hours to 48 hours for the production of the required specific radioactivity. (author)

  15. Commissioning of the new heat exchanger for the research nuclear reactor IEA-R1

    Energy Technology Data Exchange (ETDEWEB)

    Castro, Alfredo Jose Alvim de; Cassiano, Douglas Alves; Umbehaun, Pedro Ernesto; Carvalho, Marcos Rodrigues de; Frajndlich, Roberto [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)]. E-mails: ajcastro@ipen.br; docass@gmail.com.br; umbehaun@ipen.br; carvalho@ipen.br; frajndli@ipen.br

    2008-07-01

    The Research Reactor IEA-R1 placed at IPEN/CNEN-SP is of the swimming pool type, light water moderated and with graphite reflectors, and was build and designed by Babcock and Wilcox Co. Start up operation was in September the 16{sup th}, 1957, being the first criticality for South Hemisphere. Although designed to operate at 5 MW, the IEA-R1 was operated until 2001 with 2 MW and was suitable for use in basic and applied research as well as the production of medical radioisotopes, industry and natural sciences applications. Due to a recent demand increase on radioisotopes in Brazil for medical diagnoses and therapies applications, IPEN /CNEN updated the IEA-R1 power to 5 MW and to work at continuous operation regime. Studies on the Ageing Management for the Research Reactor IEA-R1 were conducted according to IAEA procedures. As result of these studies critical components within the Ageing Management Program were identified. Also were made recommendations on the implementation of test scheduling and standardization procedures to organize data and documents. One of the main results was the need of monitoring the two heat exchangers, the two primary circuit pumps and the data acquisition system. During monitoring procedures, issues were observed on the IEA-R1 operation at 5 MW mainly due to the ageing of the Babcox and Wilcox TCA heat exchanger, and excessive vibrations at high flow rates on CBC's TCB heat exchanger. So, from 2005 on, it was decided to work with 3,5 MW and provide a new IESA heat exchanger with 5 MW capacity, to substitute the TCA heat exchanger. This work presents results on the commissioning of the new heat exchanger and compares against the values calculated in the IESA project. The results show that the IEA-R1 Reactor can be operated more safety and continuously at 5 MW with the new IESA heat exchanger. (author)

  16. An overview-probabilistic safety analysis for research reactors

    International Nuclear Information System (INIS)

    Liu Jinlin; Peng Changhong

    2015-01-01

    For long-term application, Probabilistic Safety Analysis (PSA) has proved to be a valuable tool for improving the safety and reliability of power reactors. In China, 'Nuclear safety and radioactive pollution prevention 'Twelfth Five Year Plan' and the 2020 vision' raises clearly that: to develop probabilistic safety analysis and aging evaluation for research reactors. Comparing with the power reactors, it reveals some specific features in research reactors: lower operating power, lower coolant temperature and pressure, etc. However, the core configurations may be changed very often and human actions play an important safety role in research reactors due to its specific experimental requirement. As a result, there is a necessary to conduct the PSA analysis of research reactors. This paper discusses the special characteristics related to the structure and operation and the methods to develop the PSA of research reactors, including initiating event analysis, event tree analysis, fault tree analysis, dependent failure analysis, human reliability analysis and quantification as well as the experimental and external event evaluation through the investigation of various research reactors and their PSAs home and abroad, to provide the current situation and features of research reactors PSAs. (author)

  17. Training and research reactor facility longevity extension program

    International Nuclear Information System (INIS)

    Carriveau, G.W.

    1991-01-01

    Since 1943, over 550 training and research reactors have been in operation. According to statistics from the International Atomic Energy Agency, ∼325 training and research reactors are currently in service. This total includes a wide variety of designs covering a range of power and research capabilities located virtually around the world. A program has been established at General Atomics (GA) that is dedicated to the support of extended longevity of training and research reactor facilities. Aspects of this program include the following: (1) new instrumentation and control systems; (2) improved and upgraded nuclear monitoring and control channels; (3) facility testing, repair and upgrade services that include (a) pool or tank integrity, (b) cooling system, and (c) water purification system; (4) fuel element testing procedures and replacement; (5) control rod drive rebuilding and upgrades; (6) control and monitoring system calibration and repair service; (7) training services, including reactor operations, maintenance, instrumentation calibration, and repair; and (8) expanded or new uses such as neutron radiography and autoradiography, isotope production, nuclear medicine, activation analysis, and material properties modification

  18. Current status of nuclear research reactor management and utilization program in Thailand

    International Nuclear Information System (INIS)

    Aramrattana, M.; Busamongkol, Y.

    1999-01-01

    The TRR1/M1 is the first research reactor and has been in operational for more than 20 years. During the three decades of research reactor operation in Thailand the utilization of research reactor have been broadened in different fields such as agriculture, medicine and industry. Limitation on utilization of the existing reactor in various fields has led to establishing of a new nuclear research center, Ongkharak Nuclear Research Center (ONRC). The ONRC comprises three major facilities, namely Reactor Island, Isotope Production Facility and Waste Processing and Storage Facility. The reactor itself is a 10 MW TRIGA-type fuels, moderated and cooled by light water with beryllium and heavy water as the reflectors. It is a multi-purpose reactor consisting of different facilities inside and around the core for radioisotope production, medical and industrial uses; and for beam experiments such as High Resolution Powder Diffractometry (HRPD), Neutron Radiography (NR), Prompt Gamma Neutron Activation Analysis (PGNAA), and Boron Neutron Capture Therapy (BNCT). The center is expected to be operational by year 2001. (author)

  19. Research reactor fuel - an update

    International Nuclear Information System (INIS)

    Finlay, M.R.; Ripley, M.I.

    2003-01-01

    In the two years since the last ANA conference there have been marked changes in the research reactor fuel scene. A new low-enriched uranium (LEU) fuel, 'monolithic' uranium molybdenum, has shown such promise in initial trials that it may be suitable to meet the objectives of the Joint Declaration signed by Presidents Bush and Putin to commit to converting all US and Russian research reactors to LEU by 2012. Development of more conventional aluminium dispersion UMo LEU fuel has continued in the meantime and is entering the final qualification stage of multiple full sized element irradiations. Despite this progress, the original 2005 timetable for UMo fuel qualification has slipped and research reactors, including the RRR, may not convert from silicide to UMo fuel before 2007. The operators of the Swedish R2 reactor have been forced to pursue the direct route of qualifying a UMo lead test assembly (LTA) in order to meet spent fuel disposal requirements of the Swedish law. The LTA has recently been fabricated and is expected to be loaded shortly into the R2 reactor. We present an update of our previous ANA paper and details of the qualification process for UMo fuel

  20. Supply of enriched uranium for research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, H. [NUKEM GmbH, Alzenau (Germany)

    1997-08-01

    Since the RERTR-meeting In Newport/USA in 1990 the author delivered a series of papers in connection with the fuel cycle for research reactors dealing with its front-end. In these papers the author underlined the need for unified specifications for enriched uranium metal suitable for the production of fuel elements and made proposals with regard to the re-use of in Europe reprocessed highly enriched uranium. With regard to the fuel cycle of research reactors the research reactor community was since 1989 more concentrating on the problems of its back-end since the USA stopped the acceptance of spent research reactor fuel on December 31, 1988. Now, since it is apparent that these back-end problem have been solved by AEA`s ability to reprocess and the preparedness of the USA to again accept physically spent research reactor fuel the author is focusing with this paper again on the front-end of the fuel cycle on the question whether there is at all a safe supply of low and high enriched uranium for research reactors in the future.

  1. Supply of enriched uranium for research reactors

    International Nuclear Information System (INIS)

    Mueller, H.

    1997-01-01

    Since the RERTR-meeting In Newport/USA in 1990 the author delivered a series of papers in connection with the fuel cycle for research reactors dealing with its front-end. In these papers the author underlined the need for unified specifications for enriched uranium metal suitable for the production of fuel elements and made proposals with regard to the re-use of in Europe reprocessed highly enriched uranium. With regard to the fuel cycle of research reactors the research reactor community was since 1989 more concentrating on the problems of its back-end since the USA stopped the acceptance of spent research reactor fuel on December 31, 1988. Now, since it is apparent that these back-end problem have been solved by AEA's ability to reprocess and the preparedness of the USA to again accept physically spent research reactor fuel the author is focusing with this paper again on the front-end of the fuel cycle on the question whether there is at all a safe supply of low and high enriched uranium for research reactors in the future

  2. Safe operation and maintenance of research reactor

    International Nuclear Information System (INIS)

    Munsorn, S.

    1999-01-01

    The first Thai Research Reactor (TRR-1) was established in 1961 at the Office of Atomic Energy for Peace (OAEP), Bangkok. The reactor was light water moderated and cooled, using HEU plate-type with U 3 O 8 - Al fuel meat and swimming pool type. The reactor went first critical on October 27, 1962 and had been licensed to operate at 1 MW (thermal). On June 30, 1975 the reactor was shutdown for modification and the core and control system was disassemble and replaced by that of TRIGA Mark III type while the pool cooling system, irradiation facilities and other were kept. Thus the name TRR-1/M1' has been designed due to this modification the fuel has been changed from HEU plate type to Uranium Zirconium Hydride (UZrH) Low Enrichment Uranium (LEU) which include 4 Fuel Follower Control Rods and 1 Air Follower Control Rod. The TRR-1/M1 went critical on November 7, 1977 and the purpose of the operation are training, isotope production and research. Nowadays the TRR-1/M1 has been operated with core loading No.12 which released power of 1,056 MWD. (as of October 1998). The TRR-1/M1 has been operated at the power of 1.2 MW, three days a week with 34 hours per week, Shut-down on Monday for weekly maintenance and Tuesday for special experiment. The everage energy released is about 40.8 MW-hour per week. Every year, the TRR-1/M1 is shut-down about 2 months between February to March for yearly maintenance. (author)

  3. Safe operation and maintenance of research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Munsorn, S. [Reactor Operation Division, Office of Atomic Energy for Peace, Chatuchak, Bangkok (Thailand)

    1999-10-01

    The first Thai Research Reactor (TRR-1) was established in 1961 at the Office of Atomic Energy for Peace (OAEP), Bangkok. The reactor was light water moderated and cooled, using HEU plate-type with U{sub 3}O{sub 8}- Al fuel meat and swimming pool type. The reactor went first critical on October 27, 1962 and had been licensed to operate at 1 MW (thermal). On June 30, 1975 the reactor was shutdown for modification and the core and control system was disassemble and replaced by that of TRIGA Mark III type while the pool cooling system, irradiation facilities and other were kept. Thus the name TRR-1/M1' has been designed due to this modification the fuel has been changed from HEU plate type to Uranium Zirconium Hydride (UZrH) Low Enrichment Uranium (LEU) which include 4 Fuel Follower Control Rods and 1 Air Follower Control Rod. The TRR-1/M1 went critical on November 7, 1977 and the purpose of the operation are training, isotope production and research. Nowadays the TRR-1/M1 has been operated with core loading No.12 which released power of 1,056 MWD. (as of October 1998). The TRR-1/M1 has been operated at the power of 1.2 MW, three days a week with 34 hours per week, Shut-down on Monday for weekly maintenance and Tuesday for special experiment. The everage energy released is about 40.8 MW-hour per week. Every year, the TRR-1/M1 is shut-down about 2 months between February to March for yearly maintenance. (author)

  4. Annual report of department of research reactor, 2000. April 1, 2000 - March 31, 2001

    International Nuclear Information System (INIS)

    2002-02-01

    The Department of Research Reactor is responsible for the operation, Maintenance, utilization of the JRR-3 and the JRR-4 and for the related R and D. Besides RI production including its R and D are carried out. This report describes the activities of the department in fiscal year of 2000 and it also includes some of the technical topics on the works mentioned above. As for the research reactors, we carried out the operation, maintenance, the utilization of irradiation and neutron beam experiments, technical management including fuels and water chemistry, radiation monitoring as related R and D works. RI Production and its R and D works were conducted as well. The international cooperations between the developing countries and the department were also made concerning the operation, utilization and safety analysis for research reactors. Although the term 'JRR-3M' was used to denote the JRR-3M modified 1990 until the 2000 annual report of the Department of Research Reactor, the term 'JRR-3' will be used from this annual report because the JRR-3 has been operated for about 10 years since the modification and is now under further modification and upgrading study. (author)

  5. Dismantling of the reactor block of the FRJ-1 research reactor (MERLIN); Abbau des Reaktorblocks des Forschungsreaktors FRJ-1 (MERLIN)

    Energy Technology Data Exchange (ETDEWEB)

    Stahn, B.; Matela, K.; Zehbe, C. [Forschungszentrum Juelich GmbH (Germany); Poeppinghaus, J. [Gesellschaft fuer Nuklear-Service mbH, Essen (Germany); Cremer, J. [Siempelkamp Nukleartechnik GmbH, Heidelberg (Germany)

    2003-07-01

    By the end of 1998 the complete secondary cooling system and the major part of the primary cooling system were dismantled. Furthermore, the experimental devices, including a rabbit system conceived as an in-core irradiation device, were disassembled and disposed of. In total, approx. 65 t of contaminated and/or activated material as well as approx. 70 t of clearance-measured material were disposed of within the framework of these activities. The dismantling of the coolant loops and experimental devices was followed in 2000 by the removal of the reactor tank internals and the subsequent draining of the reactor tank water. The reactor tank internals were essentially the core support plate, the core box, the flow channel and the neutron flux bridges (s. Fig. 2, detailed reactor core). All components consisted of aluminium, the connecting elements such as bolts and nuts, however, of stainless steel. Due to the high activation of the core internals, disassembly had to be remotely controlled under water. All removal work was carried out from a tank intermediate floor (s. Fig. 2). These activities, which served for preparing the dismantling of the reactor block, were completed in summer 2001. The waste parts arising were transferred to the Service Department for Decontamination of the Research Centre. This included approx. 2.5 t of waste parts with a total activity of approx. 8 x 10{sup 11} Bq. (orig.)

  6. Demolition to Green-Field conditions of the FRJ-1 (MERLIN) research reactor. Successes and hurdles in the demolition of a research reactor of the megawatt class; Der Rueckbau des Forschungsreaktors FRJ-1 (MERLIN) bis zur 'Gruenen Wiese'. Erfolge und Huerden beim Rueckbau eines Forschungsreaktors der Megawatt-Klasse

    Energy Technology Data Exchange (ETDEWEB)

    Stahn, Burkhard; Printz, Rudolf; Matela, Karel; Zehbe, Carsten; Stauch, Bernhard; Zander, Iven [Forschungszentrum Juelich GmbH, Juelich (Germany)

    2010-02-15

    The Juelich-1 Research Reactor (FRJ-1), also referred to as MERLIN (Medium Energy Research Light Water Moderated Industrial Nuclear Reactor), was a light-water moderated and cooled swimming pool reactor of British design. The cornerstone in the erection of the reactor building was laid on June 11, 1958. Reactor operation was started on February 23, 1962. The plant was last run at a thermal power of 10 MW and shut down for good in 1985 after 23 years of operation. After the fuel elements had been removed and most of the experimental installations dismantled, some first steps towards demolition were taken in 1995. Demolition on a large scale began in 1996. September 8, 2008 was a special day: On the area of the former reactor hall, an oak tree was planted as a symbol of the 'green field' and of the original oak wood which had to make way for the construction of reactors in Juelich. An oak tree now stands in the place of the reactor unit. Was that all? It was not, for there were ancillary systems, operations, utility and hygiene buildings which had to be pulled down. Decontamination and clearance measurements were completed. The application for clearance was prepared and completed. Conventional demolition was started in 2009. After completion of that step, the last chapter about demolition of the FRJ-1 research reactor has been written, and the book can be closed. (orig.)

  7. Design of a digital system for operational parameters simulation of IPR-R1 TRIGA nuclear research reactor

    International Nuclear Information System (INIS)

    Lage, Aldo M.F.; Mesquita, Amir Z.; Felippe, Adriano de A.M.

    2017-01-01

    The instrumentation of nuclear reactors is designed based on the reliability, redundancy and diversification of control systems. The monitoring of the parameters is of crucial importance with regard to the operational efficiency and safety of the installation. Since the first criticality of a nuclear reactor, achieved by Fermi et al. in 1942, there has been concern about the reliable monitoring of the parameters involved in the chain reaction. This paper presents the current stage of the system of simulation, which is under development at the CDTN, which intends to simulate the operation of the TRIGA IPR-R1 nuclear reactor, involving the evolution of neutron flux and reactor power related events. The system will be developed using LabVIEW® software, using the modern concept of virtual instruments (VIs) that are visualized in a video monitor. For the implementation of this model, computational tools and systems analysis are necessary, which help and facilitate the implementation of the simulator. In this article we will show some of these techniques and the initial design of the model to be implemented. The design of a computational system is of great importance, since it guides in the implementation stages and generates the documentation for later maintenance and updating of the computational system. It is noteworthy that the innovations developed in research reactors are normally used in power reactors. The relatively low costs enable research reactors to be an excellent laboratory for developing techniques for future reactors. (author)

  8. Design of a digital system for operational parameters simulation of IPR-R1 TRIGA nuclear research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lage, Aldo M.F.; Mesquita, Amir Z.; Felippe, Adriano de A.M., E-mail: aldo@cdtn.br, E-mail: amir@cdtn.br, E-mail: adrianoamfelippe@gmail.com [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN /CNEN-MG), Belo Horizonte, MG (Brazil); Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil)

    2017-11-01

    The instrumentation of nuclear reactors is designed based on the reliability, redundancy and diversification of control systems. The monitoring of the parameters is of crucial importance with regard to the operational efficiency and safety of the installation. Since the first criticality of a nuclear reactor, achieved by Fermi et al. in 1942, there has been concern about the reliable monitoring of the parameters involved in the chain reaction. This paper presents the current stage of the system of simulation, which is under development at the CDTN, which intends to simulate the operation of the TRIGA IPR-R1 nuclear reactor, involving the evolution of neutron flux and reactor power related events. The system will be developed using LabVIEW® software, using the modern concept of virtual instruments (VIs) that are visualized in a video monitor. For the implementation of this model, computational tools and systems analysis are necessary, which help and facilitate the implementation of the simulator. In this article we will show some of these techniques and the initial design of the model to be implemented. The design of a computational system is of great importance, since it guides in the implementation stages and generates the documentation for later maintenance and updating of the computational system. It is noteworthy that the innovations developed in research reactors are normally used in power reactors. The relatively low costs enable research reactors to be an excellent laboratory for developing techniques for future reactors. (author)

  9. Semiconductor research with reactor neutrons

    International Nuclear Information System (INIS)

    Kimura, Itsuro

    1992-01-01

    Reactor neutrons play an important role for characterization of semiconductor materials as same as other advanced materials. On the other hand reactor neutrons bring about not only malignant irradiation effects called radiation damage, but also useful effects such as neutron transmutation doping and defect formation for opto-electronics. Research works on semiconductor materials with the reactor neutrons of the Kyoto University Reactor (KUR) are briefly reviewed. In this review, a stress is laid on the present author's works. (author)

  10. Research nuclear reactor RA - Annual report 1992

    International Nuclear Information System (INIS)

    Sotic, O.

    1992-12-01

    Research reactor RA Annual report for year 1992 is divided into two main parts to cover: (1) operation and maintenance and (2) activities related to radiation protection. First part includes 8 annexes describing reactor operation, activities of services for maintenance of reactor components and instrumentation, financial report and staffing. Second annex B is a paper by Z. Vukadin 'Recurrence formulas for evaluating expansion series of depletion functions' published in 'Kerntechnik' 56, (1991) No.6 (INIS record no. 23024136. Second part of the report is devoted to radiation protection issues and contains 4 annexes with data about radiation control of the working environment and reactor environment, description of decontamination activities, collection of radioactive wastes, and meteorology data [sr

  11. Growing dimensions. Spent fuel management at research reactors

    International Nuclear Information System (INIS)

    Ritchie, I.G.

    1998-01-01

    More than 550 nuclear research reactors are operating or shout down around the world. At many of these reactors, spent fuel from their operations is stored, pending decisions on its final disposition. In recent years, problems associated with this spent fuel storage have loomed larger in the international nuclear community. In efforts to determine the overall scope of problems and to develop a database on the subject, the IAEA has surveyed research reactor operators in its Member States. Information for the Research Reactor Spent Fuel Database (RRSFDB) so far has been obtained from a limited but representative number of research reactors. It supplements data already on hand in the Agency's more established Research Reactor Database (RRDB). Drawing upon these database resources, this article presents an overall picture of spent fuel management and storage at the world's research reactors, in the context of associated national and international programmes in the field

  12. Review of computer models used for post closure safety assessment of nuclear waste repositories in the FRG

    International Nuclear Information System (INIS)

    Bogorinski, P.; Baltes, B.; Martens, K.H.

    1987-01-01

    In the FRG, disposal of nuclear wastes takes place in deep geologic formations. For longterm safety assessment of such a repository, groundwater transport provides a release scenario for the radionuclides to the biosphere. GRs reviewed a methodology that was implemented by the research group of PSE to simulate migration of radionuclides in the geosphere. The examination included the applicability of theoretical models, numerical experiments, comparison to results of diverse computer codes as well as experience from international intercomparison studies. The review concluded that the hydrological model may be applied to full extent unless density effects have to be considered whereas there are some restrictions in the use of the nuclide transport model

  13. State-of-the-art report on accident analysis and risk analysis of reprocessing plants in European countries

    International Nuclear Information System (INIS)

    Nomura, Yasushi

    1985-12-01

    This report summarizes informations obtained from America, England, France and FRG concerning methodology, computer code, fundamental data and calculational model on accident/risk analyses of spent fuel reprocessing plants. As a result, the followings are revealed. (1) The system analysis codes developed for reactor plants can be used for reprocessing plants with some code modification. (2) Calculational models and programs have been developed for accidental phenomenological analyses in FRG, but with insufficient data to prove them. (3) The release tree analysis codes developed in FRG are available to estimate radioactivity release amount/probability via off-gas/exhaustair lines in the case of accidents. (4) The computer codes developed in America for reactor-plant environmental transport/safety analyses of released radioactivity can be applied to reprocessing facilities. (author)

  14. Training courses at VR-1 reactor

    International Nuclear Information System (INIS)

    Sklenka, L.; Kropik, M.

    2006-01-01

    This paper describes one of the main purposes of the VR-1 training reactor utilization - i.e. extensive educational program. The educational program is intended for the training of university students and selected nuclear power plant personnel. The training courses provide them experience in reactor and neutron physics, dosimetry, nuclear safety and operation of nuclear facilities. At present, the training course participants can go through more than 20 standard experimental exercises; particular exercises for special training can be prepared. Approximately 200 university students become familiar with the reactor (lectures, experiments, experimental and diploma works, etc.) every year. About 12 different faculties from Czech universities use the reactor. International co-operation with European universities in Germany, Hungary, Austria, Slovakia, Holland and UK is frequent. The VR-1 reactor takes also part in Eugene Wigner Course on Reactor Physics Experiments in the framework of European Nuclear Educational Network (ENEN) association. Recently, training courses for Bulgarian research reactor specialists supported by IAEA were carried out. An attractive program including demonstration of reactor operation is prepared also for high school students. Every year, more than 1500 high school students come to visit the reactor, as do many foreigner visitors. (author)

  15. Diagnostic measurement on research reactors

    International Nuclear Information System (INIS)

    Dach, K.; Zbytovsky, A.

    A comparison is made of noise experiments on zero power and power reactors. The general characteristics of noise experiments on power reactors is their ''passivity'', i.e., the experiment does not require any interruption of the normal operating regime of the reactor system. On zero power research reactors where the fission reaction constitutes the dominant noise source such conditions have to be created in the study of noise components as to make the investigated noise dominant and the noise of the fission reaction the background. The simultaneous use of both methods makes it possible to determine the spectral composition of reactivity fluctuations, which facilitates the identification of noise sources. The conditions are described of the recordability of noise components. The possibilities are listed provided for research work in Czechoslovakia and the possibility is studied of setting up an expert team to organize the respective experimental programme on an international scale. Power reactors manufactured in the GDR are considered as the suitable experimental base. (J.P.)

  16. Proceedings of the international symposium on research reactor safety operations and modifications

    International Nuclear Information System (INIS)

    1990-03-01

    The International Symposium on Research Reactor Safety, Operations and Modifications was organized by the International Atomic Energy Agency in cooperation with Atomic Energy of Canada Limited-Research Company. The main objectives of this Symposium were: (1) to exchange information and to discuss current perspectives and concerns relating to all aspects to research reactor safety, operations, and modifications; and, (2) to present views and to discuss future initiatives and directions for research reactor design, operations, utilization, and safety. The symposium topics included: research reactor programmes and experience; research reactor design safety and analysis; research reactor modifications and decommissioning; research reactor licensing; and new research reactors. These topics were covered during eight oral sessions and three poster sessions. These Proceedings include the full text of the 93 papers presented. The subject of Symposium was quite wide-ranging in that it covered essentially all aspects of research reactor safety, operations, and modifications. This was considered to be appropriate and timely given the 326 research reactors currently in operation in some 56 countries; given the degree of their utilization which ranges from pure and applied research to radioisotopes production to basic training and manpower development; and given that many of these reactors are undergoing extensive modifications, core conversions, power upratings, and are becoming the subject of safety reassessment and regulatory reviews. Although the Symposium covered many topics, the majority of papers and discussions tended to focus mainly on research reactor safety. This was seen as a clear sign of the continuing recognition of the fundamental importance of identifying and addressing, particularly through international cooperation, issues and concerns associated with research reactor safety

  17. Gas reactor international cooperative program interim report: German Pebble Bed Reactor design and technology review

    International Nuclear Information System (INIS)

    1978-09-01

    This report describes and evaluates several gas-cooled reactor plant concepts under development within the Federal Republic of Germany (FRG). The concepts, based upon the use of a proven Pebble Bed Reactor (PBR) fuel element design, include nuclear heat generation for chemical processes and electrical power generation. Processes under consideration for the nuclear process heat plant (PNP) include hydrogasification of coal, steam gasification of coal, combined process, and long-distance chemical heat transportation. The electric plant emphasized in the report is the steam turbine cycle (HTR-K), although the gas turbine cycle (HHT) is also discussed. The study is a detailed description and evaluation of the nuclear portion of the various plants. The general conclusions are that the PBR technology is sound and that the HTR-K and PNP plant concepts appear to be achievable through appropriate continuing development programs, most of which are either under way or planned

  18. Proposal to the United States Energy Research and Development Administration for continuation of fusion reactor technology studies. Progress report October 1, 1977--July 1, 1978

    Energy Technology Data Exchange (ETDEWEB)

    Conn, R.W.; Kulcinski, G.L.; Maynard, C.W.

    1978-01-01

    Since the last progress report we have concentrated on three main areas of research: (1) the study of the NUWMAK reactor design, (2) the study of rf heating for tokamak reactors, and (3) the initiation of a tandem mirror reactor study. The initial work on the tandem mirror reactor is included as background in the technical proposal. Summaries of our work on recent assessments of lithium reserves and neutral transport codes are included.

  19. Proposal to the United States Energy Research and Development Administration for continuation of fusion reactor technology studies. Progress report October 1, 1977--July 1, 1978

    International Nuclear Information System (INIS)

    Conn, R.W.; Kulcinski, G.L.; Maynard, C.W.

    1978-01-01

    Since the last progress report we have concentrated on three main areas of research: (1) the study of the NUWMAK reactor design, (2) the study of rf heating for tokamak reactors, and (3) the initiation of a tandem mirror reactor study. The initial work on the tandem mirror reactor is included as background in the technical proposal. Summaries of our work on recent assessments of lithium reserves and neutral transport codes are included

  20. Experimental research in neutron physic and thermal-hydraulic at the CDTN Triga reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mesquita, Amir Z.; Souza, Rose Mary G.P.; Ferreira, Andrea V.; Pinto, Antonio J.; Costa, Antonio C.L.; Rezende, Hugo C., E-mail: amir@cdtn.b, E-mail: souzarm@cdtn.b, E-mail: avf@cdtn.b, E-mail: ajp@cdtn.b, E-mail: aclc@cdtn.b, E-mail: hcr@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-07-01

    The IPR-R1 TRIGA (Training, Research, Isotopes production, General Atomics) at Nuclear Technology Development Center (CDTN) is a pool type reactor cooled by natural circulation of light water and an open surface. TRIGA reactors, developed by General Atomics (GA), are the most widely used research reactor in the world and characterized by inherent safety. The IPR-R1 is the only Brazilian nuclear research reactor available and able to perform experiments in which interaction between neutronic and thermal-hydraulic areas occurs. The IPR-R1 has started up on November 11th, 1960. At that time the maximum thermal power was 30 kW. The present forced cooling system was built in the 70th and the power was upgraded to 100 kW. Recently the core configuration and instrumentation was upgraded again to 250 kW at steady state, and is awaiting the license of CNEN to operate definitely at this new power. This paper describes the experimental research project carried out in the IPR-R1 reactor that has as objective evaluate the behaviour of the reactor operational parameters, and mainly to investigate the influence of temperature on the neutronic variables. The research was supported by Research Support Foundation of the State of Minas Gerais (FAPEMIG) and Brazilian Council for Scientific and Technological Development (CNPq). The research project meets the recommendations of the IAEA, for safety, modernization and development of strategic plan for research reactors utilization. This work is in line with the strategic objectives of Brazil, which aims to design and construct the Brazilian Multipurpose research Reactor (RMB). (author)

  1. Experimental research in neutron physic and thermal-hydraulic at the CDTN Triga reactor

    International Nuclear Information System (INIS)

    Mesquita, Amir Z.; Souza, Rose Mary G.P.; Ferreira, Andrea V.; Pinto, Antonio J.; Costa, Antonio C.L.; Rezende, Hugo C.

    2011-01-01

    The IPR-R1 TRIGA (Training, Research, Isotopes production, General Atomics) at Nuclear Technology Development Center (CDTN) is a pool type reactor cooled by natural circulation of light water and an open surface. TRIGA reactors, developed by General Atomics (GA), are the most widely used research reactor in the world and characterized by inherent safety. The IPR-R1 is the only Brazilian nuclear research reactor available and able to perform experiments in which interaction between neutronic and thermal-hydraulic areas occurs. The IPR-R1 has started up on November 11th, 1960. At that time the maximum thermal power was 30 kW. The present forced cooling system was built in the 70th and the power was upgraded to 100 kW. Recently the core configuration and instrumentation was upgraded again to 250 kW at steady state, and is awaiting the license of CNEN to operate definitely at this new power. This paper describes the experimental research project carried out in the IPR-R1 reactor that has as objective evaluate the behaviour of the reactor operational parameters, and mainly to investigate the influence of temperature on the neutronic variables. The research was supported by Research Support Foundation of the State of Minas Gerais (FAPEMIG) and Brazilian Council for Scientific and Technological Development (CNPq). The research project meets the recommendations of the IAEA, for safety, modernization and development of strategic plan for research reactors utilization. This work is in line with the strategic objectives of Brazil, which aims to design and construct the Brazilian Multipurpose research Reactor (RMB). (author)

  2. The Canadian research reactor spent fuel situation

    International Nuclear Information System (INIS)

    Ernst, P.C.

    1996-01-01

    This paper summarizes the present research reactor spent fuel situation in Canada. The research reactors currently operating are listed along with the types of fuel that they utilize. Other shut down research reactors contributing to the storage volume are included for completeness. The spent fuel storage facilities associated with these reactors and the methods used to determine criticality safety are described. Finally the current inventory of spent fuel and where it is stored is presented along with concerns for future storage. (author). 3 figs

  3. Reactor operations Brookhaven medical research reactor, Brookhaven high flux beam reactor informal monthly report

    International Nuclear Information System (INIS)

    Hauptman, H.M.; Petro, J.N.; Jacobi, O.

    1995-04-01

    This document is the April 1995 summary report on reactor operations at the Brookhaven Medical Research Reactor and the Brookhaven High Flux Beam Reactor. Ongoing experiments/irradiations in each are listed, and other significant operations functions are also noted. The HFBR surveillance testing schedule is also listed

  4. Annual report of department of research reactors, 2001. April 1, 2001 - March 31, 2002

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2002-12-01

    The Department of Research Reactor is responsible for the operation, maintenance, utilization of the JRR-3 and the JRR-4 and for the related R and D. Besides RI production including its R and D are carried out. This report describes the activities of the department in fiscal year of 2001 and it also includes some of the technical topics on the works mentioned above. As for the research reactors, we carried out the operation, maintenance, the utilization of irradiation and neutron beam experiments, technical management including fuels and water chemistry, radiation monitoring as related R and D works. RI production and its R and D works were conducted as well. The international cooperations between the developing countries and the department were also made concerning the operation, utilization and safety analysis for research reactors. (author)

  5. Ageing investigation and upgrading of components/systems of Kartini research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Syarip,; Setiawan, Widi [Yogyakarta Nuclear Research Centre, Yogyakarta (Indonesia)

    1998-10-01

    Kartini research reactor has been operated in good condition and has demonstrated successful operation for the past 18 years, utilized for: reactor kinetic and control studies, instrumentation tests, neutronic and thermohydraulic studies, routine neutron activation analysis, reactor safety studies, training for research reactor operators and supervisors, and reactor physics experiments. Several components of Kartini reactor use components from the abandoned IRT-2000 Project at Serpong and from Bandung Reactor Centre such as: reactor tank, reactor core, heat exchanger, motor blower for ventilation system, fuel elements, etc. To maintain a good operating performance and also for aging investigation purposes, the component failure data collection has been done. The method used is based on the Manual on Reliability Data Collection For Research Reactor PSAs, IAEA TECDOC 636, and analyzed by using Data Entry System (DES) computer code. Analysis result shows that the components/systems failure rate of Kartini reactor is around 1,5.10{sup -4} up to 2,8.10{sup -4} per hour, these values are within the ranges of the values indicated in IAEA TECDOC 478. Whereas from the analysis of irradiation history shows that the neutron fluence of fuel element with highest burn-up (2,05 gram U-235 in average) is around 1.04.10{sup 16} n Cm{sup -2} and this value is still far below its limiting value. Some reactor components/systems have been replaced and upgraded such as heat exchanger, instrumentation and control system (ICS), etc. The new reactor ICS was installed in 1994 which is designed as a distributed structure by using microprocessor based systems and bus system technology. The characteristic and operating performance of the new reactor ICS, as well as the operation history and improvement of the Kartini research reactor is presented. (J.P.N.)

  6. German research reactor back-end provisions

    International Nuclear Information System (INIS)

    Koester, Siegfried; Gruber, Gerhard

    2002-01-01

    Germany has several types of Research Reactors in operation. These reactors use fuel containing uranium of U.S. origin. Basically all the fuel which will be spent until May 2006 will be returned to the U.S. under existing contracts with the U.S. Department of Energy. The contracts are based on the U.S. FRR SNF (Foreign Research Reactor Spent Nuclear Fuel) Program which started in May 1996 and which will last for 10 years. In 1990, the German Federal Government started a program to long-term store (approx. 40 years) and finally dispose of spent fuel in Germany after the so-called U.S. fuel return window will be closed. In order to long-term store the fuel, a special container was designed which covers all different types of spent fuel from the Research Reactors. The container called 'CASTOR MTR 2' is basically licensed and is already in use for the spent fuel of Russian origin from the 'Research Reactor Rossendorf' in the eastern part of Germany. All that fuel is expected to be stored in the existing intermediate storage facility, the so-called BZA (Brennelemente Zwischenlager Ahaus). BZA already accomodates spent fuel from the former THTR-300 high temperature reactor. A final repository does not yet exist in Germany. Alternative provisions to close the back-end of the Research Reactor fuel cycle are reprocessing at COGEMA (France) or in Russian facilities, perspectively. Waste return in a form to be agreed will be mandatory, at least in France. (author)

  7. Education and Training on ISIS Research Reactor

    International Nuclear Information System (INIS)

    Foulon, F.; Badeau, G.; Lescop, B.; Wohleber, X.

    2013-01-01

    In the frame of academic and vocational programs the National Institute for Nuclear Science and Technology uses the ISIS research reactor as a major tool to ensure a practical and comprehensive understanding of the nuclear reactor physics, principles and operation. A large set of training courses have been developed on ISIS, optimising both the content of the courses and the pedagogical approach. Programs with duration ranging from 3 hours (introduction to reactor operation) to 24 hours (full program for the future operators of research reactors) are carried out on ISIS reactor. The reactor is operated about 350 hours/year for education and training, about 40 % of the courses being carried out in English. Thus, every year about 400 trainees attend training courses on ISIS reactor. We present here the ISIS research reactor and the practical courses that have been developed on ISIS reactor. Emphasis is given to the pedagogical method which is used to focus on the operational and safety aspects, both in normal and incidental operation. We will present the curricula of the academic and vocational courses in which the practical courses are integrated, the courses being targeted to a wide public, including operators of research reactors, engineers involved in the design and operation of nuclear reactors as well as staff of the regulatory body. We address the very positive impact of the courses on the development of the competences and skills of participants. Finally, we describe the Internet Reactor Laboratories (IRL) that are under development and will consist in broadcasting the training courses via internet to remote facilities or institutions

  8. Education and Training on ISIS Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Foulon, F.; Badeau, G.; Lescop, B.; Wohleber, X. [French Atomic Energy and Alternative Energies Commission, Paris (France)

    2013-07-01

    In the frame of academic and vocational programs the National Institute for Nuclear Science and Technology uses the ISIS research reactor as a major tool to ensure a practical and comprehensive understanding of the nuclear reactor physics, principles and operation. A large set of training courses have been developed on ISIS, optimising both the content of the courses and the pedagogical approach. Programs with duration ranging from 3 hours (introduction to reactor operation) to 24 hours (full program for the future operators of research reactors) are carried out on ISIS reactor. The reactor is operated about 350 hours/year for education and training, about 40 % of the courses being carried out in English. Thus, every year about 400 trainees attend training courses on ISIS reactor. We present here the ISIS research reactor and the practical courses that have been developed on ISIS reactor. Emphasis is given to the pedagogical method which is used to focus on the operational and safety aspects, both in normal and incidental operation. We will present the curricula of the academic and vocational courses in which the practical courses are integrated, the courses being targeted to a wide public, including operators of research reactors, engineers involved in the design and operation of nuclear reactors as well as staff of the regulatory body. We address the very positive impact of the courses on the development of the competences and skills of participants. Finally, we describe the Internet Reactor Laboratories (IRL) that are under development and will consist in broadcasting the training courses via internet to remote facilities or institutions.

  9. Twenty years of health physics research reactor operation

    International Nuclear Information System (INIS)

    Sims, C.S.; Gilley, L.W.

    1983-01-01

    The Health Physics Research Reactor at the Oak Ridge National Laboratory has been in regular use for more than two decades. Safe operation of this fast reactor over this extended period indicates that (1) fundamental design, (2) operational procedures, (3) operator training and performance, (4) maintenance activites, and (5) management have all been eminently satisfactory. The reactor and its uses are described, the operational history and significant events are reviewed, and operational improvements and maintenance are discussed

  10. The first university research reactor in India

    International Nuclear Information System (INIS)

    Murty, G.S.

    1999-01-01

    As the first university research reactor in India, the low power, pool type with fixed core and low enriched uranium fuel research reactor is under construction in the Andhra university campus, Andhra Pradesh, India. The reactor is expected to be commissioned during 2001-2002. The mission of the reactor is to play the research center as a regional research facility catering to the needs of academic institutions and industrial organizations of this region of the country. Further, to encourage interdisplinary and multidisplinary research activities, to supply radioisotope and labelled compounds to the user institutions and to create awareness towards the peaceful uses of atomic energy. This report describes its objectives, status and future plans in brief. (H. Itami)

  11. The first university research reactor in India

    Energy Technology Data Exchange (ETDEWEB)

    Murty, G.S. [Co-ordinator, Low Power Research Reactor, Andhra Univ., Visakapatnam (India)

    1999-08-01

    As the first university research reactor in India, the low power, pool type with fixed core and low enriched uranium fuel research reactor is under construction in the Andhra university campus, Andhra Pradesh, India. The reactor is expected to be commissioned during 2001-2002. The mission of the reactor is to play the research center as a regional research facility catering to the needs of academic institutions and industrial organizations of this region of the country. Further, to encourage interdisplinary and multidisplinary research activities, to supply radioisotope and labelled compounds to the user institutions and to create awareness towards the peaceful uses of atomic energy. This report describes its objectives, status and future plans in brief. (H. Itami)

  12. Proceedings of the FNCA 2006 workshop on the utilization of research reactors (Contract research)

    International Nuclear Information System (INIS)

    2007-09-01

    The FNCA 2006 Workshop on the Utilization of Research Reactors was held in Manila, Philippines from August 28 to September 1, 2006. This workshop was executed based on the agreement in the seventh Coordinator's Meeting of Forum for Nuclear Cooperation in Asia (FNCA) held in Tokyo, Mach 2006. The workshop consisted of three groups under the themes of the following fields; 1) Neutron Activation Analysis, 2) Research Reactor Technology and 3) Tc-99m Generator Technology. The total number of participants for the workshop was 45 people from 9 countries; China, Indonesia, Korea, Malaysia, the Philippines, Thailand, Vietnam, Bangladesh and Japan. This report consists of 10 papers for Neutron Activation Analysis, 7 papers for Tc-99m Generator Technology, 9 papers for Research Reactor Technology and a summary report. (author)

  13. Proton induced nuclide production cross section by HETC-3STEP/FRG-R

    Energy Technology Data Exchange (ETDEWEB)

    Shigyo, Nobuhiro; Ishibashi, Kenji [Kyushu Univ., Fukuoka (Japan); Yoshizawa, Nobuaki; Takada, Hiroshi

    1998-03-01

    High Energy Transport Code (HETC) based on the intranuclear-cascade-evaporation model is modified to calculate the fragmentation cross section. The exciton model is adopted for improvement of backward nucleon-emission cross section for low-energy nucleon-incident events. The level density parameter depending on the excitation energy is taken in the evaporation process. The fragmentation reaction is incorporated into HETC as a subroutine set by the use of the systematics of the reaction. The modified HETC (HETC-3STEP/FRG-R) reproduces experimental fragment yields to a reasonable degree. (author)

  14. GKSS. Annual report 2007/2008. Science use

    International Nuclear Information System (INIS)

    2008-06-01

    The following topics are dealt with: Titanium aluminide alloys, microscopy in materials research, a new generation of membranes based on nanostructured polymers, the GKSS Engineering Materials Science Synchrotron, regenerativ medicine, the colour of the coastal sea, news from the North German Climate Office, the climate change for the Baltic sea basin, polyfluorinated compounds as pollutant in the coastal environment, the FRG 1 research reactor. (HSI)

  15. GKSS. Annual report 2007/2008. Science use; GKSS. Jahresbericht 2007/2008. Wissenschaft nutzen

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2008-06-15

    The following topics are dealt with: Titanium aluminide alloys, microscopy in materials research, a new generation of membranes based on nanostructured polymers, the GKSS Engineering Materials Science Synchrotron, regenerativ medicine, the colour of the coastal sea, news from the North German Climate Office, the climate change for the Baltic sea basin, polyfluorinated compounds as pollutant in the coastal environment, the FRG 1 research reactor. (HSI)

  16. Proceedings of the FNCA 2004 workshop on the utilization of research reactors (Contract research)

    International Nuclear Information System (INIS)

    2006-07-01

    The FNCA 2004 Workshop on the Utilization of Research Reactors, which is the twelfth workshop on the theme of research reactor utilization, was held in Bangkok, Thailand from January 13 to 21, 2005. This workshop was executed based on the agreement in the fifth Coordinator's Meeting of Forum for Nuclear Cooperation in Asia (FNCA) held in Tokyo, March 2004. The workshop consisted of three groups under the themes of the following fields; 1) Neutron Activation Analysis, 2) Research Reactor Technology and 3) Tc-99m Generator Technology. The total number of participants for the workshop was 59 people from 8 countries; China, Indonesia, Korea, Malaysia, the Philippines, Thailand, Vietnam and Japan. This report consists of 6 papers for Neutron Activation Analysis, 5 papers for Research Reactor Technology, 5 Papers for Tc-99m Generator Technology and a summary report. The 15 of the presented papers are indexed individually. (J.P.N.)

  17. Cost aspects of the research reactor fuel cycle

    International Nuclear Information System (INIS)

    2010-01-01

    Research reactors have made valuable contributions to the development of nuclear power, basic science, materials development, radioisotope production for medicine and industry, and education and training. In doing so, they have provided an invaluable service to humanity. Research reactors are expected to make important contributions in the coming decades to further development of the peaceful uses of nuclear technology, in particular for advanced nuclear fission reactors and fuel cycles, fusion, high energy physics, basic research, materials science, nuclear medicine, and biological sciences. However, in the context of decreased public sector support, research reactors are increasingly faced with financial constraints. It is therefore of great importance that their operations are based on a sound understanding of the costs of the complete research reactor fuel cycle, and that they are managed according to sound financial and economic principles. This publication is targeted at individuals and organizations involved with research reactor operations, with the aim of providing both information and an analytical framework for assessing and determining the cost structure of fuel cycle related activities. Efficient management of fuel cycle expenditures is an important component in developing strategies for sustainable future operation of a research reactor. The elements of the fuel cycle are presented with a description of how they can affect the cost efficient operation of a research reactor. A systematic review of fuel cycle choices is particularly important when a new reactor is being planned or when an existing reactor is facing major changes in its fuel cycle structure, for example because of conversion of the core from high enriched uranium (HEU) to low enriched uranium (LEU) fuel, or the changes in spent fuel management provision. Review and optimization of fuel cycle issues is also recommended for existing research reactors, even in cases where research reactor

  18. Developments in the regulation of research reactors

    International Nuclear Information System (INIS)

    Loy, J.

    2003-01-01

    The International Atomic Energy Agency (IAEA) has data on over 670 research reactors in the world. Fewer than half of them are operational and a significant number are in a shutdown but not decommissioned state. The International Nuclear Safety Advisory Group (INSAG) has expressed concerns about the safety of many research reactors and this has resulted in a process to draw up an international Code of Conduct on the Safety of Research Reactors. The IAEA is also reviewing its safety standards applying to research reactors. On the home front, regulation of the construction of the Replacement Research Reactor continues. During the construction phase, regulation has centred around the consideration of Requests for Approval (RFA) for the manufacture and installation of systems, structures and components important for safety. Quality control of construction of systems, structures and components is the central issue. The process for regulation of commissioning is under consideration

  19. Quality assurance in the manufacture of metallic uranium fuel for research reactors

    International Nuclear Information System (INIS)

    Shah, B.K.; Kumar, Arbind; Nanekar, P.P.; Vaidya, P.R.

    2009-01-01

    Two Research Reactors viz. CIRUS and DHRUVA are operating at Trombay since 1960 and 1985 respectively. Cirus is a 40 MWth reactor using heavy water as moderator and light water as coolant. Dhruva is a 100 MWth reactor using heavy water as moderator and coolant. The maximum neutron flux of these reactors are 6.7 x 10 13 n/cm 2 /s (Cirus) and 1.8 x 10 14 n/cm 2 /s (Dhruva). Both these reactors are used for basic research, R and D in reactor technology, isotope production and operator training. Fuel material for these reactors is natural uranium metallic rods claded in finned aluminium (99.5%) tubes. This presentation will discuss various issues related to fabrication quality assurance and reactor behavior of metallic uranium fuel used in research reactors

  20. Research reactor collaboration in the Asia-Pacific region

    International Nuclear Information System (INIS)

    Jun, Byung Jin

    2006-01-01

    The number of research reactors over the world has been decreasing since its peak in the middle of the 1970s, and it is predicted to decrease more rapidly than before in the future. International collaboration on research reactors is an effective way for their continued safe service to human welfare in various technical areas. The number of new research reactors under construction or planned for in the Asia-Pacific region is the greatest in the world. Among the regional collaboration activities on research reactors, safety has been the most important subject followed by neutron activation analysis, radioisotope production and neutron beam applications. It is understood that more regional collaboration on basic technologies important for the safety, management and utilization of the research reactors is demanding. The new project proposal of the Forum for Nuclear Cooperation in Asia on 'Research Reactor Technology for Effective Utilization' is understood to meet the demands. Meanwhile, there is a consensus on the need for research reactor resource sharing in the region. As a result of the review on the international collaboration activities in the region, the author suggests a linkage between the above new project and IAEA/RCA project considering a possible sharing of research reactor resources in the region. (author)

  1. Sodium fast reactor safety and licensing research plan - Volume II

    International Nuclear Information System (INIS)

    Ludewig, H.; Powers, D.A.; Hewson, John C.; LaChance, Jeffrey L.; Wright, A.; Phillips, J.; Zeyen, R.; Clement, B.; Garner, Frank; Walters, Leon; Wright, Steve; Ott, Larry J.; Suo-Anttila, Ahti Jorma; Denning, Richard; Ohshima, Hiroyuki; Ohno, S.; Miyhara, S.; Yacout, Abdellatif; Farmer, M.; Wade, D.; Grandy, C.; Schmidt, R.; Cahalen, J.; Olivier, Tara Jean; Budnitz, R.; Tobita, Yoshiharu; Serre, Frederic; Natesan, Ken; Carbajo, Juan J.; Jeong, Hae-Yong; Wigeland, Roald; Corradini, Michael; Thomas, Justin; Wei, Tom; Sofu, Tanju; Flanagan, George F.; Bari, R.; Porter D.

    2012-01-01

    Expert panels comprised of subject matter experts identified at the U.S. National Laboratories (SNL, ANL, INL, ORNL, LBL, and BNL), universities (University of Wisconsin and Ohio State University), international agencies (IRSN, CEA, JAEA, KAERI, and JRC-IE) and private consultation companies (Radiation Effects Consulting) were assembled to perform a gap analysis for sodium fast reactor licensing. Expert-opinion elicitation was performed to qualitatively assess the current state of sodium fast reactor technologies. Five independent gap analyses were performed resulting in the following topical reports: (1) Accident Initiators and Sequences (i.e., Initiators/Sequences Technology Gap Analysis), (2) Sodium Technology Phenomena (i.e., Advanced Burner Reactor Sodium Technology Gap Analysis), (3) Fuels and Materials (i.e., Sodium Fast Reactor Fuels and Materials: Research Needs), (4) Source Term Characterization (i.e., Advanced Sodium Fast Reactor Accident Source Terms: Research Needs), and (5) Computer Codes and Models (i.e., Sodium Fast Reactor Gaps Analysis of Computer Codes and Models for Accident Analysis and Reactor Safety). Volume II of the Sodium Research Plan consolidates the five gap analysis reports produced by each expert panel, wherein the importance of the identified phenomena and necessities of further experimental research and code development were addressed. The findings from these five reports comprised the basis for the analysis in Sodium Fast Reactor Research Plan Volume I.

  2. Sodium fast reactor safety and licensing research plan. Volume II.

    Energy Technology Data Exchange (ETDEWEB)

    Ludewig, H. (Brokhaven National Laboratory, Upton, NY); Powers, D. A.; Hewson, John C.; LaChance, Jeffrey L.; Wright, A. (Argonne National Laboratory, Argonne, IL); Phillips, J.; Zeyen, R. (Institute for Energy Petten, Saint-Paul-lez-Durance, France); Clement, B. (IRSN/DPAM.SEMIC Bt 702, Saint-Paul-lez-Durance, France); Garner, Frank (Radiation Effects Consulting, Richland, WA); Walters, Leon (Advanced Reactor Concepts, Los Alamos, NM); Wright, Steve; Ott, Larry J. (Oak Ridge National Laboratory, Oak Ridge, TN); Suo-Anttila, Ahti Jorma; Denning, Richard (Ohio State University, Columbus, OH); Ohshima, Hiroyuki (Japan Atomic Energy Agency, Ibaraki, Japan); Ohno, S. (Japan Atomic Energy Agency, Ibaraki, Japan); Miyhara, S. (Japan Atomic Energy Agency, Ibaraki, Japan); Yacout, Abdellatif (Argonne National Laboratory, Argonne, IL); Farmer, M. (Argonne National Laboratory, Argonne, IL); Wade, D. (Argonne National Laboratory, Argonne, IL); Grandy, C. (Argonne National Laboratory, Argonne, IL); Schmidt, R.; Cahalen, J. (Argonne National Laboratory, Argonne, IL); Olivier, Tara Jean; Budnitz, R. (Lawrence Berkeley National Laboratory, Berkeley, CA); Tobita, Yoshiharu (Japan Atomic Energy Agency, Ibaraki, Japan); Serre, Frederic (Centre d' %C3%94etudes nucl%C3%94eaires de Cadarache, Cea, France); Natesan, Ken (Argonne National Laboratory, Argonne, IL); Carbajo, Juan J. (Oak Ridge National Laboratory, Oak Ridge, TN); Jeong, Hae-Yong (Korea Atomic Energy Research Institute, Daejeon, Korea); Wigeland, Roald (Idaho National Laboratory, Idaho Falls, ID); Corradini, Michael (University of Wisconsin-Madison, Madison, WI); Thomas, Justin (Argonne National Laboratory, Argonne, IL); Wei, Tom (Argonne National Laboratory, Argonne, IL); Sofu, Tanju (Argonne National Laboratory, Argonne, IL); Flanagan, George F. (Oak Ridge National Laboratory, Oak Ridge, TN); Bari, R. (Brokhaven National Laboratory, Upton, NY); Porter D. (Idaho National Laboratory, Idaho Falls, ID); Lambert, J. (Argonne National Laboratory, Argonne, IL); Hayes, S. (Idaho National Laboratory, Idaho Falls, ID); Sackett, J. (Idaho National Laboratory, Idaho Falls, ID); Denman, Matthew R.

    2012-05-01

    Expert panels comprised of subject matter experts identified at the U.S. National Laboratories (SNL, ANL, INL, ORNL, LBL, and BNL), universities (University of Wisconsin and Ohio State University), international agencies (IRSN, CEA, JAEA, KAERI, and JRC-IE) and private consultation companies (Radiation Effects Consulting) were assembled to perform a gap analysis for sodium fast reactor licensing. Expert-opinion elicitation was performed to qualitatively assess the current state of sodium fast reactor technologies. Five independent gap analyses were performed resulting in the following topical reports: (1) Accident Initiators and Sequences (i.e., Initiators/Sequences Technology Gap Analysis), (2) Sodium Technology Phenomena (i.e., Advanced Burner Reactor Sodium Technology Gap Analysis), (3) Fuels and Materials (i.e., Sodium Fast Reactor Fuels and Materials: Research Needs), (4) Source Term Characterization (i.e., Advanced Sodium Fast Reactor Accident Source Terms: Research Needs), and (5) Computer Codes and Models (i.e., Sodium Fast Reactor Gaps Analysis of Computer Codes and Models for Accident Analysis and Reactor Safety). Volume II of the Sodium Research Plan consolidates the five gap analysis reports produced by each expert panel, wherein the importance of the identified phenomena and necessities of further experimental research and code development were addressed. The findings from these five reports comprised the basis for the analysis in Sodium Fast Reactor Research Plan Volume I.

  3. History, Development and Future of TRIGA Research Reactors

    International Nuclear Information System (INIS)

    2016-01-01

    Due to its particular fuel design and resulting enhanced inherent safety features, TRIGA reactors (Training, Research, Isotopes, General Atomics) constitute a ‘class of their own’ among the large variety of research reactors built world-wide. This publication summarizes in a single document the information on the past and present of TRIGA research reactors and presents an outlook in view of potential issues to be solved by TRIGA operating organizations in the near future. It covers the historical development and basic TRIGA characteristics, followed by utilization, fuel conversion and ageing management of TRIGA research reactors. It continues with issues and challenges, introduction to the global TRIGA research reactor network and concludes with future perspectives. The publication is complemented with a CD-ROM to illustrate the historical developments of TRIGA research reactors through individual facility examples and experiences

  4. Current utilization and long term strategy of the Finnish TRIGA research reactor FiR 1

    International Nuclear Information System (INIS)

    Auterinen, Iiro; Salmenhaara, Seppo

    2008-01-01

    FiR 1 (TRIGA Mark II, 250 kW) has an important international role in the development of boron neutron capture therapy (BNCT) for cancer. The safety and efficacy of BNCT is studied for several different cancers: - primary glioblastoma, a highly malignant brain tumour (since 1999); - recurrent glioblastoma or anaplastic astrocytoma (since 2001); - recurrent inoperable head and neck carcinoma (since 2003). It is one of the few facilities in the world providing this kind of treatments. The successes in the BNCT development have now created a demand for these treatments, although they are given on an experimental basis. Well over 100 patients treated now since May 1999: - at least 1 patient irradiation / week, often 2 (Tuesday and Thursday) - patients are referred to BNCT-treatments from several hospitals, also outside research protocols; - the hospitals pay for the treatment. The FiR 1 reactor has proven to be a reliable neutron source for the BNCT treatments; no patient irradiations have been cancelled because of a failure of the reactor. The BNCT facility has become a center of extensive academic research especially in medical physics. Nuclear education and training continue to play also a role at FiR 1 in the form of university courses and training of nuclear industry personnel. FiR 1 is one of the two sources in Scandinavia for short lived radioisotopes used in tracer studies in industry. The main isotope produced is Br-82 in the form of either KBr or ethylene bromide. Other typical isotopes are Na-24, Ar-41, La-140. The isotopes are used mainly in tracer studies in industry (Indmeas Inc., Finland). Typical activity of one irradiated Br-sample is 20 - 80 GBq; total activity produced in one year is over 3 TBq; the reactor operating time needed for the isotope production is one or two days per week. Accelerator based neutron sources are developed for BNCT. The prospect is that when BNCT will achieve a status of a fully accepted and efficient treatment modality for

  5. Reactor operations at SAFARI-1

    International Nuclear Information System (INIS)

    Vlok, J.W.H.

    2003-01-01

    A vigorous commercial programme of isotope production and other radiation services has been followed by the SAFARI-1 research reactor over the past ten years - superimposed on the original purpose of the reactor to provide a basic tool for nuclear research, development and education to the country at an institutional level. A combination of the binding nature of the resulting contractual obligations and tighter regulatory control has demanded an equally vigorous programme of upgrading, replacement and renovation of many systems in order to improve the safety and reliability of the reactor. Not least among these changes is the more effective training and deployment of operations personnel that has been necessitated as the operational demands on the reactor evolved from five days per week to twenty four hours per day, seven days per week, with more than 300 days per year at full power. This paper briefly sketches the operational history of SAFARI-1 and then focuses on the training and structuring currently in place to meet the operational needs. There is a detailed step-by-step look at the operator?s career plan and pre-defined milestones. Shift work, especially the shift cycle, has a negative influence on the operator's career path development, especially due to his unavailability for training. Methods utilised to minimise this influence are presented. The increase of responsibilities regarding the operation of the reactor, ancillaries and experimental facilities as the operator progresses with his career are discussed. (author)

  6. Research reactors: a tool for science and medicine

    International Nuclear Information System (INIS)

    Ordonez, Juan

    2001-01-01

    The types and uses of research reactors are reviewed. After an analysis of the world situation, the demand of new research reactors of about 20 MW is foreseen. The experience and competitiveness of INVAP S.E. as designer and constructor of research reactors is outlined and the general specifications of the reactors designed by INVAP for Egypt and Australia are given

  7. Reactor safety research. The CEC contribution

    International Nuclear Information System (INIS)

    Krischer, W.

    1990-01-01

    The involvement of the EC Commission in the reactor safety research dates back almost to the implementation of the EURATOM Treaty and has thus lasted for thirty years. The need for close collaboration and for general consensus on some crucial problems of concern to the public, has made the role of international organizations and, as far as Europe is concerned, the role of the European Community particularly important. The areas in which the CEC has been active during the last five years are widespread. This is partly due to the fact that, after TMI and Chernobyl, the effort and the interest of the different countries in reactor safety was considerable. Reactor Safety Research represents the proceedings of a seminar held by the Commission at the end of its research programme 1984-88 on reactor safety. As such it gives a comprehensive overview of the recent activities and main results achieved in the CEC Joint Research Centre and in national laboratories throughout Europe on the basis of shared cost actions. In a concluding chapter the book reports on the opinions, expressed during a panel by a group of major exponents, on the needs for future research. The main topics addressed are, with particular reference to Light Water Reactors (LWRS): reliability and risk evaluation, inspection of steel components, primary circuit components end-of-life prediction, and abnormal behaviour of reactor cooling systems. As far as LMFBRs are concerned, the topics covered are: severe accident modelling, material properties and structural behaviour studies. There are 67 pages, all of which are indexed separately. Reactor Safety Research will be of particular interest to reliability and safety engineers, nuclear engineers and technicians, and mechanical and structural engineers. (author)

  8. The Jules Horowitz Reactor project, a driver for revival of the research reactor community

    International Nuclear Information System (INIS)

    Pere, P.; Cavailler, C.; Pascal, C.

    2010-01-01

    The first concrete of the nuclear island for the Jules Horowitz Reactor (JHR) was poured at the end of July 2009 and construction is ongoing. The JHR is the largest new platform for irradiation experiments supporting Generation II and III reactors, Generation IV technologies, and radioisotope production. This facility, composed of a unique grouping of workshops, hot cells and hot laboratories together with a first-rate MTR research reactor, will ensure that the process, from preparations for irradiation experiments through post-irradiation non-destructive examination, is completed expediently, efficiently and, of course, safely. In addition to the performance requirements to be met in terms of neutron fluxes on the samples (5x10 14 n.cm -2 /sec -1 E>1 MeV in core and 3,6x10 14 n.cm -2 /sec -1 E<0.625 eV in the reflector) and the JHR's considerable irradiation capabilities (more than 20 experiments and one-tenth of irradiation area for simultaneous radioisotope production), the JHR is the first MTR to be built since the end of the 1960s, making this an especially challenging project. The presentation will provide an overview of the reactor, hot cells and laboratories and an outline of the key milestones in the project schedule, including initial criticality in early 2014 and radioisotope production in 2015. This will be followed by a description of the project organization set up by the CEA as owner and future operator and AREVA TA as prime contractor and supplier of critical systems, and a discussion of project challenges, especially those dealing with the following items:accommodation of a broad experimental domain; involvement by international partners making in-kind contributions to the project; ? development of components critical to safety and performance; the revival of engineering of research reactors and experimental devices involving France's historical players in the field of research reactors, and; tools to carry out the project, including computer codes

  9. Improvement of colloid sampling techniques in groundwater and actinide characterisation of the groundwater systems at Gorleben (FRG) and El Berrocal (E)

    International Nuclear Information System (INIS)

    Dearlove, J.P.L.; Longworth, G.; Ivanovich, M.

    1990-08-01

    Two sites, the Gorleben site (FRG) and the El Berrocal Experimental station (E) have been studied to evaluate different sampling and analytical techniques for the characterisation of particulates (> 1000 nm size), colloids (1-1000 nm size) and the solution phase (<1 nm) in groundwaters in terms of their physical, chemical and actinide composition. The uptake characteristics of the field ultrafiltration system used to separate the colloid fraction from the solution phase in the groundwater have also been studied. (Author)

  10. Annual report of department of research reactor, 2000. April 1, 2000 - March 31, 2001

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2002-02-01

    The Department of Research Reactor is responsible for the operation, Maintenance, utilization of the JRR-3 and the JRR-4 and for the related R and D. Besides RI production including its R and D are carried out. This report describes the activities of the department in fiscal year of 2000 and it also includes some of the technical topics on the works mentioned above. As for the research reactors, we carried out the operation, maintenance, the utilization of irradiation and neutron beam experiments, technical management including fuels and water chemistry, radiation monitoring as related R and D works. RI Production and its R and D works were conducted as well. The international cooperations between the developing countries and the department were also made concerning the operation, utilization and safety analysis for research reactors. Although the term 'JRR-3M' was used to denote the JRR-3M modified 1990 until the 2000 annual report of the Department of Research Reactor, the term 'JRR-3' will be used from this annual report because the JRR-3 has been operated for about 10 years since the modification and is now under further modification and upgrading study. (author)

  11. Neutron Flux Variation in the Nigeria Research Reactor-1 (NIRR-1)

    International Nuclear Information System (INIS)

    Yahaya, M.; Ahmed, Y.A.

    2013-01-01

    In order to ascertain the level of flux variation in one of the inner irradiation channels of the Nigeria Research Reactor-1 (NIRR-1), the irradiation container used for routine activation analysis was employed with copper wires as flux monitors. Measurements were carried out with copper wires arranged in axial direction to determine the thermal neutron flux at selected positions using absolute foil activation method. Our results show that there exists a slight flux variation from one position to another ranging from (4.57±0.21) x 10 11 to (5.20± 0.20) x 10 11 cm -2 s -1 .Individual foil shows slight flux variation from one position to another in the same irradiation container but they all pointed toward a level of consistency in variation in spite of the recent installation of the cadmium lined irradiation channel. The values obtained in this work are in good agreement with the previously measured value of (5.14±0.24) x 10 11 cm -2 s -1 after commissioning of NIRR-1 (Jonah et al., 2005). This shows that the cadmium lined installation does not affect the flux stability. In order to improve the accuracy of NAA using NIRR-l facility, there is need for flux corrections to be made by MNSR users during NAA particularly for samples in the axial position for long irradiation.

  12. The AFR. An approved network of research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hampel, Gabriele [Mainz Univ. (Germany). Arbeitsgemeinschaft fuer Betriebs- und Sicherheitsfragen an Forschungsreaktoren (AFR)

    2012-10-15

    AFR (Arbeitsgemeinschaft fuer Betriebs- und Sicherheitsfragen an Forschungsreaktoren) is the German acronym for 'Association for Research Reactor Operation and Safety Issues' which was founded in 1959. Reactor managers of European research reactors mainly from the German linguistic area meet regularly for their mutual benefit to exchange experience and knowledge in all areas of operating, managing and utilization of research reactors. In the last 2 years joint meetings were held together with the French association of research reactors CER (Club d'Exploitants des Reacteurs). In this contribution the AFR, its members, work and aims as well as the French partner CER are presented. (orig.)

  13. Annual report of Department of Research Reactor, 2003. April 1, 2003 - March 31, 2004

    International Nuclear Information System (INIS)

    2005-02-01

    The Department of Research Reactor is responsible for the operation, maintenance, utilization of the JRR-3 and the JRR-4 and for the related R and D. Besides the RI production and its R and D are carried out. This report describes the activities of the department in fiscal year of 2003 and also includes some of the technical topics on the works mentioned above. As for the research reactors, we carried out the operation, maintenance, utilization of irradiation and neutron beam experiments, technical management including management of fuels and water chemistry, and related R and D works. The RI production and its R and D works were conducted as well. The international co-operations between the developing countries and the department were also made concerning the operation, utilization and safety analysis for research reactors. (author)

  14. Safety of Research Reactors. Safety Requirements

    International Nuclear Information System (INIS)

    2010-01-01

    The main objective of this Safety Requirements publication is to provide a basis for safety and a basis for safety assessment for all stages in the lifetime of a research reactor. Another objective is to establish requirements on aspects relating to regulatory control, the management of safety, site evaluation, design, operation and decommissioning. Technical and administrative requirements for the safety of research reactors are established in accordance with these objectives. This Safety Requirements publication is intended for use by organizations engaged in the site evaluation, design, manufacturing, construction, operation and decommissioning of research reactors as well as by regulatory bodies

  15. IRT-type research reactor physical calculation methodology

    International Nuclear Information System (INIS)

    Carrera, W.; Castaneda, S.; Garcia, F.; Garcia, L.; Reyes, O.

    1990-01-01

    In the present paper an established physical calculation procedure for the research reactor of the Nuclear Research Center (CIN) is described. The results obtained by the method are compared with the ones reported during the physical start up of a reactor with similar characteristics to the CIN reactor. 11 refs

  16. RA Research reactor Annual report 1981 - Part 1, Operation, maintenance and utilization of the RA reactor

    International Nuclear Information System (INIS)

    Sotic, O.; Milosevic, M.; Martinc, R.; Kozomara-Maic, S.; Cupac, S.; Radivojevic, J.; Stamenkovic, D.; Skoric, M.

    1981-12-01

    The RA nuclear reactor stopped operation after March 1979 campaign due to appearance of aluminium oxyhydrates deposits on the surface of fuel element claddings. Relevant decisions of the Sanitary inspection body of the Ministry of health and the Director General of the 'Boris Kidric' Institute of nuclear sciences, Vinca, banned further reactor operation until reasons caused aluminium oxyhydrates deposition are investigated and removed to enable regular reactor operation. Until the end of 1979 and during 1980, after a series of analyses and findings that caused cease of reactor operation, all the preparatory actions needed for restart were performed. Due to the fact that there is no emergency cooling system and no appropriate filtering system at the reactor, and according to the new regulations about start up of nuclear facilities, the Sanitary inspection body made a decision about temporary licence for reactor start-up meaning performance of the 'zero experiment' limiting the operating power to 1% of the nominal power. Accordingly the reactor was restarted on January 21 1981. Criticality was reached with the core made of 80% enriched fuel elements only. After the experiment was finished by the end of March a permission was demanded for operation at higher power levels at full power. Taking into account the state of the reactor components the operating licence was issued limiting the power to 2 MW until reconstruction of the ventilation system and construction of the emergency cooling system are fulfilled. Program of testing operation started on September 15 1981 increasing gradually the operating power. Thus the reactor was operated at 2 MW power for 15 days during November and December. The total production achieved in 1981 was 1698 MWh. This enabled isotopes production at the reactor during last two months. Control and maintenance of the reactor components and systems was done regularly and efficiently within limits imposed by availability of spare parts. The

  17. RMB: the new Brazilian Multipurpose Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Perrotta, Jose Augusto, E-mail: perrotta@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2016-07-01

    Full text: The Brazilian research reactors have a limited capacity for radioisotopes production, leading to a high dependence on external supply for radioisotopes used in nuclear medicine. In order to overcome this condition and due to the old age of these research reactors, the Brazilian Nuclear Energy Commission decided, in 2008, to construct a new research reactor. The new reactor named RMB (Brazilian Multipurpose Reactor) will be part of a new nuclear research center, to be built on a site about 100 kilometers from São Paulo city, in the southern part of Brazil. The new nuclear research center will have a 30 MW open pool type research reactor using low enriched uranium fuel, and several associated laboratories in order to produce radioisotopes for medical and industrial use, to use neutron beams in scientific and technological research; to perform neutron activation analysis; and to perform materials and fuels irradiation tests. Regarding the neutron beams use, the RMB design provides thermal and cold neutron beams. From one side of the reactor, the neutron guides will extend to an experimental hall of instruments named Neutron Guide Hall where it will be installed the scattering instruments. In the initial stage of the reactor operation, the intent is to implement two neutron guides for thermal neutrons and another two for cold neutrons. The 2015 SBPMAT symposium has presented the technical overview of the RMB project and its main buildings, structures and components. At this year symposium, the RMB presentation updates some technical information and the development status of the project, discussing the negative results of the Brazilian political and economic crisis to the project development and its future perspectives. (author)

  18. Present status of research reactor and future prospects

    International Nuclear Information System (INIS)

    Nakajima, Ken

    2013-01-01

    Research reactors have been playing an important role in the research and development of the various fields, such as physics, chemistry, biology, engineering, agriculture, medicine, etc. as well as human resource development. However, the most of them are older than 40 years, and the ageing management is an important issue. In Japan, only two research reactors are operational after the Great East Japan Earthquake in 2011. JAEA's reactors suffered from the quake and they are under inspections. Kyoto University Research Reactor, one of the operational reactors, has been widely used for research and human resource development, and the additional safety measures against the station blackout were installed. Besides the affect of the quake, the disposal or treatment of spent fuel becomes an inevitable problem for research reactors. The way of spent fuel disposal or treatment should be determined with the nation-wide and/or international coalition. (author)

  19. TRIGA 14 MW Research Reactor Status and Utilization

    International Nuclear Information System (INIS)

    Barbos, D.; Ciocanescu, M.; Paunoiu, C.

    2016-01-01

    Institute for Nuclear Research is the owner of the largest family TRIGA research reactor, TRIGA14 MW research reactor. TRIGA14 MW reactor was designed to be operated with HEU nuclear fuel but now the reactor core was fully converted to LEU nuclear fuel. The full conversion of the core was a necessary step to ensure the continuous operation of the reactor. The core conversion took place gradually, using fuel manufactured in different batches by two qualified suppliers based on the same well qualified technology for TRIGA fuel, including some variability which might lead to a peculiar behaviour under specific conditions of reactor utilization. After the completion of the conversion a modernization program for the reactor systems was initiated in order to achieve two main objectives: safe operation of the reactor and reactor utilization in a competitive environment to satisfy the current and future demands and requirements. The 14 MW TRIGA research reactor operated by the Institute for Nuclear Research in Pitesti, Romania, is a relatively new reactor, commissioned 37 years ago. It is expected to operate for another 15-20 years, sustaining new fuel and testing of materials for future generations of power reactors, supporting radioisotopes production through the development of more efficient new technologies, sustaining research or enhanced safety, extended burn up and verification of new developments concerning nuclear power plants life extension, to sustain neutron application in physics research, thus becoming a centre for instruction and training in the near future. A main objective of the TRIGA14MW research reactor is the testing of nuclear fuel and nuclear material. The TRIGA 14 MW reactor is used for medical and industrial radioisotopes production ( 131 I, 125 I, 192 Ir etc.) and a method for 99 Mo- 99 Tc production from fission is under development. For nuclear materials properties investigation, neutron radiography methods have been developed in the INR. The

  20. Application of research reactors for radiation education

    International Nuclear Information System (INIS)

    Ito, Yasuo; Harasawa, Susumu; Hayashi, Shu A.; Tomura, Kenji; Matsuura, Tatsuo; Nakanishi, Tomoko M.; Yamamoto, Yusuke

    1999-01-01

    Nuclear research Reactors are, as well as being necessary for research purposes, indispensable educational tools for a country whose electric power resources are strongly dependent on nuclear energy. Both large and small research reactors are available, but small ones are highly useful from the viewpoint of radiation education. This paper oders a brief review of how small research reactors can, and must, be used for radiation education for high school students, college and graduate students, as well as for the public. (author)

  1. Application of research reactors for radiation education

    Energy Technology Data Exchange (ETDEWEB)

    Ito, Yasuo [Tokyo Univ. (Japan). Research Center for Nuclear Science and Technology; Harasawa, Susumu; Hayashi, Shu A.; Tomura, Kenji; Matsuura, Tatsuo; Nakanishi, Tomoko M.; Yamamoto, Yusuke

    1999-09-01

    Nuclear research Reactors are, as well as being necessary for research purposes, indispensable educational tools for a country whose electric power resources are strongly dependent on nuclear energy. Both large and small research reactors are available, but small ones are highly useful from the viewpoint of radiation education. This paper oders a brief review of how small research reactors can, and must, be used for radiation education for high school students, college and graduate students, as well as for the public. (author)

  2. Safety requirements in the design of research reactors: A Canadian perspective

    International Nuclear Information System (INIS)

    Lee, A.G.; Langman, V.J.

    2000-01-01

    In Canada, the formal development of safety requirements for the design of research reactors in general began under an inter-organizational Small Reactor Criteria Committee. This committee developed safety and licensing criteria for use by several small reactor projects in their licensing discussions with the Atomic Energy Control Board. The small reactor projects or facilities represented included the MAPLE-X10 reactor, the proposed SES-10 heating reactor and its prototype, the SDR reactor at the Whiteshell Laboratories, the Korea Multipurpose Research Reactor (a.k.a., HANARO) in Korea, the SCORE project, and the McMaster University Nuclear Reactor. The top level set of criteria which form a safety philosophy and serve as a framework for more detailed developments was presented at an IAEA Conference in 1989. AECL continued this work to develop safety principles and design criteria for new small reactors. The first major application of this work has been to the design, safety analysis and licensing of the MAPLE 1 and 2 reactors for the MDS Nordion Medical Isotope Reactor Project. This paper provides an overview of the safety principles and design criteria. Examples of an implementation of these safety principles and design criteria are drawn from the work to design the MAPLE 1 and 2 reactors. (author)

  3. The national standards program for research reactors

    International Nuclear Information System (INIS)

    Whittemore, W.L.

    1977-01-01

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

  4. Contributions of research Reactors in science and technology

    International Nuclear Information System (INIS)

    Butt, N.M.; Bashir, J.

    1992-12-01

    In the present paper, after defining a research reactor, its basic constituents, types of reactors, their distribution in the world, some typical examples of their uses are given. Particular emphasis in placed on the contribution of PARR-I (Pakistan Research Reactor-I), the 5 MW Swimming Pool Research reactor which first became critical at the Pakistan Institute of Nuclear Science and Technology (PINSTECH) in Dec. 1965 and attained its full power in June 1966. This is still the major research facility at PINSTECH for research and development. (author)

  5. Extensive utilization of training reactor VR-1

    International Nuclear Information System (INIS)

    Karel, Matejka; Lubomir, Sklenka

    2005-01-01

    This paper describes one of the main purposes of the VR-1 training reactor utilisation - i.e. extensive educational programme. The educational programme is intended for the training of university students (all technical universities in Czech Republic) and selected nuclear power plant personnel. At the present, students can go through more than 20 different experimental exercises. An attractive programme including demonstration of reactor operation is prepared also for high school students. Moreover, research and development works and information programmes proceed at the VR-1 reactor as well

  6. Development of Pneumatic Transfer Irradiation Facility (PTS no.1) for Neutron Activation Analysis at HANARO Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Y. S.; Moon, J. H.; Kim, S. H.; Sun, G. M.; Baek, S. Y.; Kim, H. R.; Kim, Y. J

    2008-03-15

    A pneumatic transfer system (PTS) is one of the most important facilities used during neutron irradiation of a target material for instrumental neutron activation analysis (INAA) in a research reactor. In particular, a fast pneumatic transfer system is essential for the measurement of a short half-life nuclide and a delayed neutron counting system. The pneumatic transfer system (PTS no.1) involving a manual system and an semiautomatic system were reconstructed with new designs of a functional improvement at the HANARO research reactor in 2006. In this technical report, the conception, design, operation and control of these system (PTS no.1) was described. Also the experimental results and the characteristic parameters measured by a mock-up test, a functional operation test and an irradiation test of these systems, such as the transfer time of irradiation capsule, the different neutron flux, the temperature of the irradiation position with an irradiation time, the radiation dose rate when the rabbit is returned, etc. are reported to provide a user information as well as a reactor's management and safety.

  7. Generic component reliability data for research reactor PSA

    International Nuclear Information System (INIS)

    1997-02-01

    The purpose of this document is to provide reference generic component-reliability information for a variety of research reactor types. As noted in Section 2 and Table IV, component data accumulated over many years is in the database. It is expected that the report should provide representative data which will remain valid for a number of years. The database provides component failure rates on a time and/or demand related basis according to the operational modes of the components. No update of the database is presently planned. As a result of the implementation of data collection systems in the research reactors represented in these studies, updating of data from individual facilities could be made available by the contributing research reactor facilities themselves. As noted in Section 1.1, the report does not include a detailed discussion of information regarding component classification and reliability parameter definitions. The report does provide some insights and discussions regarding the practicalities of the data collection process and some guidelines for database usage. 9 refs, 7 tabs

  8. Experience and prospects for developing research reactors of different types

    International Nuclear Information System (INIS)

    Kuatbekov, R.P.; Tretyakov, I.T.; Romanov, N.V.; Lukasevich, I.B.

    2015-01-01

    NIKIET has a 60-year experience in the development of research reactors. Altogether, there have been more than 25 NIKIET-designed plants of different types built in Russia and 20 more in other countries, including pool-type water-cooled and water moderated research reactors, tank-type and pressure-tube research reactors, pressurized high-flux, heavy-water, pulsed and other research reactors. Most of the research reactors were designed as multipurpose plants for operation at research centers in a broad range of applications. Besides, unique research reactors were developed for specific application fields. Apart from the experience in the development of research reactor designs and the participation in the reactor construction, a unique amount of knowledge has been gained on the operation of research reactors. This makes it possible to use highly reliable technical solutions in the designs of new research reactors to ensure increased safety, greater economic efficiency and maintainability of the reactor systems. A multipurpose pool-type research reactor of a new generation is planned to be built at the Center for Nuclear Energy Science & Technology (CNEST) in the Socialist Republic of Vietnam to be used to support a spectrum of research activities, training of skilled personnel for Vietnam nuclear industry and efficient production of isotopes. It is exactly the applications a research reactor is designed for that defines the reactor type, design and capacity, and the selection of fuel and components subject to all requirements of industry regulations. The design of the new research reactor has a great potential in terms of upgrading and installation of extra experimental devices. (author)

  9. Selecting a MAPLE research reactor core for 1-10 mW operation

    International Nuclear Information System (INIS)

    Smith, H.J.; Roy, M.-F.; Carlson, P.A.

    1986-06-01

    The MAPLE class of research reactors is designed so that a single reactor concept can satisfy a wide range of practical applications. This paper reports the results of physics studies performed on a number of potential core configurations fuelled with either 5 w/o or 8 w/o enriched UO 2 or 20 w/o U 3 Si-Al and assesses the relative merits of each. Recommended core designs are given to maximize the neutron fluxes available for scientific application and isotope production

  10. Reliability studies in research reactors

    International Nuclear Information System (INIS)

    Albuquerque, Tob Rodrigues de

    2013-01-01

    Fault trees and event trees are widely used in industry to model and to evaluate the reliability of safety systems. Detailed analyzes in nuclear installations require the combination of these two techniques. This study uses the methods of FT (Fault Tree) and ET (Event Tree) to accomplish the PSA (Probabilistic Safety Assessment) in research reactors. According to IAEA (lnternational Atomic Energy Agency), the PSA is divided into Level 1, Level 2 and Level 3. At the Level 1, conceptually, the security systems perform to prevent the occurrence of accidents, At the Level 2, once accidents happened, this Level seeks to minimize consequences, known as stage management of accident, and at Level 3 accident impacts are determined. This study focuses on analyzing the Level 1, and searching through the acquisition of knowledge, the consolidation of methodologies for future reliability studies. The Greek Research Reactor, GRR-1, is a case example. The LOCA (Loss of Coolant Accident) was chosen as the initiating event and from it, using ET, possible accidental sequences were developed, which could lead damage to the core. Moreover, for each of affected systems, probabilities of each event top of FT were developed and evaluated in possible accidental sequences. Also, the estimates of importance measures for basic events are presented in this work. The studies of this research were conducted using a commercial computational tool SAPHIRE. Additionally, achieved results thus were considered satisfactory for the performance or the failure of analyzed systems. (author)

  11. Conversion of research reactors to low-enrichment uranium fuels

    International Nuclear Information System (INIS)

    Muranaka, R.G.

    1983-01-01

    There are at present approximately 350 research reactors in 52 countries ranging in power from less than 1 watt to 100 Megawatt and over. In the 1970's, many people became concerned about the possibility that some fuels and fuel cycles could provide an easy route to the acquisition of nuclear weapons. Since enrichment to less than 20% is internationally recognized as a fully adequate barrier to weapons usability, certain Member States have moved to minimize the international trade in highly enriched uranium and have established programmes to develop the technical means to help convert research reactors to the use of low-enrichment fuels with minimum penalties. This could involve modifications in the design of the reactor and development of new fuels. As a result of these programmes, it is expected that most research reactors can be converted to the use of low-enriched fuel

  12. Research reactor's role in Korea

    International Nuclear Information System (INIS)

    Choi, C-O.

    1995-01-01

    After a TRIGA MARK-II was constructed in 1962, new research activity of a general nature, utilizing neutrons, prevailed in Korea. Radioisotopes produced from the MARK-II played a good role in the 1960's in educating people as to what could be achieved by a neutron source. Because the research reactor had implanted neutron science in the country, another TRIGA MARK-III had to be constructed within 10 years after importing the first reactor, due to increased neutron demand from the nuclear community. With the sudden growth of nuclear power, however, the emphasis of research changed. For a while research activities were almost all oriented to nuclear power plant technology. However, the specifics of nuclear power plant technology created a need for a more highly capable research reactor like HANARO 30MWt. HANARO will perform well with irradiation testing and other nuclear programs in the future, including: production of key radioisotopes, doping of silicon by transmutation, neutron activation analysis, neutron beam experiments, cold neutron source. 3 tabs., 2 figs

  13. Making better use of research reactors

    International Nuclear Information System (INIS)

    1964-01-01

    Some 250 research reactors are in operation in the world today, and there are problems in putting them to the most fruitful use. The difficulties - of trained manpower, of auxiliary equipment, of satisfactory research programmes, of co-ordination, between the various disciplines - are common to all users. But as is only to be expected, they press more heavily on the newly-established centres, particularly those in the developing countries which are lacking in long experience in research and usually severely limited as to technical manpower and money. The IAEA has been turning its attention to this question for the past three or four years - ever since, in fact, its early assistance missions and other field operations brought it into close contact with the operations of numerous Member States. The task of providing assistance and advice in this matter is growing. Many centres have been building research reactors under bilateral arrangements; with the completion of their projects this form of aid usually ends, and they look to IAEA for help in operating the reactors. Although some critics consider that difficulties have been caused by premature construction of research reactors, before well-founded programmes of nuclear research had been developed in the countries concerned, several valid motives have led to the establishment of some of these centres at an early stage. A research reactor often provides an effective stimulant for scientific research in the country. It is a remarkably versatile tool for workers in many fields of science and technology. There have been instances where the establishment of a research reactor has had a great impact on the scientific education of a country and has led to a salutary reappraisal and reforms. A reactor is sometimes considered to be a particularly effective means of retaining in the country men trained in the nuclear field. This particular problem is common to most countries. In fact, it is a feature of the present age that

  14. Keeping research reactors relevant: A pro-active approach for SLOWPOKE-2

    International Nuclear Information System (INIS)

    Cosby, L.R.; Bennett, L.G.I.; Nielsen, K.; Weir, R.

    2010-01-01

    The SLOWPOKE is a small, inherently safe, pool-type research reactor that was engineered and marketed by Atomic Energy of Canada Limited (AECL) in the 1970s and 80s. The original reactor, SLOWPOKE-1, was moved from Chalk River to the University of Toronto in 1970 and was operated until upgraded to the SLOWPOKE-2 reactor in 1973. In all, eight reactors in the two versions were produced and five are still in operation today, three having been decommissioned. All of the remaining reactors are designated as SLOWPOKE-2 reactors. These research reactors are prone to two major issues: aging components and lack of relevance to a younger audience. In order to combat these problems, one SLOWPOKE -2 facility has embraced a strategy that involves modernizing their reactor in order to keep the reactor up to date and relevant. In 2001, this facility replaced its aging analogue reactor control system with a digital control system. The system was successfully commissioned and has provided a renewed platform for student learning and research. The digital control system provides a better interface and allows flexibility in data storage and retrieval that was never possible with the analogue control system. This facility has started work on another upgrade to the digital control and instrumentation system that will be installed in 2010. The upgrade includes new computer hardware, updated software and a web-based simulation and training system that will allow licensed operators, students and researchers to use an online simulation tool for training, education and research. The tool consists of: 1) A dynamic simulation for reactor kinetics (e.g., core flux, power, core temperatures, etc). This tool is useful for operator training and student education; 2) Dynamic mapping of the reactor and pool container gamma and neutron fluxes as well as the vertical neutron beam tube flux. This research planning tool is used for various researchers who wish to do irradiations (e.g., neutron

  15. Safety-related parameters for the MAPLE research reactor and a comparison with the IAEA generic 10-MW research reactor

    International Nuclear Information System (INIS)

    Carlson, P.A.; Lee, A.G.; Smith, H.J.; Ellis, R.J.

    1989-07-01

    A summary is presented of some of the principle safety-related physics parameters for the MAPLE Research Reactor, and a comparison with the IAEA Generic 10-MW Reactor is given. This provides a means to assess the operating conditions and fuelling requirements for safe operation of the MAPLE Research Reactor under accepted standards

  16. Annual report of department of research reactor, 1999. April 1, 1999 - March 31, 2000

    International Nuclear Information System (INIS)

    2001-03-01

    The Department of Research Reactor is responsible for the operation, maintenance, utilization of the JRR-3M (new JRR-3) and the JRR-4 and for the related R and D. Besides the decommissioning of the JRR-2 and RI production including its R and D are carried out. This report describes the activities of the department in fiscal year of 1999 and it also includes some of the technical topics on the works mentioned above. As for the research reactors, we carried out the operation, maintenance, the utilization of irradiation and neutron beam experiments, technical management including fuels and water chemistry, radiation monitoring as related R and D works. RI production and its R and D works were conducted as well. The international cooperations between the developing countries and the department were also made concerning the operation, utilization and safety analysis for research reactors. (author)

  17. Biological dosimetry studies for boron neutron capture therapy at the RA-1 research reactor facility

    International Nuclear Information System (INIS)

    Trivillin, Veronica A.; Heber, Elisa M.; Itoiz, Maria E.; Schwint, Amanda E.; Castillo, Jorge

    2004-01-01

    Initial physical dosimetry measurements have been completed using activation spectrometry and thermoluminescent dosimeters to characterize the BNCT facility developed at the RA-1 research reactor operated by the National Atomic Energy Commission in Buenos Aires. Biological dosimetry was performed employing the hamster cheek pouch oral cancer model previously validated for BNCT studies by our group. Results indicate that the RA-1 neutron source produces useful dose rates for BNCT studies but that some improvements in the initial configuration will be needed to optimize the spectrum for thermal-neutron BNCT research applications. (author)

  18. Research and development of a super fast reactor (12). Considerations for the reactor characteristics

    International Nuclear Information System (INIS)

    Goto, Shoji; Ishiwatari, Yuki; Oka, Yoshiaki

    2008-01-01

    A research program aimed at developing the Super Fast Reactor (Super FR) has been entrusted by the Ministry of Education, Culture, Sports, Science and Technology (MEXT) of Japan since December 2005. It includes the following three projects. (A) Development of the Super Fast Reactor concept. (B)Thermal-hydraulic experiments. (C) Materials development. Tokyo Electric Power Company (TEPCO) has joined this program and works on part (A) together with the University of Tokyo. From the utility's viewpoint, it is important to consider the most desirable characteristics for Super FR to have. Four issues were identified in project (A), (1) Fuel design, (2) Reactor core design, (3) Safety, and (4) Plant characteristics of Super FR. This report describes the desired characteristics of Super FR with respect to item (1) fuel design and item (2) Reactor core design, as compared with a boiling water reactor (BWR) plant. The other two issues will be discussed in this project, and will also be considered in the design process of Super FR. (author)

  19. Neutrons down-under: Australia's research reactor review

    International Nuclear Information System (INIS)

    Murray, Allan

    1995-01-01

    Australian research reactor review commenced in September 1992, the Review had the following Terms of Reference: Whether, on review of the benefits and costs for scientific, commercial, industrial and national interest reasons, Australia has a need for a new reactor; a review of the present reactor, HIFAR, to include: an assessment of national and commercial benefits and costs of operations, its likely remaining useful life and its eventual closure and decommissioning; if Australia has a need for a new nuclear research reactor, the Review will consider: possible locations for a new reactor, its environmental impact at alternative locations, recommend a preferred location, and evaluate matters associated with regulation of the facility and organisational arrangements for reactor-based research. From the Review findings the following recommendations were stated: keep HIFAR going; commission a PRA to ascertain HIFAR's remaining life and refurbishment possibilities; identify and establish a HLW repository; accept that neither HIFAR nor a new reactor can be completely commercial; any decision on a new neutron source must rest primarily on benefits to science and Australia's national interest; make a decision on a new neutron source in about five years' time (1998). Design Proposals for a New Reactor are specified

  20. An Open Source-based Approach to the Development of Research Reactor Simulator

    International Nuclear Information System (INIS)

    Joo, Sung Moon; Suh, Yong Suk; Park, Cheol Park

    2016-01-01

    In reactor design, operator training, safety analysis, or research using a reactor, it is essential to simulate time dependent reactor behaviors such as neutron population, fluid flow, and heat transfer. Furthermore, in order to use the simulator to train and educate operators, a mockup of the reactor user interface is required. There are commercial software tools available for reactor simulator development. However, it is costly to use those commercial software tools. Especially for research reactors, it is difficult to justify the high cost as regulations on research reactor simulators are not as strict as those for commercial Nuclear Power Plants(NPPs). An open source-based simulator for a research reactor is configured as a distributed control system based on EPICS framework. To demonstrate the use of the simulation framework proposed in this work, we consider a toy example. This example approximates a 1-second impulse reactivity insertion in a reactor, which represents the instantaneous removal and reinsertion of a control rod. The change in reactivity results in a slightly delayed change in power and corresponding increases in temperatures throughout the system. We proposed an approach for developing research reactor simulator using open source software tools, and showed preliminary results. The results demonstrate that the approach presented in this work can provide economical and viable way of developing research reactor simulators

  1. Safety aspects on dependability management for a TRIGA research reactor in Romania

    International Nuclear Information System (INIS)

    Vieru, Gheorghe

    2002-01-01

    Safety on the management for a nuclear research reactor involves a 'good dependability management' of the activities, such as: reliability, availability, maintainability and maintenance support. In order to evaluate the safety management aspects intended to be applied at a research reactor management, the performance dependability indicators and their impact over the availability and reactor safety have to be established. The document ISO 9000-4/IEC 300-1 'Dependability Management' (1995), describes five internationally agreed indicators of the reactor equipment dependability, each of them can be used for corrective maintenance or for preventive maintenance, such as: I 1 - equipment Maintenance Frequency; I 2 - equipment Maintenance Effort; I-3 - equipment Maintenance Downtime Factor; I 4 - equipment Maintenance Contribution to the System Function Downtime Factor; I 5 - equipment Maintenance Contribution to the reactor Capability Loss Factor. The paper presents an evaluation of those 5 mentioned indicators with referring only at the primary circuit of the INR's TRIGA research reactor and conclusion. The analyzed period was stated between 1994-1999. It is to be noted that this type of analyze is performed for the first time for a research reactor. (author)

  2. Safety considerations for research reactors in extended shutdown

    International Nuclear Information System (INIS)

    2004-01-01

    According to the IAEA Research Reactor Database, in the last 20 years, 367 research reactors have been shut down. Of these, 109 have undergone decommissioning and the rest are in extended shutdown with no clear definition about their future. Still other research reactors are infrequently operated with no meaningful utilization programme. These two situations present concerns related to safety such as loss of corporate memory, personnel qualification, maintenance of components and systems and preparation and maintenance of documentation. There are many reasons to shut down a reactor; these may include: - the need to carry out modifications in the reactor systems; - the need for refurbishment to extend the lifetime of the reactor; - the need to repair reactor structures, systems, or components; - the need to remedy technical problems; - regulatory or public concerns; - local conflicts or wars; - political convenience; - the lack of resources. While any one of these reasons may lead to shutdown of a reactor, each will present unique problems to the reactor management. The large variations from one research reactor to the next also will contribute to the uniqueness of the problems. Any option that the reactor management adopts will affect the future of the facility. Options may include dealing with the cause of the shutdown and returning to normal operation, extending the shutdown period waiting a future decision, or decommissioning. Such options are carefully and properly analysed to ensure that the solution selected is the best in terms of reactor type and size, period of shutdown and legal, economic and social considerations. This publication provides information in support of the IAEA safety standards for research reactors

  3. Improvements at the biological shielding of BNCT research facility in the IEA-R1 reactor

    International Nuclear Information System (INIS)

    Souza, Gregorio Soares de

    2011-01-01

    The technique of neutron capture in boron is a promising technique in cancer treatment, it uses the high LET particles from the reaction 10 B (n, α) 7 Li to destroy cancer cells.The development of this technique began in the mid-'50s and even today it is the object of study and research in various centers around the world, Brazil has built a facility that aims to conduct research in BNCT, this facility is located next to irradiation channel number three at the research nuclear reactor IEA-R1 and has a biological shielding designed to meet the radiation protection standards. This biological shielding was developed to allow them to conduct experiments with the reactor at maximum power, so it is not necessary to turn on and off the reactor to irradiate samples. However, when the channel is opened for experiments the background radiation in the experiments salon increases and this background variation makes it impossible to perform measurements in a neutron diffraction research that utilizes the irradiation channel number six. This study aims to further improve the shielding in order to minimize the variation of background making it possible to perform the research facility in BNCT without interfering with the action of the research group of the irradiation channel number six. To reach this purpose, the code MCNP5, dosimeters and activation detectors were used to plan improvements in the biological shielding. It was calculated with the help of the code an improvement that can reduce the average heat flow in 71.2% ± 13 and verified experimentally a mean reduce of 70 ± 9% in dose due to thermal neutrons. (author)

  4. Usage of burnable poison on research reactors

    International Nuclear Information System (INIS)

    Villarino, Eduardo Anibal

    2002-01-01

    The fuel assemblies with burnable poison are widely used on power reactors, but there are not commonly used on research reactors. This paper shows a neutronic analysis of the advantages and disadvantages of the burnable poison usage on research reactors. This paper analyses both burnable poison design used on research reactors: Boron on the lateral wall and Cadmium wires. Both designs include a parametric study on the design parameters like the amount and geometry of the burnable poison. This paper presents the design flexibility using burnable poisons, it does not find an optimal or final design, which it will strongly depend on the core characteristics and fuel management strategy. (author)

  5. Study on the decommissioning of research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Suh, Doo Hwan; Jun, Kwan Sik; Choi, Yoon Dong; Lee, Tae Yung; Kwon, Sang Woon; Lee, Jong Il [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-01-01

    Currently, KAERI operates TRIGA Mark-II and TRIGA Mark-III research reactors as a general purpose research and training facility. As these are, however, situated at Seoul office site of KAERI which is scheduled to be transferred to KEPCO as well as 30 MW HANARO research reactor which is expected to reach the first criticality in 1995 is under construction at head site of KAERI, decommissioning of TRIGA reactors has become an important topic. The objective of this study is to prepare and present TRIGA facility decontamination and decommissioning plan. Estimation of the radioactive inventory in TRIGA research reactor was carried out by the use of computational method. In addition, summarized in particular were the methodologies associated with decontamination, segmenting processes for activated metallic components, disposition of wastes. Particular consideration in this study was focused available technology applicable to decommissioning of TRIGA research reactor. State-of-the-art summaries of the available technology for decommissioning presented here will serve a useful document for preparations for decommissioning in the future. 6 figs, 41 tabs, 30 refs. (Author).

  6. Calculation of photon dose for Dalat research reactor in case of loss of reactor tank water

    International Nuclear Information System (INIS)

    Le Vinh Vinh; Huynh Ton Nghiem; Nguyen Kien Cuong

    2007-01-01

    Photon sources of actinides and fission products were estimated by ORIGEN2 code with the modified cross-section library for Dalat research reactor (DRR) using new cross-section generated by WIMS-ANL code. Photon sources of reactor tank water calculated from the experimental data. MCNP4C2 with available non-analog Monte Carlo model and ANSI/ANL-6.1.1-1977 flux-to-dose factors were used for dose estimation. The agreement between calculation results and those of measurements showed that the methods and models used to get photon sources and dose were acceptable. In case the reactor water totally leaks out from the reactor tank, the calculated dose is very high at the top of reactor tank while still low in control room. In the reactor hall, the operation staffs can access for emergency works but with time limits. (author)

  7. Ageing Management for Research Reactors. Specific Safety Guide

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2010-10-15

    This Safety Guide was developed under the IAEA programme for safety standards for research reactors, which covers all the important areas of research reactor safety. It supplements and elaborates upon the safety requirements for ageing management of research reactors that are established in paras 6.68-6.70 and 7.109 of the IAEA Safety Requirements publication, Safety of Research Reactors. The safety of a research reactor requires that provisions be made in its design to facilitate ageing management. Throughout the lifetime of a research reactor, including its decommissioning, ageing management of its structures, systems and components (SSCs) important to safety is required, to ensure continued adequacy of the safety level, reliable operation of the reactor, and compliance with the operational limits and conditions. Managing the safety aspects of research reactor ageing requires implementation of an effective programme for the monitoring, prediction, and timely detection and mitigation of degradation of SSCs important to safety, and for maintaining their integrity and functional capability throughout their service lives. Ageing management is defined as engineering, operation, and maintenance strategy and actions to control within acceptable limits the ageing degradation of SSCs. Ageing management includes activities such as repair, refurbishment and replacement of SSCs, which are similar to other activities carried out at a research reactor in maintenance and testing or when a modification project takes place. However, it is important to recognize that effective management of ageing requires the use of a methodology that will detect and evaluate ageing degradation as a consequence of the service conditions, and involves the application of countermeasures for prevention and mitigation of ageing degradation. The objective of this Safety Guide is to provide recommendations on managing ageing of SSCs important to safety at research reactors on the basis of international

  8. Ageing Management for Research Reactors. Specific Safety Guide

    International Nuclear Information System (INIS)

    2010-01-01

    This Safety Guide was developed under the IAEA programme for safety standards for research reactors, which covers all the important areas of research reactor safety. It supplements and elaborates upon the safety requirements for ageing management of research reactors that are established in paras 6.68-6.70 and 7.109 of the IAEA Safety Requirements publication, Safety of Research Reactors. The safety of a research reactor requires that provisions be made in its design to facilitate ageing management. Throughout the lifetime of a research reactor, including its decommissioning, ageing management of its structures, systems and components (SSCs) important to safety is required, to ensure continued adequacy of the safety level, reliable operation of the reactor, and compliance with the operational limits and conditions. Managing the safety aspects of research reactor ageing requires implementation of an effective programme for the monitoring, prediction, and timely detection and mitigation of degradation of SSCs important to safety, and for maintaining their integrity and functional capability throughout their service lives. Ageing management is defined as engineering, operation, and maintenance strategy and actions to control within acceptable limits the ageing degradation of SSCs. Ageing management includes activities such as repair, refurbishment and replacement of SSCs, which are similar to other activities carried out at a research reactor in maintenance and testing or when a modification project takes place. However, it is important to recognize that effective management of ageing requires the use of a methodology that will detect and evaluate ageing degradation as a consequence of the service conditions, and involves the application of countermeasures for prevention and mitigation of ageing degradation. The objective of this Safety Guide is to provide recommendations on managing ageing of SSCs important to safety at research reactors on the basis of international

  9. Some considerations for assurance of reactor safety from experiences in research reactors

    International Nuclear Information System (INIS)

    Okamoto, Sunao; Nishihara, Hideaki; Shibata, Toshikazu

    1981-01-01

    For the purpose of assuring reactor safety and strengthening research in the related fields, a multi-disciplinary group was formed among university researchers, including social scientists, with a special allocation of the Grant-in-Aid from the Ministry of Education, Science and Culture. An excerpt from the first year's report (1979 -- 1980) is edited here, which contains an interpretation of Murphy's reliability engineering law, a scope of reactor diagnostic studies to be pursued at universities, and safety measures already implemented or suggested to be implemented in university research reactors. (author)

  10. Integrated Management System, Configuration and Document Control for Research Reactors

    International Nuclear Information System (INIS)

    Steynberg, B.J.; Bruyn, J.F. du

    2017-01-01

    . Document control helps ensure that: 1. Important facility documents are properly stored; 2. Revisions to documents are controlled, tracked, and completed in a timely manner; 3. Revised documents are formally distributed to designated users; and 4. Information concerning pending revisions is made available. Configuration and document management within an integrated management system are essential requirements for the safe operation, utilisation and modification of any research reactor. (author)

  11. The modification of the Rossendorf Research Reactor

    International Nuclear Information System (INIS)

    Gehre, G.; Hieronymus, W.; Kampf, T.; Ringel, V.; Robbander, W.

    1990-01-01

    The Rossendorf Research Reactor is of the WWR-SM type. It is a heterogeneous water moderated and cooled tank reactor with a thermal power of 10 MW, which was in operation from 1957 to 1986. It was shut down in 1987 for comprehensive modifications to increase its safety and to improve the efficiency of irradiation and experimentals. The modifications will be implemented in two steps. The first one to be finished in 1989 comprises: 1) the replacement of the reactor tank and its components, the reactor cooling system, the ventilation system and the electric power installation; 2) the construction of a new reactor control room and of filtering equipment; 3) the renewal of process instrumentation and control engineering equipment for reactor operation, equipment for radiation protection monitoring, and reactor operation and safety documentation. The second step, to be implemented in the nineties, is to comprise: 1) the enlargement of the capacity for storage of spent fuel; 2) the modernization of reactor operations by computer-aided control; 3) the installation of an automated measuring systems for accident and environmental monitoring. Two objects of the modification, the replacement of the reactor tank and the design of a new and safer one as well as the increase of the redundancy of the core emergency cooling system are described in detail. For the tank replacement the exposure data are also given. Furthermore, the licensing procedures based on national ordinances and standards as well as on international standards and recommendations and the mutual responsibilities and activities of the licensing authority and of the reactor manager are presented. Finally, the present state of the modifications and the schedule up to the reactor recommissioning and test operation at full power is outlined

  12. Kartini Research Reactor prospective studies for neutron scattering application

    International Nuclear Information System (INIS)

    Widarto

    1999-01-01

    The Kartini Research Reactor (KRR) is located in Yogyakarta Nuclear Research Center, Yogyakarta - Indonesia. The reactor is operated for 100 kW thermal power used for research, experiments and training of nuclear technology. There are 4 beam ports and 1 column thermal are available at the reactor. Those beam ports have thermal neutron flux around 10 7 n/cm 2 s each other and used for sub critical assembly, neutron radiography studies and Neutron Activation Analysis (NAA). Design of neutron collimator has been done for piercing radial beam port and the calculation result of collimated neutron flux is around 10 9 n/cm 2 s. This paper describes experiment facilities and parameters of the Kartini research reactor, and further more the prospective studies for neutron scattering application. The purpose of this paper is to optimize in utilization of the beam ports facilities and enhance the manpower specialty. The special characteristic of the beam ports and preliminary studies, pre activities regarding with neutron scattering studies for KKR is presented. (author)

  13. 1986 international meeting of nuclear reactor safety committees

    International Nuclear Information System (INIS)

    Moeller, D.W.

    1987-01-01

    During the week of Oct. 20-23, 1986, nuclear power-plant safety representatives from the Federal Republic of Germany (FRG), France, Japan, and the US assembled to discuss subjects of mutual interest. The meeting, the first of its kind, was organized under the leadership and direction of David A. Ward, Chairman, Advisory Committee on Reactor Safeguards, US Nuclear Regulatory Commission (NRC), and held at the Wingspread Conference Center of the Johnson Foundation in Racine, Wis. Approximately 40 representatives of the several countries attended. Discussions were candid and provided the participants an opportunity to share thoughts and information on nuclear safety concerns and solutions

  14. Probabilistic safety assessment for research reactors

    International Nuclear Information System (INIS)

    1986-12-01

    Increasing interest in using Probabilistic Safety Assessment (PSA) methods for research reactor safety is being observed in many countries throughout the world. This is mainly because of the great ability of this approach in achieving safe and reliable operation of research reactors. There is also a need to assist developing countries to apply Probabilistic Safety Assessment to existing nuclear facilities which are simpler and therefore less complicated to analyse than a large Nuclear Power Plant. It may be important, therefore, to develop PSA for research reactors. This might also help to better understand the safety characteristics of the reactor and to base any backfitting on a cost-benefit analysis which would ensure that only necessary changes are made. This document touches on all the key aspects of PSA but placed greater emphasis on so-called systems analysis aspects rather than the in-plant or ex-plant consequences

  15. Safety evaluation of the Dalat research reactor operation

    International Nuclear Information System (INIS)

    Long, V.H.; Lam, P.V.; An, T.K.

    1989-01-01

    After an introduction presenting the essential characteristics of the Dalat Nuclear Research Reactor, the document presents i) The safety assurance condition of the reactor, ii) Its safety behaviour after 5 years of operation, iii) Safety research being realized on the reactor. Following is questionnaire of safety evaluation and a list of attachments, which concern the reactor

  16. Enhancement of research reactor utilization in the developing countries

    International Nuclear Information System (INIS)

    Bashir, J.; Butt, N.M.

    1994-06-01

    As the research reactor represents a significant capital investment on the part of any institution and in addition there are recurring annual operating costs, therefore, the subject of its effective utilization has always been of interest. World wide there are about three hundred research reactors. Of these, 92 are located in the developing countries. Together, these reactors represent quite significant research potential. In the present paper, reasons of under utilization, procedures necessary to measure the productivity, ways and means of enhancing the utilization of research reactors are described. In the end, use of two research reactors at PINSTECH are described to illustrate some of the ways in which a successful utilization of a research reactor can made in the developing country. (author) 9 figs

  17. COOLOD, Steady-State Thermal Hydraulics of Research Reactors

    International Nuclear Information System (INIS)

    Kaminaga, Masanori

    1997-01-01

    1 - Description of program or function: The COOLOD-N2 code provides a capability for the analyses of the steady-state thermal-hydraulics of research reactors. This code is a revised version of the COOLOD-N code, and is applicable not only for research reactors in which plate-type fuel is adopted, but also for research reactors in which rod-type fuel is adopted. In the code, subroutines to calculate temperature distribution in rod-type fuel have been newly added to the COOLOD-N code. The COOLOD-N2 code can calculate fuel temperatures under both forced convection cooling mode and natural convection cooling mode. A 'Heat Transfer package' is used for calculating heat transfer coefficient, DNB heat flux etc. The 'Heat Transfer package' is a subroutine program and is especially developed for research reactors in which plate-type fuel is adopted. In case of rod-type fuel, DNB heat flux is calculated by both the 'Heat Transfer package' and Lund DNB heat flux correlation which is popular for TRIGA reactor. The COOLOD-N2 code also has a capability of calculating ONB temperature, the heat flux at onset of flow instability as well as DNB heat flux. 2 - Method of solution: The 'Heat Transfer Package' is a subprogram for calculating heat transfer coefficients, ONB temperature, heat flux at onset of flow instability and DNB heat flux. The 'Heat transfer package' was especially developed for research reactors which are operated under low pressure and low temperature conditions using plate-type fuel, just like the JRR-3M. Heat transfer correlations adopted in the 'Heat Transfer Package' were obtained or estimated based on the heat transfer experiments in which thermal-hydraulic features of the upgraded JRR-3 core were properly reflected. The 'Heat Transfer Package' is applicable to upward and downward flow

  18. Advanced Research Reactor Fuel Development

    Energy Technology Data Exchange (ETDEWEB)

    Kim, C. K.; Park, H. D.; Kim, K. H. (and others)

    2006-04-15

    RERTR program for non-proliferation has propelled to develop high-density U-Mo dispersion fuels, reprocessable and available as nuclear fuel for high performance research reactors in the world. As the centrifugal atomization technology, invented in KAERI, is optimum to fabricate high-density U-Mo fuel powders, it has a great possibility to be applied in commercialization if the atomized fuel shows an acceptable in-reactor performance in irradiation test for qualification. In addition, if rod-type U-Mo dispersion fuel is developed for qualification, it is a great possibility to export the HANARO technology and the U-Mo dispersion fuel to the research reactors supplied in foreign countries in future. In this project, reprocessable rod-type U-Mo test fuel was fabricated, and irradiated in HANARO. New U-Mo fuel to suppress the interaction between U-Mo and Al matrix was designed and evaluated for in-reactor irradiation test. The fabrication process of new U-Mo fuel developed, and the irradiation test fuel was fabricated. In-reactor irradiation data for practical use of U-Mo fuel was collected and evaluated. Application plan of atomized U-Mo powder to the commercialization of U-Mo fuel was investigated.

  19. Development of Neutron Imaging System for Neutron Tomography at Thai Research Reactor TRR-1/M1

    Science.gov (United States)

    Wonglee, S.; Khaweerat, S.; Channuie, J.; Picha, R.; Liamsuwan, T.; Ratanatongchai, W.

    2017-09-01

    The neutron imaging is a powerful non-destructive technique to investigate the internal structure and provides the information which is different from the conventional X-ray/Gamma radiography. By reconstruction of the obtained 2-dimentional (2D) images from the taken different angle around the specimen, the tomographic image can be obtained and it can provide the information in more detail. The neutron imaging system at Thai Research Reactor TRR-1/M1 of Thailand Institute of Nuclear Technology (Public Organization) has been developed to conduct the neutron tomography since 2014. The primary goal of this work is to serve the investigation of archeological samples, however, this technique can also be applied to various fields, such as investigation of industrial specimen and others. This research paper presents the performance study of a compact neutron camera manufactured by Neutron Optics such as speed and sensitivity. Furthermore, the 3-dimentional (3D) neutron image was successfully reconstructed at the developed neutron imaging system of TRR-1/M1.

  20. The first university research reactor in India

    International Nuclear Information System (INIS)

    Murthy, G.S.

    1999-01-01

    At low power research reactor is being set up in Andhra University to cater to the needs of researchers and isotope users by the Department of Atomic Energy in collaboration with Andhra University. This reactor is expected to be commissioned by 2001-02. Departments like Chemistry, Earth Sciences, Physics, Life Sciences, Pharmacy, Medicine and Engineering would be the beneficiaries of the availability of this reactor. In this paper, details of the envisaged research programme and training activities are discussed. (author)

  1. Proceedings of the European Research Reactor Conference - RRFM 2013 Transactions

    International Nuclear Information System (INIS)

    2013-01-01

    In 2013 RRFM, the European Research Reactor Conference is jointly organised by ENS and Atomexpo LLC. This time the Research Reactor community meet in St. Petersburg, Russia. The conference programme will revolve around a series of Plenary Sessions dedicated to the latest global developments with regards to research reactor technology and management. Parallel sessions will focus on all areas of the Fuel Cycle of Research Reactors, their Utilisation, Operation and Management as well as specific research projects and innovative methods in research reactor analysis and design. In 2013 the European Research Reactor Conference will for the first time give special attention to complementary safety assessments of Research Reactors, following the Fukushima-Dai-Ichi NPP's Accident. (authors)

  2. Review of Operation and Maintenance Support Systems for Research Reactors

    International Nuclear Information System (INIS)

    Jin, Kyungho; Heo, Gyunyoung; Park, Jaekwan

    2014-01-01

    Operation support systems do not directly control the plant but it can aid decision making itself by obtaining and analyzing large amounts of data. Recently, the demand of research reactor is growing and the need for operation support systems is increasing, but it has not been applied for research reactors. This study analyzes operation and maintenance support systems of NPPs and suggests appropriate systems for research reactors based on analysis. In this paper, operation support systems for research reactors are suggested by comparing with those of power reactors. Currently, research reactors do not cover special systems in order to improve safety and operability in comparison with power reactors. Therefore we expect to improve worth to use by introducing appropriate systems for research reactors. In further research, we will develop an appropriate system such as applications or tools that can be applied to the research reactor

  3. Testing of research reactor fuel in the high flux reactor (Petten)

    International Nuclear Information System (INIS)

    Guidez, J.; Markgraf, J.W.; Sordon, G.; Wijtsma, F.J.; Thijssen, P.J.M.; Hendriks, J.A.

    1999-01-01

    The two types of fuel most frequently used by the main research reactors are metallic: highly enriched uranium (>90%) and silicide low enriched uranium ( 3 . However, a need exists for research on new reactor fuel. This would permit some plants to convert without losses in flux or in cycle length and would allow new reactor projects to achieve higher possibilities especially in fluxes. In these cases research is made either on silicide with higher density, or on other types of fuel (UMo, etc.). In all cases when new fuel is proposed, there is a need, for safety reasons, to test it, especially regarding the mechanical evolution due to burn-up (swelling, etc.). Initially, such tests are often made with separate plates, but lately, using entire elements. Destructive examinations are often necessary. For this type of test, the High Flux Reactor, located in Petten (The Netherlands) has many specific advantages: a large core, providing a variety of interesting positions with high fluence rate; a downward coolant flow simplifies the engineering of the device; there exists easy access with all handling possibilities to the hot-cells; the high number of operating days (>280 days/year), together with the high flux, gives a possibility to reach quickly the high burn-up needs; an experienced engineering department capable of translating specific requirements to tailor-made experimental devices; a well equipped hot-cell laboratory on site to perform all necessary measurements (swelling, γ-scanning, profilometry) and all destructive examinations. In conclusion, the HFR reactor readily permits experimental research on specific fuels used for research reactors with all the necessary facilities on the Petten site. (author)

  4. The U.S. reduced enrichment research and test reactor (RERTR) program

    International Nuclear Information System (INIS)

    Travelli, A.

    1993-01-01

    Research and test reactors are widely deployed to study the irradiation behavior of materials of interest in nuclear engineering, to produce radioisotopes for medicine, industry, and agriculture, and as a basic research and teaching tool. In order to maximize neutron flux per unit power and/or to minimize capital costs and fuel cycle costs, most of these reactors were de- signed to utilize uranium with very high enrichment (in the 70% to 95% range). Research reactor fuels with such high uranium enrichment cause a potential risk of nuclear weapons proliferation. Over 140 research and test reactors of significant power (between 10 kW and 250 MW) are in operation with very highly enriched uranium in more than 35 countries, with total power in excess of 1,700 MW. The overall annual fuel requirement of these reactors corresponds to approximately 1,200 kg of 235 U. This highly strategic material is normally exported from the United States, converted to metal form, shipped to a fuel fabricator, and then shipped to the reactor site in finished fuel element form. At the reactor site the fuel is first stored, then irradiated, stored again, and eventually shipped back to the United States for reprocessing. The whole cycle takes approximately four years to complete, bringing the total required fuel inventory to approximately 5,000 kg of 235 U. The resulting international trade in highly-enriched uranium may involve several countries in the process of refueling a single reactor and creates a considerable concern that the highly-enriched uranium may be diverted for non-peaceful purposes while in fabrication, transport, or storage, particularly when it is in the unirradiated form. The proliferation resistance of nuclear fuels used in research and test reactors can be considerably improved by reducing their uranium enrichment to a value less than 20%, but significantly greater than natural to avoid excessive plutonium production

  5. Proceedings of the Twenty-First Water Reactor Safety Information Meeting: Volume 1, Plenary session; Advanced reactor research; advanced control system technology; advanced instrumentation and control hardware; human factors research; probabilistic risk assessment topics; thermal hydraulics; thermal hydraulic research for advanced passive LWRs

    International Nuclear Information System (INIS)

    Monteleone, S.

    1994-04-01

    This three-volume report contains 90 papers out of the 102 that were presented at the Twenty-First Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 25--27, 1993. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Germany, Japan, Russia, Switzerland, Taiwan, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. Individual papers have been cataloged separately. This document, Volume 1 covers the following topics: Advanced Reactor Research; Advanced Instrumentation and Control Hardware; Advanced Control System Technology; Human Factors Research; Probabilistic Risk Assessment Topics; Thermal Hydraulics; and Thermal Hydraulic Research for Advanced Passive Light Water Reactors

  6. Proceedings of the Twenty-First Water Reactor Safety Information Meeting: Volume 1, Plenary session; Advanced reactor research; advanced control system technology; advanced instrumentation and control hardware; human factors research; probabilistic risk assessment topics; thermal hydraulics; thermal hydraulic research for advanced passive LWRs

    Energy Technology Data Exchange (ETDEWEB)

    Monteleone, S. [Brookhaven National Lab., Upton, NY (United States)] [comp.

    1994-04-01

    This three-volume report contains 90 papers out of the 102 that were presented at the Twenty-First Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 25--27, 1993. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Germany, Japan, Russia, Switzerland, Taiwan, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. Individual papers have been cataloged separately. This document, Volume 1 covers the following topics: Advanced Reactor Research; Advanced Instrumentation and Control Hardware; Advanced Control System Technology; Human Factors Research; Probabilistic Risk Assessment Topics; Thermal Hydraulics; and Thermal Hydraulic Research for Advanced Passive Light Water Reactors.

  7. Research reactor utilization. Summary reports of three study group meetings: Irradiation techniques at research reactors, held in Istanbul 15-19 November 1965; Research reactor operation and maintenance problems, held in Caracas 6-10 December 1965; and Research reactor utilization in the Far East, held in Lucas Heights 28 February - 4 March 1966

    International Nuclear Information System (INIS)

    1967-01-01

    The three sections of this book, which are summary reports of three Study Group meetings of the IAEA: Irradiation techniques at research reactors, Istanbul, 15-19 November 1965; Research reactor operation and maintenance problems, Caracas, 6-10 December 1965; and Research reactor utilization in the Far East, Lucas Heights, Australia, 28 February - 4 March 1966. These meetings were the latest in a series designed to promote efficient utilization of research reactors, to disseminate information on advances in techniques, to discuss common problems in reactor operations, and to outline some advanced areas of reactor-based research. (author)

  8. Factors affecting nuclear research reactor utilization across countries

    International Nuclear Information System (INIS)

    Hien, P.D.

    2000-01-01

    In view of the worldwide declining trend of research reactor utilization and the fact that many reactors in developing countries are under-utilised, a question naturally arises as to whether the investment in a research reactor is justifiable. Statistical analyses were applied to reveal relationships between the status of reactor utilization and socio-economic conditions among countries, that may provide a guidance for reactor planning and cost benefit assessment. The reactor power has significant regression relationships with size indicators such as GNP, electricity consumption and R and D expenditure. Concerning the effectiveness of investment in research reactors, the number of reactor operation days per year only weakly correlates with electricity consumption and R and D expenditure, implying that there are controlling factors specific of each group of countries. In the case of less developed countries, the low customer demands on reactor operation may be associated with the failure in achieving quality assurance for the reactor products and services, inadequate investment in the infrastructure for reactor exploitation, the shortage of R and D funding and well trained manpower and the lack of measures to get the scientific community involved in the application of nuclear techniques. (author)

  9. The present situations and perspectives on utilization of research reactors in Thailand

    Science.gov (United States)

    Chongkum, Somporn

    2002-01-01

    The Thai Research Reactor 1/Modification 1, a TRIGA Mark III reactor, went critical on November 7, 1977. It has been playing a central role in the development of both Office of Atomic Energy for Peace (OAEP) and nuclear application in Thailand. It has a maximum power of 2 MW (thermal) at steady state and a pulsing capacity of 2000 MW. The highest thermal neutron flux at a central thimber is 1×10 13 n/cm 2/s, which is extensively utilized for radioisotope production, neutron activation analysis and neutron beam experiments, i.e. neutron scattering, prompt gamma analysis and neutron radiography. Following the nuclear technological development, the OAEP is in the process of establishing the Ongkharak Nuclear Research Center (ONRC). The center is being built in Nakhon Nayok province, 60 km northeast of Bangkok. The centerpiece of the ONRC is a multipurpose 10 MW TRIGA research reactor. Facilities are included for the production of radioisotopes for medicine, industry and agriculture, neutron transmutation doping of silicon, and neutron capture therapy. The neutron beam facilities will also be utilized for applied research and technology development as well as training in reactor operations, performance of experiments and reactor physics. This paper describes a recent program of utilization as well as a new research reactor for enlarging the perspectives of its utilization in the future.

  10. The current status of Kartini research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tjiptono, Tri Wulan; Syarip,

    1998-10-01

    The Kartini reactor reached the first criticality on January 25, 1979. In the first three years, the reactor power is limited up to 50 kW thermal power and on July 1, 1982 has been increased to 100 kW. It has been used as experiments facility by researcher of Atomic Energy National Agency and students of the Universities. Three beam tubes used as experiments facilities, the first, is used as a neutron source for H{sub 2}O-Natural Uranium Subcritical Assembly, the second, is developed for neutron radiography facility and the third, is used for gamma radiography facility. The other facilities are rotary rack and two pneumatic transfer systems, one for delayed neutron counting system and the other for the new Neutron Activation Analysis (NAA) facility. The rotary rack used for isotope production for NAA purpose (for long time irradiation), the delayed neutron counting system used for analysis the Uranium contents of the ores and the new NAA is provided for short live elements analysis. In the last three years the Reactor Division has a joint use program with the Nuclear Component and Engineering Center in research reactor instrumentation and control development. (author)

  11. The Jules Horowitz reactor project, a driver for revival of the research reactor community

    Energy Technology Data Exchange (ETDEWEB)

    Pere, P.; Cavailler, C.; Pascal, C. [AREVA TA, CEA Cadarache - Etablissement d' AREVA TA - Chantier RJH - MOE - BV2 - BP no. 9 - 13115 Saint Paul lez Durance (France); CS 50497 - 1100, rue JR Gauthier de la Lauziere, 13593 Aix en Provence cedex 3 (France)

    2010-07-01

    The first concrete of the nuclear island for the Jules Horowitz Reactor (JHR) was poured at the end of July 2009 and construction is ongoing. The JHR is the largest new platform for irradiation experiments supporting Generation II and III reactors, Generation IV technologies, and radioisotope production. This facility, composed of a unique grouping of workshops, hot cells and hot laboratories together with a first -rate MTR research reactor, will ensure that the process, from preparations for irradiation experiments through post-irradiation non-destructive examination, is completed expediently, efficiently and, of course, safely. In addition to the performance requirements to be met in terms of neutron fluxes on the samples (5x10{sup 14} n.cm{sup -2}/sec{sup -1} E> 1 MeV in core and 3,6x10{sup 14} n.cm{sup -2}/sec{sup -1} E<0.625 eV in the reflector) and the JHR's considerable irradiation capabilities (more than 20 experiments and one-tenth of irradiation area for simultaneous radioisotope production), the JHR is the first MTR to be built since the end of the 1960's, making this an especially challenging project. The presentation will provide an overview of the reactor, hot cells and laboratories and an outline of the key milestones in the project schedule, including initial criticality in early 2014 and radioisotope production in 2015. This will be followed by a description of the project organization set up by the CEA as owner and future operator and AREVA TA as prime contractor and supplier of critical systems, and a discussion of project challenges, especially those dealing with the following items: - accommodation of a broad experimental domain, - involvement by international partners making in-kind contributions to the project, - development of components critical to safety and performance, - the revival of engineering of research reactors and experimental devices involving France's historical players in the field of research reactors, and

  12. The replacement research reactor

    International Nuclear Information System (INIS)

    Cameron, R.; Horlock, K.

    2001-01-01

    The contract for the design, construction and commissioning of the Replacement Research Reactor was signed in July 2000. This was followed by the completion of the detailed design and an application for a construction licence was made in May 2001. This paper will describe the main elements of the design and their relation to the proposed applications of the reactor. The future stages in the project leading to full operation are also described

  13. MAPLE research reactor beam-tube performance

    International Nuclear Information System (INIS)

    Lee, A.G.; Lidstone, R.F.; Gillespie, G.E.

    1989-05-01

    Atomic Energy of Canada Limited (AECL) has been developing the MAPLE (Multipurpose Applied Physics Lattice Experimental) reactor concept as a medium-flux neutron source to meet contemporary research reactor applications. This paper gives a brief description of the MAPLE reactor and presents some results of computer simulations used to analyze the neutronic performance. The computer simulations were performed to identify how the MAPLE reactor may be adapted to beam-tube applications such as neutron radiography

  14. RA Research reactor Annual report 1982 - Part 1, Operation, maintenance and utilization of the RA reactor

    International Nuclear Information System (INIS)

    Sotic, O.; Martinc, R.; Kozomara-Maic, S.; Cupac, S.; Radivojevic, J.; Stamenkovic, D.; Skoric, M.; Miokovic, J.

    1982-12-01

    Reactor test operation started in September 1981 at 2 MW power with 80% enriched fuel continued during 1982 according to the previous plan. The initial reactor core was made of 44 fuel channel each containing 10 fuel slugs. The first half of 1982 was used for the needed measurements and analysis of operating parameters and functioning of reactor systems and equipment under operating conditions. Program concerned with the testing operation at higher power levels was started in the second half of this year. It was found that the inherent excess reactivity and control rod worths ensure safe operation according to the IAEA safety standards. Excess reactivity is high enough to enable higher power level of 4.7 MW during 4 monthly cycles each lasting 15-20 days. Favourable conditions for cooling exist for the initial core configuration. Effects of poisoning at startup on the reactivity and power density distribution were measured as well as initial spatial distribution of the neutron flux which was 3,9 10 13 cm -2 s -1 at 2 MW power. Modification of the calibration coefficient in the system for automated power level control was determined. All the results show that all the safety criteria and limitations concerned with fuel utilization are fulfilled if reactor power would be 4.7 MW. Additional testing operation at 3, 4, and 4.7 MW power levels will be needed after obtaining the licence for operating at nominal power. Transition from the initial core with 44 fuel channels to the equilibrium lattice configuration with 72 fuel channels each containing 10 fuel slugs, would be done gradually. Reactor was not operated in September because of the secondary coolant pipes were exchanged between Danube and the horizontal sedimentary. Control and maintenance of the reactor equipment was done regularly and efficiently dependent on the availability of the spare parts. Difficulties in maintenance of the reactor instrumentation were caused by unavailability of the outdated spare parts

  15. Design of boron carbide-shielded irradiation channel of the outer irradiation channel of the Ghana Research Reactor-1 using MCNP.

    Science.gov (United States)

    Abrefah, R G; Sogbadji, R B M; Ampomah-Amoako, E; Birikorang, S A; Odoi, H C; Nyarko, B J B

    2011-01-01

    The MCNP model for the Ghana Research Reactor-1 was redesigned to incorporate a boron carbide-shielded irradiation channel in one of the outer irradiation channels. Extensive investigations were made before arriving at the final design of only one boron carbide covered outer irradiation channel; as all the other designs that were considered did not give desirable results of neutronic performance. The concept of redesigning a new MCNP model, which has a boron carbide-shielded channel is to equip the Ghana Research Reactor-1 with the means of performing efficient epithermal neutron activation analysis. After the simulation, a comparison of the results from the original MCNP model for the Ghana Research Reactor-1 and the new redesigned model of the boron carbide shielded channel was made. The final effective criticality of the original MCNP model for the GHARR-1 was recorded as 1.00402 while that of the new boron carbide designed model was recorded as 1.00282. Also, a final prompt neutron lifetime of 1.5245 × 10(-4)s was recorded for the new boron carbide designed model while a value of 1.5571 × 10(-7)s was recorded for the original MCNP design of the GHARR-1. Copyright © 2010 Elsevier Ltd. All rights reserved.

  16. Time-integrated thyroid dose for accidental releases from Pakistan Research Reactor-1

    International Nuclear Information System (INIS)

    Raza, S Shoaib; Iqbal, M; Salahuddin, A; Avila, R; Pervez, S

    2004-01-01

    The two-hourly time-integrated thyroid dose due to radio-iodines released to the atmosphere through the exhaust stack of Pakistan Research Reactor-1 (PARR-1), under accident conditions, has been calculated. A computer program, PAKRAD (which was developed under an IAEA research grant, PAK/RCA/8990), was used for the dose calculations. The sensitivity of the dose results to different exhaust flow rates and atmospheric stability classes was studied. The effect of assuming a constant activity concentration (as a function of time) within the containment air volume and an exponentially decreasing air concentration on the time-integrated dose was also studied for various flow rates (1000-50,000 m 3 h -1 ). The comparison indicated that the results were insensitive to the containment air exhaust rates up to or below 2000 m 3 h -1 , when the prediction with the constant activity concentration assumption was compared to an exponentially decreasing activity concentration model. The results also indicated that the plume touchdown distance increases with increasing atmospheric stability. (note)

  17. Computer codes developed in FRG to analyse hypothetical meltdown accidents

    International Nuclear Information System (INIS)

    Hassmann, K.; Hosemann, J.P.; Koerber, H.; Reineke, H.

    1978-01-01

    It is the purpose of this paper to give the status of all significant computer codes developed in the core melt-down project which is incorporated in the light water reactor safety research program of the Federal Ministry of Research and Technology. For standard pressurized water reactors, results of some computer codes will be presented, describing the course and the duration of the hypothetical core meltdown accident. (author)

  18. Regulatory Framework for Controlling the Research Reactor Decommissioning Project

    International Nuclear Information System (INIS)

    Melani, Ai; Chang, Soon Heung

    2009-01-01

    Decommissioning is one of important stages in construction and operation of research reactors. Currently, there are three research reactors operating in Indonesia. These reactors are operated by the National Nuclear Energy Agency (BATAN). The age of the three research reactors varies from 22 to 45 years since the reactors reached their first criticality. Regulatory control of the three reactors is conducted by the Nuclear Energy Regulatory Agency (BAPETEN). Controlling the reactors is carried out based on the Act No. 10/1997 on Nuclear Energy, Government Regulations and BAPETEN Chairman Decrees concerning the nuclear safety, security and safeguards. Nevertheless, BAPETEN still lack of the regulation, especially for controlling the decommissioning project. Therefore, in the near future BAPETEN has to prepare the regulations for decommissioning, particularly to anticipate the decommissioning of the oldest research reactors, which probably will be done in the next ten years. In this papers author give a list of regulations should be prepared by BAPETEN for the decommissioning stage of research reactor in Indonesia based on the international regulatory practice

  19. Research on reactor physics data

    International Nuclear Information System (INIS)

    1961-01-01

    In the early years of nuclear reactor research, each national program tended to develop its own reactor physics information. The Government of Norway proposed to the Agency the undertaking of a joint program in reactor physics utilizing the facilities and staff of its zero power reactor NORA then under construction. Following the approval by the Board of Governors in February, the Agency invited Member States to submit the names and qualifications of scientists they wished to suggest for the project. All the results and information gained through the program, which is expected to last about three years, will be placed at the disposal of the Agency's Member States

  20. Data acquisition and signal processing system for IPR R1 TRIGA-Mark I nuclear research reactor of CDTN

    International Nuclear Information System (INIS)

    Mesquita, A.Z.; Maretti, F. Jr.; Rezende, H.C.; Tambourgi, E.B.

    2004-01-01

    The TRIGA IPR-R1 Nuclear Research Reactor, located at the Nuclear Technology Development Center (CDTN/CNEN) in Belo Horizonte, Brazil, is being operated since 44 years ago. The main operational parameters were monitored by analog recorders and counters located in the reactor control console. The reactor operators registered the most important operational parameters and data in the reactor logbook. This process is quite useful, but it can involve some human errors. It is also impossible for the operators to take notes of all variables involving the process mainly during fast power transients in some operations. A PC-based data acquisition was developed for the reactor that allows online monitoring, through graphic interfaces, and shows operational parameters evolution to the operators. Some parameters that were not measured, like the power and the coolant flow rate at the primary loop, are monitored now in the computer video monitor. The developed system allows measuring out all parameters in a frequency up to 1 kHz. These data is also recorded in text files available for consults and analysis. (author)

  1. Applications of Research Reactors

    International Nuclear Information System (INIS)

    2014-01-01

    One of the IAEA's statutory objectives is to 'seek to accelerate and enlarge the contribution of atomic energy to peace, health and prosperity throughout the world.' One way this objective is achieved is through the publication of a range of technical series. Two of these are the IAEA Nuclear Energy Series and the IAEA Safety Standards Series. According to Article III.A.6 of the IAEA Statute, the safety standards establish 'standards of safety for protection of health and minimization of danger to life and property'. The safety standards include the Safety Fundamentals, Safety Requirements and Safety Guides. These standards are written primarily in a regulatory style, and are binding on the IAEA for its own programmes. The principal users are the regulatory bodies in Member States and other national authorities. The IAEA Nuclear Energy Series comprises reports designed to encourage and assist R and D on, and application of, nuclear energy for peaceful uses. This includes practical examples to be used by owners and operators of utilities in Member States, implementing organizations, academia, and government officials, among others. This information is presented in guides, reports on technology status and advances, and best practices for peaceful uses of nuclear energy based on inputs from international experts. The IAEA Nuclear Energy Series complements the IAEA Safety Standards Series. The purpose of the earlier publication, The Application of Research Reactors, IAEA-TECDOC-1234, was to present descriptions of the typical forms of research reactor use. The necessary criteria to enable an application to be performed were outlined for each one, and, in many cases, the minimum as well as the desirable requirements were given. This revision of the publication over a decade later maintains the original purpose and now specifically takes into account the changes in service requirements demanded by the relevant stakeholders. In particular, the significant improvements in

  2. Research report on the users' needs for next research reactor

    International Nuclear Information System (INIS)

    Takahashi, Hiroyuki; Tamura, Itaru; Hosoya, Toshiaki; Horiguchi, Hironori

    2015-03-01

    JRR-3 has been operated for more than 25 years for that it is time to investigate the role of a next research reactor. A task force under the Committee for Promotion of JRR-3 Neutron Beam Application has been organized by Department of Research Reactor and Tandem Accelerator to survey neutron beam application trends in the future. This is a report on the survey results and users' requirements for the next research reactor have been summarized in this report carried by the task force. (author)

  3. The determination of neutron energy spectrum in reactor core C1 of reactor VR-1 Sparrow

    Energy Technology Data Exchange (ETDEWEB)

    Vins, M. [Department of Nuclear Reactors, Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University, V Holesovickach 2, 180 00 Prague 8 (Czech Republic)], E-mail: vinsmiro@seznam.cz

    2008-07-15

    This contribution overviews neutron spectrum measurement, which was done on training reactor VR-1 Sparrow with a new nuclear fuel. Former nuclear fuel IRT-3M was changed for current nuclear fuel IRT-4M with lower enrichment of 235U (enrichment was reduced from former 36% to 20%) in terms of Reduced Enrichment for Research and Test Reactors (RERTR) Program. Neutron spectrum measurement was obtained by irradiation of activation foils at the end of pipe of rabit system and consecutive deconvolution of obtained saturated activities. Deconvolution was performed by computer iterative code SAND-II with 620 groups' structure. All gamma measurements were performed on Canberra HPGe. Activation foils were chosen according physical and nuclear parameters from the set of certificated foils. The Resulting differential flux at the end of pipe of rabit system agreed well with typical spectrum of light water reactor. Measurement of neutron spectrum has brought better knowledge about new reactor core C1 and improved methodology of activation measurement. (author)

  4. IAEA Activities supporting education and training at research reactors

    International Nuclear Information System (INIS)

    Peld, N.D.; Ridikas, D.

    2013-01-01

    Full-text: Through the provision of neutrons for experiments and their historical association with universities, research reactors have played a prominent role in nuclear education and training of students, scientists and radiation workers. Today education and training remains the foremost application of research reactors, involving close to 160 facilities out of 246 operational. As part of its mandate to facilitate and expand the contribution of atomic energy to peace, health and prosperity throughout the world, the IAEA administers a number of activities intended to promote nuclear research and enable access to nuclear technology for peaceful purposes, one of which is the support of various education and training measures involving research reactors. In the last 5 years, education and training has formed one pillar for the creation of research reactor coalitions and networks to pool their resources and offer joint programmes, such as the on-going Group Fellowship Training Course. Conducted mainly through the Eastern European Research Reactor Initiative, this programme is a periodic sic week course for young scientists and engineers on nuclear techniques and administration jointly conducted at several member research reactor institutes. Organization of similar courses is under consideration in Latin America and the Asia-Pacific Region, also with support from the IAEA. Additionally, four research reactor institutes have begun offering practical education courses through virtual reactor experiments and operation known as the Internet Reactor Laboratory. Through little more than an internet connection and projection screens, university science departments can be connected regionally or bilaterally with the control room o a research reactor for various training activities. Finally, two publications are being prepared, namely Hands-On Training Courses Using Research Reactors and Accelerators, and Compendium on Education and training Based on Research Reactors. These

  5. Generic component reliability data for research reactors PSA. Final report of the CRP on data acquisition for research reactor PSA. Working material

    International Nuclear Information System (INIS)

    1993-01-01

    The scope of this document is to provide the final reference generic component reliability database information for a variety of research reactor types. As noted in Section 2.1 and Table 3a, many years of component data are represented in the database so that it is expected that the report should provide representative data valid for a number of years. The database provides component failure rates on a time and/or a demand related basis according to the operational modes of the components. At the current time an update of the database is not planned. As a result of the implementation of data collection systems in the research reactors represented in these studies, updating of data from individual facilities could be made available from the contributing research reactor facilities themselves. As noted in Section 1.2, the report does not include detailed discussion of information regarding component classification and reliability parameter definitions. The report does provide some insights and discussion regarding the practicalities of the data collection process and some guidelines for database usage. 9 refs, tabs

  6. Generic component reliability data for research reactors PSA. Final report of the CRP on data acquisition for research reactor PSA. Working material

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-12-31

    The scope of this document is to provide the final reference generic component reliability database information for a variety of research reactor types. As noted in Section 2.1 and Table 3a, many years of component data are represented in the database so that it is expected that the report should provide representative data valid for a number of years. The database provides component failure rates on a time and/or a demand related basis according to the operational modes of the components. At the current time an update of the database is not planned. As a result of the implementation of data collection systems in the research reactors represented in these studies, updating of data from individual facilities could be made available from the contributing research reactor facilities themselves. As noted in Section 1.2, the report does not include detailed discussion of information regarding component classification and reliability parameter definitions. The report does provide some insights and discussion regarding the practicalities of the data collection process and some guidelines for database usage. 9 refs, tabs.

  7. The FRJ-1 (MERLIN) research reactor: its main activity inventory has been removed by successful demolition of the reactor block

    International Nuclear Information System (INIS)

    Stahn, B.; Printz, R.; Matela, K.; Zehbe, C.; Poeppinghaus, J.; Cremer, J.

    2004-01-01

    The FRJ-1 (MERLIN) research reactor was decommissioned in 1985 after twenty-three years of operation. Demolition of the plant was begun in 1996. The article contains a survey of the demolition steps carried out so far within the framework of three partial permits. The main activity is the demolition of the reactor core structures as a precondition for subsequent measures to ensure clearance measurements of the building. The core structures are demolished which were exposed to high neutron fluxes during reactor operation and now show the highest activity and dose rate levels, except for the core internals. For demolition and disassembly of the metal structures in this part of the plant, the tools specially designed and made include a remotely operated sawing system and a pipe cutting system for internal segmentation of the beam lines. The universal demolition tool for use also above and beyond the concrete structures has been found to be a remotely controlled electrohydraulic demolition shovel. Spreading contamination in the course of the demolition work was avoided. One major reason for this success was the fact that no major airborne contamination existed at any time as a consequence of the quality of the material demolished and also of the consistent use of technical tools. While the reactor block was being demolished, an application for clearance measurement of the reactor hall and subsequent release from the scope of the Atomic Energy Act was filed as early as in mid-2003. The fourth partial permit covering these activities is expected to be issued in the spring of 2004. (orig.)

  8. Operation and utilizations of Dalat nuclear research reactor

    International Nuclear Information System (INIS)

    Hien, P.Z.

    1988-01-01

    The reconstructed Dalat nuclear research reactor was commissioned in March 1984 and up to September 1988 more than 6200 hours of operation at nominal power have been recorded. The major utilizations of the reactor include radioisotope production, activation analysis, nuclear data research and training. A brief review of the utilizations of the reactor is presented. Some aspects of reactor safety are also discussed. (author)

  9. Management and storage of nuclear fuel from Belgian research reactors

    International Nuclear Information System (INIS)

    Gubel, P.

    1996-01-01

    Experiences and problems with the storage of irradiated fuel at research reactors in Belgium are described. In particular, interim storage problems exist for spent fuel elements at the BR2 and the shut down BR3 reactors in Mol. (author). 1 ref

  10. Status of research reactor spent fuel world-wide

    International Nuclear Information System (INIS)

    Ritchie, I.G.

    2004-01-01

    Results compiled in the research reactor spent fuel database are used to assess the status of research reactor spent fuel world-wide. Fuel assemblies, their types, enrichment, origin of enrichment and geological distribution among the industrialised and developed countries of the world are discussed. Fuel management practices in wet and dry storage facilities and the concerns of reactor operators about long-term storage of their spent fuel are presented and some of the activities carried out by the International Atomic Energy Agency to address the issues associated with research reactor spent fuel are outlined. (author)

  11. Research reactor status for future nuclear research in Europe

    Energy Technology Data Exchange (ETDEWEB)

    Raymond, Patrick; Bignan, Gilles; Guidez, Joel [Commissariat a l' Energie Atomique - CEA (France)

    2010-07-01

    During the 1950's and 60's, the European countries built several research reactors, partially to support their emerging nuclear-powered electricity programs. Now, over forty years later, the use and operation of these reactors have both widened and grown more specialized. The irradiation reactors test materials and fuels for power reactors, produce radio-isotopes for medicine, neutro-graphies, doping silicon, and other materials. The neutron beam reactors are crucial to science of matter and provide vital support to the development of nano-technologies. Other reactors are used for other specialized services such as teaching, safety tests, neutron physics measurements... The modifications to the operating uses and the ageing of the nuclear facilities have led to increasing closures year after year. Since last ENC, for example, we have seen, only in France, the closure of the training reactor Ulysse in 2007, the closure of the safety test dedicated reactor Phebus in 2008 and recently the Phenix reactor, last fast breeder in operation in the European Community, has been shut down after a set of 'end of life' technological and physical tests. For other research reactors, safety re-evaluations have had to take place, to enable extension of reactor life. However, in the current context of streamlining and reorganization, new European tools have emerged to optimally meet the changing demands for research. However the operation market of these reactors seems now increasing in all fields. For the neutron beams reactors (FRMII, ORPHEE, ILL, ISIS,..) the experimental needs are increasing years after years, especially for nano sciences and bio sciences new needs. The measurement of residual stress on manufactured materials is also more and more utilised. All these reactors have increasing utilizations, and their future seems promising. A new project project based on a neutron spallation is under definition in Sweden (ESSS: European Spallation Source

  12. Research reactor status for future nuclear research in Europe

    International Nuclear Information System (INIS)

    Raymond, Patrick; Bignan, Gilles; Guidez, Joel

    2010-01-01

    During the 1950's and 60's, the European countries built several research reactors, partially to support their emerging nuclear-powered electricity programs. Now, over forty years later, the use and operation of these reactors have both widened and grown more specialized. The irradiation reactors test materials and fuels for power reactors, produce radio-isotopes for medicine, neutro-graphies, doping silicon, and other materials. The neutron beam reactors are crucial to science of matter and provide vital support to the development of nano-technologies. Other reactors are used for other specialized services such as teaching, safety tests, neutron physics measurements... The modifications to the operating uses and the ageing of the nuclear facilities have led to increasing closures year after year. Since last ENC, for example, we have seen, only in France, the closure of the training reactor Ulysse in 2007, the closure of the safety test dedicated reactor Phebus in 2008 and recently the Phenix reactor, last fast breeder in operation in the European Community, has been shut down after a set of 'end of life' technological and physical tests. For other research reactors, safety re-evaluations have had to take place, to enable extension of reactor life. However, in the current context of streamlining and reorganization, new European tools have emerged to optimally meet the changing demands for research. However the operation market of these reactors seems now increasing in all fields. For the neutron beams reactors (FRMII, ORPHEE, ILL, ISIS,..) the experimental needs are increasing years after years, especially for nano sciences and bio sciences new needs. The measurement of residual stress on manufactured materials is also more and more utilised. All these reactors have increasing utilizations, and their future seems promising. A new project project based on a neutron spallation is under definition in Sweden (ESSS: European Spallation Source Scandinavia). The nuclear

  13. Proceedings of the FNCA 2003 workshop on the utilization of research reactors (Contract research)

    International Nuclear Information System (INIS)

    2006-03-01

    The FNCA 2003 Workshop on the Utilization of Research Reactors, which is the twelfth workshop on the theme of research reactor utilization, was held in Dalat, Vietnam and Jakarta and Serpong, Indonesia from January 12 to 16, 2004. This workshop was executed based on the agreement in the fourth Coordinator's Meeting of Forum for Nuclear Cooperation in Asia (FNCA) held in Tokyo, March 2003. The workshop consisted of four groups under the theme of the following fields; 1) Neutron Activation Analysis, 2) Research Reactors, 3) Tc-99m Generator Technology and 4) Neutron Scattering. The total number of participants for the workshop was 93 people from 8 countries; China, Indonesia, Korea, Malaysia, the Philippines, Thailand, Vietnam and Japan. The 30 of the presented papers are indexed individually. (J.P.N.)

  14. Thorium-based nuclear fuel: current status and perspectives

    International Nuclear Information System (INIS)

    1987-03-01

    Until the present time considerable efforts have already been made in the area of fabrication, utilization and reprocessing of Th-based fuels for different types of reactors, namely: by FRG and USA - for HTRs; FRG and Brazil, Italy - for LWRs; India - for HWRs and FBRs. Basic research of thorium fuels and thorium fuel cycles are also being undertaken by Australia, Canada, China, France, FRG, Romania, USSR and other countries. Main emphasis has been given to the utilization of thorium fuels in once-through nuclear fuel cycles, but in some projects closed thorium-uranium or thorium-plutonium fuel cycles are also considered. The purpose of the Technical Committee on the Utilization of Thorium-Based Nuclear Fuel: Current Status and Perspective was to review the world thorium resources, incentives for further exploration, obtained experience in the utilization of Th-based fuels in different types of reactors, basic research, fabrication and reprocessing of Th-based fuels. As a result of the panel discussion the recommendations on future Agency activities and list of major worldwide activities in the area of Th-based fuel were developed. A separate abstract was prepared for each of the 9 papers in this proceedings series

  15. Reactor Engineering Department annual report (April 1, 1988 - March 31, 1989)

    International Nuclear Information System (INIS)

    1989-09-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1988 (April 1, 1988 - March 31, 1989). The Department has promoted cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and also to PNC's fast reactor project. Other major Department's programs are the assessment of the high conversion light water reactor and the design activities of advanced reactor system. Application of a high energy accelerator to the nuclear engineering is also preliminarily assessed. The report also contains the latest progress in various basic researches as nuclear data and group constants, theoretical methods and code development, reactor physics experiments and analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/ diagnosis and technical developments related to the reactor physics facilities. The activities of the Research Committee on Reactor Physics are also summarized. (author)

  16. International collaboration between nuclear research centres and the role of research reactors

    International Nuclear Information System (INIS)

    Dodd, B.

    2001-01-01

    A research reactor is a core facility in many nuclear research centres (NRCs) of Member States and it is logical that it should be the focus of any international collaboration between such centres. There are several large and sophisticated research reactors in operation in both developed and developing Member States, such as Belgium, China, Egypt, France, Hungary, Indonesia, India, Japan, ROK, Netherlands, South Africa and the USA. There are also several new, large reactors under construction or being planned such as those in Australia, Canada, China, France, Germany, and Thailand. It is felt that the utilization of these reactors can be enhanced by international co-operation to achieve common goals in research and applications. (author)

  17. Simulation of a pool type research reactor

    International Nuclear Information System (INIS)

    Oliveira, Andre Felipe da Silva de; Moreira, Maria de Lourdes

    2011-01-01

    Computational fluid dynamic is used to simulate natural circulation condition after a research reactor shutdown. A benchmark problem was used to test the viability of usage such code to simulate the reactor model. A model which contains the core, the pool, the reflector tank, the circulation pipes and chimney was simulated. The reactor core contained in the full scale model was represented by a porous media. The parameters of porous media were obtained from a separate CFD analysis of the full core model. Results demonstrate that such studies can be carried out for research and test of reactors design. (author)

  18. Transient behaviour study program of research reactors fuel elements at the Hydra Pulse Reactor

    International Nuclear Information System (INIS)

    Khvostionov, V.E.; Egorenkov, P.M.; Malankin, P.V.

    2004-01-01

    Program on behavior study of research reactor Fuel Elements (FE) under transient regimes initiated by excessive reactivity insertion is being presented. Program would be realized at HYDRA pulse reactor at Russian Research Center 'Kurchatov Institute' (RRC 'K1'). HYDRA uses aqueous solution of uranyl sulfate (UO 2 SO 4 ) as a fuel. Up to 30 MJ of energy can be released inside the core during the single pulse, effective power pulse width varying from 2 to 10 ms. Reactor facility allows to investigate behaviour of FE consisting of different types of fuel composition, being developed according to Russian RERTR. First part of program is aimed at transient behaviour studying of FE MR, IRT-3M, WWR-M5 types containing meats based on dioxide uranium in aluminum matrix. Mentioned FEs use 90% and 36% enriched uranium. (author)

  19. Status report of Indonesian research reactor

    International Nuclear Information System (INIS)

    Arbie, B.; Supadi, S.

    1992-01-01

    A general description of three Indonesian research reactor, its irradiation facilities and its future prospect are described. Since 1965 Triga Mark II 250 KW Bandung, has been in operation and in 1972 the design powers were increased to 1000 KW. Using core grid form Triga 250 KW BATAN has designed and constructed Kartini Reactor in Yogyakarta which started its operation in 1979. Both of this Triga type reactors have served a wide spectrum of utilization such as training manpower in nuclear engineering, radiochemistry, isotope production and beam research in solid state physics. Each of this reactor have strong cooperation with Bandung Institute of Technology at Bandung and Gajah Mada University at Yogyakarta which has a faculty of Nuclear Engineering. Since 1976 the idea to have high flux reactor has been foreseen appropriate to Indonesian intention to prepare infrastructure for nuclear industry for both energy and non-energy related activities. The idea come to realization with the first criticality of RSG-GAS (Multipurpose Reactor G.A. Siwabessy) in July 1987 at PUSPIPTEK Serpong area. It is expected that by early 1992 the reactor will reached its full power of 30 MW and by end 1992 its expected that irradiation facilities will be utilized in the future for nuclear scientific and engineering work. (author)

  20. Technical and safe development features of modern research reactor

    International Nuclear Information System (INIS)

    Wang Jiaying; Dong Duo

    1998-01-01

    The development trend of research reactor in the world, and development situation in China are introduced. Up to now, some research reactors have serviced for long time and equipment have aged, not to be satisfied for requirement of science and technology development. New research reactors must been developed. The technical features and safe features of new type research reactor in China, for example: multi-pile utilization, compact core of high flux, high automation level of control, reactor two independent shutdown systems, great coefficient of negative temperature, passive safety systems, reliable residual heat removal system are studied

  1. Homogeneous SLOWPOKE reactors for replacing SLOWPOKE-2 research reactors and the production of radioisotopes

    International Nuclear Information System (INIS)

    Bonin, H.W.; Hilborn, J.W.; Carlin, G.E.; Gagnon, R.; Busatta, P.

    2014-01-01

    Inspired from the inherently safe SLOWPOKE-2 research reactor, the Homogeneous SLOWPOKE reactor was conceived with a double goal: replacing the heterogeneous SLOWPOKE-2 reactors when they reach end-of-core life to continue their missions of neutron activation analysis and neutron radiography at universities, and to produce radioisotopes such as 99 Mo for medical applications. A homogeneous reactor core allows a much simpler extraction of radioisotopes (such as 99 Mo) for applications in industry and nuclear medicine. The 20 kW Homogeneous SLOWPOKE reactor was modelled using both the deterministic WIMS-AECL and the probabilistic MCNP 5 reactor simulation codes. The homogeneous fuel mixture was a dilute aqueous solution of Uranyl Sulfate (UO 2 SO 4 ) with 994.2 g of 235 U (enrichment at 20%) providing an excess reactivity at operating temperature (40 o C) of 3.8 mk for a molality determined as 1.46 mol kg -1 for a Zircaloy-2 reactor vessel. Because this reactor is intended to replace the core of SLOWPOKE-2 reactors, the Homogeneous SLOWPOKE reactor core had a height about twice its diameter. The reactor could be controlled by mechanical absorber rods in the beryllium reflector, chemical control in the core, or a combination of both. The safety of the Homogeneous SLOWPOKE reactor was analysed for both normal operation and transient conditions. Thermal-hydraulics calculations used COMSOL Multiphysics and the results showed that natural convection was sufficient to ensure adequate reactor cooling in all situations. The most severe transient simulated resulted from a 5.87 mk step positive reactivity insertion to the reactor in operation at critical and at steady state at 20 o C. Peak temperature and power were determined as 83 o C and 546 kW, respectively, reached 5.1 s after the reactivity insertion. However, the power fell rapidly to values below 20 kW some 35 s after the peak and remained below that value thereafter. Both the temperature and void coefficients are

  2. Homogeneous SLOWPOKE reactors for replacing SLOWPOKE-2 research reactors and the production of radioisotopes

    Energy Technology Data Exchange (ETDEWEB)

    Bonin, H.W., E-mail: bonin-h@rmc.ca [Royal Military College of Canada, Kingston, Ontario (Canada); Hilborn, J.W. [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada); Carlin, G.E. [Ontario Power Generation, Toronto, Ontario (Canada); Gagnon, R.; Busatta, P. [Canadian Forces (Canada)

    2014-07-01

    Inspired from the inherently safe SLOWPOKE-2 research reactor, the Homogeneous SLOWPOKE reactor was conceived with a double goal: replacing the heterogeneous SLOWPOKE-2 reactors when they reach end-of-core life to continue their missions of neutron activation analysis and neutron radiography at universities, and to produce radioisotopes such as {sup 99}Mo for medical applications. A homogeneous reactor core allows a much simpler extraction of radioisotopes (such as {sup 99}Mo) for applications in industry and nuclear medicine. The 20 kW Homogeneous SLOWPOKE reactor was modelled using both the deterministic WIMS-AECL and the probabilistic MCNP 5 reactor simulation codes. The homogeneous fuel mixture was a dilute aqueous solution of Uranyl Sulfate (UO{sub 2}SO{sub 4}) with 994.2 g of {sup 235}U (enrichment at 20%) providing an excess reactivity at operating temperature (40 {sup o}C) of 3.8 mk for a molality determined as 1.46 mol kg{sup -1} for a Zircaloy-2 reactor vessel. Because this reactor is intended to replace the core of SLOWPOKE-2 reactors, the Homogeneous SLOWPOKE reactor core had a height about twice its diameter. The reactor could be controlled by mechanical absorber rods in the beryllium reflector, chemical control in the core, or a combination of both. The safety of the Homogeneous SLOWPOKE reactor was analysed for both normal operation and transient conditions. Thermal-hydraulics calculations used COMSOL Multiphysics and the results showed that natural convection was sufficient to ensure adequate reactor cooling in all situations. The most severe transient simulated resulted from a 5.87 mk step positive reactivity insertion to the reactor in operation at critical and at steady state at 20 {sup o}C. Peak temperature and power were determined as 83 {sup o}C and 546 kW, respectively, reached 5.1 s after the reactivity insertion. However, the power fell rapidly to values below 20 kW some 35 s after the peak and remained below that value thereafter. Both the

  3. Progress of nuclear safety research, (1)

    International Nuclear Information System (INIS)

    Amano, Hiroshi; Nakamura, Hiroei; Nozawa, Masao

    1981-01-01

    The Japan Atomic Energy Research Institute was established in 1956 in conformity with the national policy to extensively conduct the research associated with nuclear energy. Since then, the research on nuclear energy safety has been conducted. In 1978, the Division of Reactor Safety was organized to conduct the large research programs with large scale test facilities. Thereafter, the Divisions of Reactor Safety Evaluation, Environmental Safety Research and Reactor Fuel Examination were organized successively in the Reactor Safety Research Center. The subjects of research have ranged from the safety of nuclear reactors to that in the recycling of nuclear fuel. In this pamphlet, the activities in JAERI associated with the safety research are reported, which have been carried out in the past two years. Also, the international cooperation research program in which JAERI participated is included. This pamphlet consists of two parts, and in this Part 1, the reactor safety research is described. The safety of nuclear fuel, the integrity and safety of pressure boundary components, the engineered safety in LOCA, fuel behavior in accident and others are reported. (Kako, I.)

  4. Training and Certification of Research Reactor Personnel

    International Nuclear Information System (INIS)

    Zarina Masood

    2011-01-01

    The safe operation of a research reactor requires that reactor personnel be fully trained and certified by the relevant authorities. Reactor operators at PUSPATI TRIGA Reactor underwent extensive training and are certified, ever since the reactor first started its operation in 1982. With the emphasis on enhancing reactor safety in recent years, reactor operator training and certification have also evolved. This paper discusses the changes that have to be implemented and the challenges encountered in developing a new training programme to be in line with the national standards. (author)

  5. Worldwide Report, Telecommunications Policy, Research and Development.

    Science.gov (United States)

    1984-02-22

    AFFAIRS Briefs Siemens Electronics to Turkey 62 FEDERAL REPUBLIC OF GERMANY FRG Awards Contracts for Integrated Digital Network (VDI NACHRICHTEN ...NETWORK Duesseldorf VDI NACHRICHTEN in German 23 Dec 83 p 1 [Article: "FRG’s Own Radio Satellite"] [Text] The German Federal Postal Service (DBP

  6. IEA-R1 reactor - Spent fuel management

    International Nuclear Information System (INIS)

    Mattos, J.R.L. De

    1996-01-01

    Brazil currently has one Swimming Pool Research Reactor (IEA-R1) at the Instituto de Pesquisas Energeticas e Nucleares - Sao Paulo. The spent fuel produced is stored both at the Reactor Pool Storage Compartment and at the Dry Well System. The present situation and future plans for spent fuel storage are described. (author). 3 refs, 2 figs, 2 tabs

  7. Conceptual design of multipurpose compact research reactor

    International Nuclear Information System (INIS)

    Nagata, Hiroshi; Kusunoki, Tsuyoshi; Hori, Naohiko; Kaminaga, Masanori

    2012-01-01

    Conceptual design of the high-performance and low-cost multipurpose compact research reactor which will be expected to construct in the nuclear power plant introduction countries, started from 2010 in JAEA and nuclear-related companies in Japan. The aims of this conceptual design are to achieve highly safe reactor, economical design, high availability factor and advanced irradiation utilization. One of the basic reactor concept was determined as swimming pool type, thermal power of 10MW and water cooled and moderated reactor with plate type fuel element same as the JMTR. It is expected that the research reactors are used for human resource development, progress of the science and technology, expansion of industry use, lifetime extension of LWRs and so on. (author)

  8. Preservation of the first research nuclear reactor in Korea

    International Nuclear Information System (INIS)

    2008-06-01

    This book describes preservation of the first research nuclear reactor in Korea and necessity of building memorial hall, sale of the Institute of Atomic Energy Research in Seoul and dismantlement of the first and the second nuclear reactor, preservation of the first research nuclear reactor and activity about memorial hall of the atomic energy reactor, assignment and leaving the report, and the list of related data.

  9. Research reactor programmes at the IAEA

    International Nuclear Information System (INIS)

    Reijonen, H.

    1978-01-01

    The activities performed according to the Agency programs for research reactors in the fields of information collection and dissemination, meetings organization, publications of the proceedings and execution of technical assistance are discussed in the paper emphasizing the services that are provided for developing countries. It is intended that the programme on research reactors should be flexible and respond to the actual needs of the countries receiving assistance

  10. Development and investigation of the prestressed reinforced concrete vessels for the water cooled reactors in the FRG

    International Nuclear Information System (INIS)

    Medovikov, A.I.; Bogopol'skij, V.G.; Nikolaev, Yu.B.; Konevskij, V.N.

    1980-01-01

    An analysis of calculation results for characteristics of stress-strained state of reactor vessel made of prestressed reinforced concrete is presented. Experimental data obtained during the investigation into a model of reactor vessel top cover are given. Thermal shielding system both for boiling water and pressurized-water reactors has been considered and its working capacity has been evaluated. An analysis of experimental data show correctness of the method assumed for calculation of the reactor top cover which permits to exactly determine its stressed-strained state as well as the nature of crack propagation in the vessel and the structure supporting power. Ceramics is suggested to be used as a heat-insulating material

  11. RA Research reactor, Annual report 1988

    International Nuclear Information System (INIS)

    Sotic, O.

    1988-12-01

    Annual report concerning the project 'RA research nuclear reactor' for 1989, financed by the Serbian ministry of science is divided into two parts. First part is concerned with RA reactor operation and maintenance, which is the task of the Division for reactor engineering of the Institute for multidisciplinary studies and RA reactor engineering. Second part deals with radiation protection activities at the RA reactor which is the responsibility of the Institute for radiation protection. Scientific council of the Institute for multidisciplinary studies and RA reactor engineering has stated that this report describes adequately the activity and tasks fulfilled at the RA reactor in 1989. The scope and the quality of the work done were considered successful both concerning the maintenance and reconstruction, as well as radiation protection activities [sr

  12. Thirteenth water reactor safety research information meeting: proceedings Volume 1

    International Nuclear Information System (INIS)

    Weiss, A.J.

    1986-02-01

    This six-volume report contains 151 papers out of the 178 that were presented at the Thirteenth Water Reactor Safety Research Information Meeting held at the National Bureau of Standards, Gaithersburg, Maryland, during the week of October 22-25, 1985. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included thirty-one different papers presented by researchers from Japan, Canada and eight European countries. The title of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. This volume presents information on: risk analysis PRA application; severe accident sequence analysis; risk analysis/dependent failure analysis; and industry safety research

  13. IGORR 2: Proceedings of the 2. meeting of the International Group On Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1992-07-01

    The International group on Research Reactors was formed to facilitate the sharing of knowledge and experience among those institutions and individuals who are actively working to design, build, and promote new research reactors or to make significant upgrades to existing facilities. Sessions during this second meeting were devoted to research reactor reports (GRENOBLE reactor, FRM-II, HIFAR, PIK, reactors at JAERI, MAPLE, ANS, NIST, MURR, TRIGA, BR-2, SIRIUS 2); other neutron sources; and two workshops were dealing with research and development results and needs and reports on progress in needed of R and D areas identified at IGORR 1.

  14. IGORR 2: Proceedings of the 2. meeting of the International Group On Research Reactors

    International Nuclear Information System (INIS)

    1992-01-01

    The International group on Research Reactors was formed to facilitate the sharing of knowledge and experience among those institutions and individuals who are actively working to design, build, and promote new research reactors or to make significant upgrades to existing facilities. Sessions during this second meeting were devoted to research reactor reports (GRENOBLE reactor, FRM-II, HIFAR, PIK, reactors at JAERI, MAPLE, ANS, NIST, MURR, TRIGA, BR-2, SIRIUS 2); other neutron sources; and two workshops were dealing with research and development results and needs and reports on progress in needed of R and D areas identified at IGORR 1

  15. Justify of implementation of a hot water layer system in swimming pool research reactor IEA-R1m

    International Nuclear Information System (INIS)

    Toyoda, Eduardo Yoshio; Gordon, Ana Maria Pinho Leite; Sordi, Gian-Maria A.A.

    2001-01-01

    The IPEN/CNEN-SP has a swimming pool research reactor (IEA-R1m) in operation since 1957 at 2 MW. In 1998, after some modifications, its nominal power increased to 5 MW. Among these modifications some adaptations had to be accomplished in the radiological protection and operational procedure. The present work aim to study the need of implementation of a hot water layer in order to reduce the dose in the workers in the vicinity of the reactor swimming pool. Applying the principles of radioprotection optimization, it was concluded that the decision of the construction of one hot water layer system in the reactor swimming pool, is not necessary. (author)

  16. Strategic Planning for Research Reactors

    International Nuclear Information System (INIS)

    2017-01-01

    This publication is a revision of IAEA-TECDOC-1212 which primarily focused on enhancing the utilization of existing research reactors. This updated version also provides guidance on how to develop and implement a strategic plan for a new research reactor project and will be of particular interest for organizations which are preparing a feasibility study to establish such a new facility. This publication will enable managers to determine more accurately the actual and potential capabilities of an existing reactor, or the intended purpose and type of a new facility. At the same time, management will be able to match these capabilities to stakeholders/users’ needs and establish the strategy of meeting such needs. In addition, several annexes are presented, including some examples as clarification to the main text and ready-to-use templates as assistance to the team drafting a strategic plan.

  17. Nuclear research reactors in the world. May 1987 ed.

    International Nuclear Information System (INIS)

    1987-01-01

    This is the second edition of Reference Data Series No.3, Nuclear Research Reactors in the World, which replaces the Agency's publications Power and Research Reactors in Member States and Research Reactors in Member States. This booklet contains general information, as of the end of May 1987, on research reactors in operation, under construction, planned, and shut down. The information is collected by the Agency through questionnaires sent to the Member States through the designated national correspondents. 11 figs, 19 tabs

  18. Self Assessment for the Safety of Research Reactor in Indonesia

    International Nuclear Information System (INIS)

    Melani, Ai; Chang, Soon Heung

    2008-01-01

    At the present Indonesia has no nuclear power plant in operation yet, although it is expected that the first nuclear power plant will be operated and commercially available in around the year of 2016 to 2017 in Muria Peninsula. National Nuclear Energy Agency (BATAN) has three research reactor; which are: Reactor Triga Mark II at Bandung, Reactor Kartini at Yogyakarta and Reactor Serbaguna (Multi Purpose Reactor) at Serpong. The Code of Conduct on the Safety of Research Reactors establishes 'best practice' guidelines for the licensing, construction and operation of research reactors. In this paper the author use the requirement in code of conduct to review the safety of research reactor in Indonesia

  19. Neutron Spectrum Parameters In Inner Irradiation Channel Of The Nigeria Research Reactor-1 (NIRR-1) For Use In Absolute And KO-NAA Methods

    International Nuclear Information System (INIS)

    Jonah, S.A; Balogun, G.I; Mayaki, M.C.

    2004-01-01

    In Nigeria, the first Nuclear Reactor achieved critically on February 03, 2004 at about 11:35 GMT and has been commissioned or training and research. It is a Miniature Neutron Source Reactor (MNSR), code-named Nigeria Research Reactor-1 (NIRR-1). NIRR-1 has a tan-in-pool structural configuration and a nominal thermal power rating of 30 Kw. With a built-in clean old core excess reactivity of 3.77 mk determined during the on-site zero and critically experimental, the reactor can operate for a n.cm-2 .s-1 in the inner irradiation channels). Under these conditions, the reactor can operate with the same fuel loading for over ten years with a burn-up of <1%. A detailed description of operating characteristics for NIRR-1, measured during the on-site zero-power and criticality experiments has been given elsewhere. In order to extend its utilization to include absolute and ko-NAA methods, the neutron spectrum parameters in the irradiation channels: power and critically experiments has been given elsewhere. In order to extend it's the irradiation channels: thermal-to-epithermal flux ration, F; and epithermal flux shape factor, a in both the inner and outer irradiation channels must be determined experimentally. In this work, we have developed and experimental procedure for monitoring the neutron spectrum parameters in an inner irradiation channel based on irradiation and gamma-ray counting of detector foils via (n,y), (n,p) and (n,a) dosimetry reactions. Results obtained indicate that a thermal neutron flux of (5.14+-0.02) x 1011 n/c m2.s determined by foil activation method in the inner irradiation channel, B2, at a power level of 15.5 kw corresponds to the flux indicators on the control console and the micro-computer control system respectively. Other parameters of the neutron spectrum determined for inner irradiation channel B2, are: a -0.0502+0.003; 18.92+-0.14; F = 3.87=0.23. The method was validated through the comparison of our result with published neutron spectrum

  20. 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

  1. Reactor Engineering Department annual report (April 1, 1990 - March 31, 1991)

    International Nuclear Information System (INIS)

    1991-09-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1990 (April 1, 1990 - March 31, 1991). The major Department's programs promoted in the year are the assessment of the high conversion light water reactor, the design activities of advanced reactor system and development of a high energy proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics, technology assessment of nuclear energy and technology developments related to the reactor physics facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project also progressed. The activities of the Research Committee on Reactor Physics are also summarized. (author)

  2. Reactor Engineering Department annual report (April 1, 1991-March 31, 1992)

    International Nuclear Information System (INIS)

    1992-08-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1991 (April 1, 1991-March 31, 1992). The major Department's programs promoted in the year are assessment of the high conversion light water reactor, the design activities of advanced reactor system and development of a high energy proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researchers on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics, technology assessment of nuclear energy and technology developments related to the reactor physics facilities. The cooperative work to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project also progressed. The activities of the Research Committee on Reactor Physics are also summarized. (author)

  3. Standards for safe operation of research reactors

    International Nuclear Information System (INIS)

    1996-01-01

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

  4. Design and research of fuel element for pulsed reactor

    International Nuclear Information System (INIS)

    Tian Sheng

    1994-05-01

    The fuel element is the key component for pulsed reactor and its design is one of kernel techniques for pulsed reactor. Following the GA Company of US the NPIC (Nuclear Power Institute of China) has mastered this technique. Up to now, the first pulsed reactor in China (PRC-1) has been safely operated for about 3 years. The design and research of fuel element undertaken by NPIC is summarized. The verification and evaluation of this design has been carried out by using the results of measured parameters during operation and test of PRC-1 as well as comparing the design parameters published by others

  5. Future directions of small research reactors

    International Nuclear Information System (INIS)

    Blotcky, A.J.; Rack, E.P.

    1986-01-01

    In prognosticating future perspectives, it is important to realize that the current number of small reactors throughout the world is not overly large and will undoubtedly decrease or at best remain constant in future generations. To survive and remain productive, small reactor facilities must concentrate on work that is unique and that cannot be performed as well by other instruments. Wherever possible, these facilities should develop some form of collaboration with universities and medical center investigators. Future development will continue and will flourish in neutron activation analysis and its applications for a diversity of fields. Fundamental research such as hot atom chemistry will continue to use neutrons from small research reactors. Finally, training of power reactor operators can be an important source of revenue for the small facility in addition to performing an important service to the nuclear power industry

  6. Review of the status of low power research reactors and considerations for its development

    International Nuclear Information System (INIS)

    Lim, In Cheol; Wu, Sang Ik; Lee, Byung Chul; Ha, Jae Joo

    2012-01-01

    At present, 232 research reactors in the world are in operation and two thirds of them have a power less than 1 MW. Many countries have used research reactors as the tools for educating and training students or engineers and for scientific service such as neutron activation analysis. As the introduction of a research reactor is considered a stepping stone for a nuclear power development program, many newcomers are considering having a low power research reactor. The IAEA has continued to provide forums for the exchange of information and experiences regarding low power research reactors. Considering these, the Agency is recently working on the preparation of a guide for the preparation of technical specification possibly for a member state to use when wanting to purchase a low power research reactor. In addition, ANS has stated that special consideration should be given to the continued national support to maintain and expand research and test reactor programs and to the efforts in identifying and addressing the future needs by working toward the development and deployment of next generation nuclear research and training facilities. Thus, more interest will be given to low power research reactors and its role as a facility for education and training. Considering these, the status of low power research reactors was reviewed, and some aspects to be considered in developing a low power research reactor were studied

  7. Manual for the operation of research reactors

    International Nuclear Information System (INIS)

    1965-01-01

    The great majority of the research reactors in newly established centres are light-water cooled and are often also light-water moderated. Consequently, the IAEA has decided to publish in its Technical Reports Series a manual dealing with the technical and practical problems associated with the safe and efficient operation of this type of reactor. Even though this manual is limited to light-water reactors in its direct application and presents the practices and experience at one specific reactor centre, it may also be useful for other reactor types because of the general relevance of the problems discussed and the long experience upon which it is based. It has, naturally, no regulatory character but it is hoped that it will be found helpful by staff occupied in all phases of the practical operation of research reactors, and also by those responsible for planning their experimental use. 23 refs, tabs

  8. Physical Characteristics of the Dalat Nuclear Research Reactor

    International Nuclear Information System (INIS)

    Ngo Quang Huy

    1994-10-01

    The operation of the TRIGA MARK II reactor of nominal power 250 KW has been stopped as all the fuel elements have been dismounted and taken away in 1968. The reconstruction of the reactor was accomplished with Russian technological assistance after 1975. The nominal power of the reconstructed reactor is of 500 KW. The recent Dalat reactor is unique of its kind in the world: Russian-designed core combined with left-over infrastructure of the American-made TRIGA II. The reactor was loaded in November 1983. It has reached physical criticality on 1/11/1983 (without central neutron trap) and on 18/12/1983 (with central neutron trap). The power start up occurred in February 1984 and from 20/3/1984 the reactor began to be operated at the nominal power 500 KW. The selected reports included in the proceedings reflect the start up procedures and numerous results obtained in the Dalat Nuclear Research Institute and the Centre of Nuclear Techniques on the determination of different physical characteristics of the reactor. These characteristics are of the first importance for the safe operation of the Dalat reactor

  9. Basic research using the 250 KW research reactor triga in Ljubljana, Yugoslavia

    International Nuclear Information System (INIS)

    Dimic, V.

    1983-01-01

    The 25 KW Triga Mark II reactor of J. 'Stefan Institute' was commissioned on May 1966. During the last two years, it has been operated for about 4200 hr/year. According to experience gained with the reactor, most of the cost of reactor operation will be earned through isotope production for local hospitals and industries, performing low cost applied experiments and organizing training courses. The rest was provided through the Research Communities of the Republic of Slovenia. The reactor has been operated for 15 years without major problems and many basic research programmes have been performed. The research is being conducted in the following mainfields: solid state physics, neutron dosimetry, neutron radiography and autoradiography, reactor physics, examination of nuclear fuel using gamma scanning, irradiation of semiconducting materials and neutron activation analysis. (A.J)

  10. An Integrated Management System (IMS) for JM-1 SLOWPOKE-2 research reactor in Jamaica: experiences in documentation

    International Nuclear Information System (INIS)

    Warner, T.

    2014-01-01

    Since the first criticality in March 1984, the Jamaica SLOWPOKE-2 research reactor at the University of the West Indies, Mona located in the department of the International Centre for Environmental and Nuclear Sciences (ICENS) has operated for approximately 52% of the lifetime of the existing core configuration. The 20kW pool type research reactor has been primarily used for neutron activation analysis in environmental, agricultural, geochemical, health-related studies and mineral exploration in Jamaica. The involvement of the JM-1 reactor for research and teaching activities has segued into commercial applications which, coupled with the current core conversion programme from HEU to LEU, has demanded the implementation of management systems to satisfy regulatory requirements and assure compliance with internationally defined quality standards. At ICENS, documentation related to the Quality Management System aspect of an Integrated Management System (IMS) is well underway. The quality system will incorporate operational and nuclear safety, training, maintenance, design, utilization, occupational health and safety, quality service, and environmental management for its Nuclear Analytical Laboratory, NAL. The IMS is being designed to meet the requirements of the IAEA GS-R-3 with additional controls from international standards including: ISO/IEC 17025:2005, ISO 9001:2008, ISO 14001:2004 and OHSAS 18001:2007. This paper reports on the experiences of the documentation process in a low power reactor facility characterized by limited human resource, where innovative mechanisms of system automation and modeling are included to increase productivity and efficiency. (author)

  11. An Integrated Management System (IMS) for JM-1 SLOWPOKE-2 research reactor in Jamaica: experiences in documentation

    Energy Technology Data Exchange (ETDEWEB)

    Warner, T., E-mail: traceyann.warner02@uwimona.edu.jm [Univ. of West Indies, Mona (Jamaica)

    2014-07-01

    Since the first criticality in March 1984, the Jamaica SLOWPOKE-2 research reactor at the University of the West Indies, Mona located in the department of the International Centre for Environmental and Nuclear Sciences (ICENS) has operated for approximately 52% of the lifetime of the existing core configuration. The 20kW pool type research reactor has been primarily used for neutron activation analysis in environmental, agricultural, geochemical, health-related studies and mineral exploration in Jamaica. The involvement of the JM-1 reactor for research and teaching activities has segued into commercial applications which, coupled with the current core conversion programme from HEU to LEU, has demanded the implementation of management systems to satisfy regulatory requirements and assure compliance with internationally defined quality standards. At ICENS, documentation related to the Quality Management System aspect of an Integrated Management System (IMS) is well underway. The quality system will incorporate operational and nuclear safety, training, maintenance, design, utilization, occupational health and safety, quality service, and environmental management for its Nuclear Analytical Laboratory, NAL. The IMS is being designed to meet the requirements of the IAEA GS-R-3 with additional controls from international standards including: ISO/IEC 17025:2005, ISO 9001:2008, ISO 14001:2004 and OHSAS 18001:2007. This paper reports on the experiences of the documentation process in a low power reactor facility characterized by limited human resource, where innovative mechanisms of system automation and modeling are included to increase productivity and efficiency. (author)

  12. The IAEA collaborating centre for neutron activation based methodologies of research reactors

    International Nuclear Information System (INIS)

    Bode, P.

    2010-01-01

    further benefit from the Institute's plans for a cos t-effective upgrade of the present HOR taking into account recent technological developments and the objectives mentioned above, an upgrade will be realized involving three relatively independent modifications: (i) The installation of a cold-neutron source in beam-tube R2; (ii) Conversion of the core geometry to a 3x3 ultra-compact core including the positioning of a Be-reflector block around the radial beam tubes and (iii) An increase of the nominal power level to 3MW: The neutron beam instruments will gain orders of magnitude in performance, opening new research avenues; the thermal neutron fluence rates in the irradiation facilities for activation will go up by a factor 10-30 and a factor of 7 higher positron output (∼1.5x10 9 s -1 ) is expected. The work plan for the collaboration between the IAEA and the Reactor Institute Delft will be outlined in the presentation as well as the current research activities at the reactor facilities and their opportunities.

  13. Technical specifications: Health Physics Research Reactor

    International Nuclear Information System (INIS)

    1979-02-01

    The technical specifications define the key limitations that must be observed for safe operation of the Health Physics Research Reactor (HPRR) and an envelope of operation within which there is assurance that these limits will not be exceeded. The specifications were written to satisfy the requirements of the Department of Energy (DOE) Manual Chapter 0540, September 1, 1972

  14. Advanced fuel in the Budapest research reactor

    International Nuclear Information System (INIS)

    Hargitai, T.; Vidovsky, I.

    1997-01-01

    The Budapest Research Reactor, the first nuclear facility of Hungary, started to operate in 1959. The main goal of the reactor is to serve neutron research, but applications as neutron radiography, radioisotope production, pressure vessel surveillance test, etc. are important as well. The Budapest Research Reactor is a tank type reactor, moderated and cooled by light water. After a reconstruction and upgrading in 1967 the VVR-SM type fuel elements were used in it. These fuel elements provided a thermal power of 5 MW in the period 1967-1986 and 10 MW after the reconstruction from 1992. In the late eighties the Russian vendor changed the fuel elements slightly, i.e. the main parameters of the fuel remained unchanged, however a higher uranium content was reached. This new fuel is called VVR-M2. The geometry of VVR-SM and VVR-M2 are identical, allowing the use to load old and new fuel assemblies together to the active core. The first new type fuel assemblies were loaded to the Budapest Research Reactor in 1996. The present paper describes the operational experience with the new type of fuel elements in Hungary. (author)

  15. Radionuclide release from research reactor spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Curtius, H., E-mail: h.curtius@fz-juelich.de [Forschungszentrum Juelich, Institut fuer Energieforschung, IEF-6 Sicherheitsforschung und Reaktortechnik, Geb. 05.3, D-52425 Juelich (Germany); Kaiser, G.; Mueller, E.; Bosbach, D. [Forschungszentrum Juelich, Institut fuer Energieforschung, IEF-6 Sicherheitsforschung und Reaktortechnik, Geb. 05.3, D-52425 Juelich (Germany)

    2011-09-01

    Numerous investigations with respect to LWR fuel under non oxidizing repository relevant conditions were performed. The results obtained indicate slow corrosion rates for the UO{sub 2} fuel matrix. Special fuel-types (mostly dispersed fuels, high enriched in {sup 235}U, cladded with aluminium) are used in German research reactors, whereas in German nuclear power plants, UO{sub 2}-fuel (LWR fuel, enrichment in {sup 235}U up to 5%, zircaloy as cladding) is used. Irradiated research reactor fuels contribute less than 1% to the total waste volume. In Germany, the state is responsible for fuel operation and for fuel back-end options. The institute for energy research (IEF-6) at the Research Center Juelich performs investigation with irradiated research reactor spent fuels under repository relevant conditions. In the study, the corrosion of research reactor spent fuel has been investigated in MgCl{sub 2}-rich salt brine and the radionuclide release fractions have been determined. Leaching experiments in brine with two different research reactor fuel-types were performed in a hot cell facility in order to determine the corrosion behaviour and the radionuclide release fractions. The corrosion of two dispersed research reactor fuel-types (UAl{sub x}-Al and U{sub 3}Si{sub 2}-Al) was studied in 400 mL MgCl{sub 2}-rich salt brine in the presence of Fe{sup 2+} under static and initially anoxic conditions. Within these experimental parameters, both fuel types corroded in the experimental time period of 3.5 years completely, and secondary alteration phases were formed. After complete corrosion of the used research reactor fuel samples, the inventories of Cs and Sr were quantitatively detected in solution. Solution concentrations of Am and Eu were lower than the solubility of Am(OH){sub 3}(s) and Eu(OH){sub 3}(s) solid phases respectively, and may be controlled by sorption processes. Pu concentrations may be controlled by Pu(IV) polymer species, but the presence of Pu(V) and Pu

  16. Research reactor developments in Australia

    International Nuclear Information System (INIS)

    Godfrey, Robert

    1998-01-01

    The Australian Nuclear Science and Technology Organization (ANSTO) operates the 10 MW research reactor, HIFAR, at the Lucas Heights site approximately 30 kilometres south of Sydney. Although recent reviews and inspections have confirmed that HIFAR operates safely by an adequate margin and has minimal impact, it was concluded that the reactor design and age places limitations on its operation and utilization, and that HIFAR is approaching the end of its economic life. In September 1997, a decision was made by the Australian Government to found ANSTO for the construction of a replacement research on the existing Lucas Heights site, subject to the requisite environmental assessment process. A draft EIS has been prepared and is currently undergoing public review. A design specification is in preparation, and a research <