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Sample records for nrx-a3 reactor

  1. Summary of loops in the Chalk River NRX and NRU reactors

    International Nuclear Information System (INIS)

    Nishimura, D.T.

    1980-12-01

    The design and operating parameters of the high pressure, high temperature light water loops in the NRX and NRU reactors are presented to assist experimenters reviewing these facilities for their experiments. The NRX and NRU reactor design and operating data of interest to the experimenters are also presented. (author)

  2. NRX and NRU reactor research facilities and irradiation and examination charges

    International Nuclear Information System (INIS)

    1960-08-01

    This report details the irradiation and examination charges on the NRX and NRU reactors at the Chalk River Nuclear Labs. It describes the NRX and NRU research facilities available to external users. It describes the various experimental holes and loops available for research. It also outlines the method used to calculate the facilities charges and the procedure for applying to use the facilities as well as the billing procedures.

  3. The poisoning of NRX pile

    International Nuclear Information System (INIS)

    Watson, W.H.

    1959-09-01

    The experimental methods used to study the poisoning of the NRX reactor are described and the operation of the reactor in relation to these methods is reviewed for the period February to September 1948. (author)

  4. Follow-up of CRNL employees involved in the NRX reactor clean-up

    International Nuclear Information System (INIS)

    Werner, M.M.; Myers, D.K.; Morrison, D.P.

    1982-07-01

    Data available to date on the mortality of continuing and retired employees of the Chalk River Nuclear Laboratories are consistent with the Σhealthy workerΣ effect that has been observed in similar studies at other nuclear facilities. Because of an accident at the NRX research reactor in December 1952, the reactor was largely dismantled and rebuilt in 1953-54. These operations involved appreciable radiation exposures to a number of employees. The follow-up of the 850 on-site AECL staff involved in the clean-up has indicated that there were no unusual patterns in the mortality of this group when compared with those of the general population of Ontario

  5. The genesis of NRX - first stop on the road to CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Arsenault, J.E.

    2015-12-15

    The CANDU (Canada Deuterium Uranium) reactor family is traceable to World War Two and to certain principal players. They gave Canada an early entry into the world of practical nuclear energy with the NRX (National Research Experimental) reactor, developed under the auspices of the ABC (America, Britain, Canada) countries. The principal players on the NRX project were, in alphabetical order: James Chadwick, C.D. Howe, Leslie R. Groves and C-J. Mackenzie. Their careers up to about 1940 are outlined in Section 10. (author)

  6. Dealing with Historical Discrepancies: The Recovery of National Research Experiment (NRX) Reactor Fuel Rods at Chalk River Laboratories (CRL) - 13324

    International Nuclear Information System (INIS)

    Vickerd, Meggan

    2013-01-01

    Following the 1952 National Research Experiment (NRX) Reactor accident, fuel rods which had short irradiation histories were 'temporarily' buried in wooden boxes at the 'disposal grounds' during the cleanup effort. The Nuclear Legacy Liabilities Program (NLLP), funded by Natural Resources Canada (NRCan), strategically retrieves legacy waste and restores lands affected by Atomic Energy of Canada Limited (AECL) early operations. Thus under this program the recovery of still buried NRX reactor fuel rods and their relocation to modern fuel storage was identified as a priority. A suspect inventory of NRX fuels was compiled from historical records and various research activities. Site characterization in 2005 verified the physical location of the fuel rods and determined the wooden boxes they were buried in had degraded such that the fuel rods were in direct contact with the soil. The fuel rods were recovered and transferred to a modern fuel storage facility in 2007. Recovered identification tags and measured radiation fields were used to identify the inventory of these fuels. During the retrieval activity, a discrepancy was discovered between the anticipated number of fuel rods and the number found during the retrieval. A total of 32 fuel rods and cans of cut end pieces were recovered from the specified site, which was greater than the anticipated 19 fuel rods and cans. This discovery delayed the completion of the project, increased the associated costs, and required more than anticipated storage space in the modern fuel storage facility. A number of lessons learned were identified following completion of this project, the most significant of which was the potential for discrepancies within the historical records. Historical discrepancies are more likely to be resolved by comprehensive historical record searches and site characterizations. It was also recommended that a complete review of the wastes generated, and the total affected lands as a result of this historic

  7. Measurement of the stored energy in the NRX reactor reflector graphite

    Energy Technology Data Exchange (ETDEWEB)

    Hilton, H. B.; Larson, E. A.G.

    1959-07-15

    With the co-operation of workers at Windscale and Harwell, whose assistance is hereby gratefully acknowledged, the stored energy content of the inner reflector graphite of NRX has been measured. Measurements made at three different elevations and at different positions through the reflector show that there is, at present, no danger to NRX from an accidental release of the energy. The energy stored in the reflector in 1958 is less by a factor five to ten than the stored energy as measured in 1953. It appears that there has been a continual release of stored energy since 1954 when, after the rehabilitation, the maximum power was raised to 40 MW. Additional thermocouples have been installed in the inner reflector, and future stored energy measurements are being scheduled. (author)

  8. Some loop experiments in the NRX reactor to study the corrosion of mild steel by flowing water at 90oF

    International Nuclear Information System (INIS)

    Allison, G.M.

    1956-11-01

    This work was undertaken to find the water conditions necessary for minimum corrosion in the mild steel thermal shield recirculating systems in NRX and NRU. This report contains the chemical and corrosion results obtained by operating three mild steel loops in which water at 85-95 o F was recirculated through test sections located in J-rod positions in the NRX reactor. Lowest corrosion rates were found when the water was maintained at pH 10.5 with or without oxygen being present. In both cases the corrosion was general in nature and no pitting occurred. At pH 7 with oxygen present in the water severe pitting took place and the corrosion rate was several times higher than similar conditions without oxygen in the water. Under oxygen-free conditions the corrosion product was Fe 3 O 4 . At pH 7 and with 3-5 ppm of O 2 in the water the corrosion product was a mixture of Fe 3 O 4 and γ-Fe 2 O 3 . At high pH with oxygen present Fe 3 O 4 predominated with some traces of Fe 2 O 3 . The systems tested may he listed in order of increasing corrosiveness: High pH with or without O 2 in the water 2 present and continual purification 2 present and no purification or pH control 2 present. (author)

  9. Research reactor put Canada in the nuclear big time

    International Nuclear Information System (INIS)

    Anon.

    1988-01-01

    The history of the NRX reactor is briefly recounted. When NRX started up in 1947, it was the most powerful neutron source in the world. It is now the oldest research reactor still operating. NRX had to be rebuilt after an accident in 1952, and its calandria was changed again in 1970. Loops in NRX were used to test fuel for the Nautilus submarine, and the first zircaloy pressure tube in the world. At the present time, NRX is in a 'hot standby' condition as a backup to the NRU reactor, which is used mainly for isotope production. NRX will be decommissioned after completion and startup of the new MAPLE-X reactor

  10. L-Area STS MTR/NRU/NRX Grapple Assembly Closure Mechanics Review

    International Nuclear Information System (INIS)

    Huizenga, D. J.

    2016-01-01

    A review of the closure mechanics associated with the Shielded Transfer System (STS) MTR/NRU/NRX grapple assembly utilized at the Savannah River Site (SRS) was performed. This review was prompted by an operational event which occurred at the Canadian Nuclear Laboratories (CNL) utilizing a DTS-XL grapple assembly which is essentially identical to the STS MTR/NRU/NRX grapple assembly used at the SRS. The CNL operational event occurred when a NRU/NRX fuel basket containing spent nuclear fuel assemblies was inadvertently released by the DTS-XL grapple assembly during a transfer. The SM review of the STS MTR/NRU/NRX grapple assembly will examine the operational aspects of the STS and the engineered features of the STS which prevent such an event at the SRS. The design requirements for the STS NRU/NRX modifications and the overall layout of the STS are provided in other documents.

  11. Some loop experiments in the NRX reactor to study the corrosion of mild steel by flowing water at 90{sup o}F

    Energy Technology Data Exchange (ETDEWEB)

    Allison, G. M.

    1956-11-15

    This work was undertaken to find the water conditions necessary for minimum corrosion in the mild steel thermal shield recirculating systems in NRX and NRU. This report contains the chemical and corrosion results obtained by operating three mild steel loops in which water at 85-95{sup o}F was recirculated through test sections located in J-rod positions in the NRX reactor. Lowest corrosion rates were found when the water was maintained at pH 10.5 with or without oxygen being present. In both cases the corrosion was general in nature and no pitting occurred. At pH 7 with oxygen present in the water severe pitting took place and the corrosion rate was several times higher than similar conditions without oxygen in the water. Under oxygen-free conditions the corrosion product was Fe{sub 3}O{sub 4}. At pH 7 and with 3-5 ppm of O{sub 2} in the water the corrosion product was a mixture of Fe{sub 3}O{sub 4} and {gamma}-Fe{sub 2}O{sub 3}. At high pH with oxygen present Fe{sub 3}O{sub 4} predominated with some traces of Fe{sub 2}O{sub 3}. The systems tested may he listed in order of increasing corrosiveness: High pH with or without O{sub 2} in the water < water at pH 7 with no O{sub 2} present and continual purification < water with no O{sub 2} present and no purification or pH control < water at pH 7 with 3-5 ppm of O{sub 2} present. (author)

  12. Performance of the NRX shut-off rods

    International Nuclear Information System (INIS)

    Manson, R.E.

    1965-08-01

    A new type of shut-off rod of electromechanical design was developed by the American Machine and Foundry Company for use in the NRX reactor following the accident of 1952. The new rods were installed in May, 1956, as part of the control system conversion program which was completed in 1958. Some problems were encountered with limit switch adjustment but minor modifications in design led to much improved operation. he performance of the rods also improved as more experience was gained in the maintenance and adjustment of the various headgear components. Each headgear is now overhauled once a year on a routine basis. The present design of shut-off rod is considered to be very satisfactory. There has only been one occasion when a shut-off rod has failed to come fully down on a trip. Rods have failed to operate correctly on five other occasions but these occurred during shutdown periods or when the reactor was being shutdown manually. (author)

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

    International Nuclear Information System (INIS)

    Winfield, D.J.

    1990-01-01

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

  14. CRNL research reactor retrofit Emergency Filtration System

    International Nuclear Information System (INIS)

    Philippi, H.M.

    1990-01-01

    This paper presents a brief history of NRX and NRU research reactor effluent air treatment systems before describing the selection and design of an appropriate retrofit Emergency Filtration System (EFS) to serve these reactors and the future MX-10 isotope production reactor. The conceptual design of the EFS began in 1984. A standby concrete shielding filter-adsorber system, sized to serve the reactor with the largest exhaust flow, was selected. The standby system, bypassed under normal operating conditions, is equipped with normal exhaust stream shutoff and diversion valves to be activated manually when an emergency is anticipated, or automatically when emergency levels of gamma radiation are detected in the exhaust stream. The first phase of the EFS installation, that is the construction of the EFS and the connection of NRU to the system, was completed in 1987. The second phase of construction, which includes the connection of NRX and provisions for the future connection of MX-10, is to be completed in 1990

  15. Power auxiliaries and research reactors. Section 3 of Symposium on the peaceful uses of atomic energy in Australia, 1958, held in Sydney, in June 1958

    Energy Technology Data Exchange (ETDEWEB)

    None

    1958-10-15

    The problems of disposing of the large amounts of highly-radioactive waste resulting from a large-scale nuclear power program are reviewed. The Canadian research reactor NRX is discussed. The DIDO reactor is briefly described and operating experience for the first year at high flux is summarized. The core of the High Flux Australian Research Reactor (HIFAR) is described, and some reactivity balance data are given (T.R.H.)

  16. Comparison Between Calculated and Measured Cross Section Changes in Natural Uranium Irradiated in NRX

    Energy Technology Data Exchange (ETDEWEB)

    Ahlstroem, P E

    1961-03-15

    It is desirable to obtain an experimental check of the reliability of the methods currently used to determine reactivity changes in a reactor and, with a view to meeting this requirement to some extent, a preliminary comparison has been made between calculated and measured cross-section changes in rods of natural uranium irradiated in NRX. The measurements were made at Harwell in the GLEEP reactor and a description has been given by, inter alia, Ward and Craig. The theory of the calculations, which is briefly described in this report, has been indicated by Littler. The investigation showed that the methods for calculating burn up used at present provides a good illustration of the long-term variations in isotope contents. A satisfactory agreement is obtained with experimental results when calculating apparent cross-section changes in uranium rods due to irradiation if the fission cross- section for {sup 239}Pu is set to 780 b. This is 34 b higher than the figure quoted in BNL - 325 (1958). However, in order to get a good idea as to whether the calculated long-term variations in reactivity really correspond to reality, it is necessary to make further investigations. For this reason the results quoted in this report should be regarded as preliminary.

  17. Fuels for Canadian research reactors

    International Nuclear Information System (INIS)

    Feraday, M.A.

    1993-01-01

    This paper includes some statements and remarks concerning the uranium silicide fuels for which there is significant fabrication in AECL, irradiation and defect performance experience; description of two Canadian high flux research reactors which use high enrichment uranium (HEU) and the fuels currently used in these reactors; limited fabrication work done on Al-U alloys to uranium contents as high as 40 wt%. The latter concerns work aimed at AECL fast neutron program. This experience in general terms is applied to the NRX and NRU designs of fuel

  18. 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

  19. Calculation of fast neutron flux in reactor pressure tubes and experimental facilities

    Energy Technology Data Exchange (ETDEWEB)

    Barnett, P. C. [Canadian General Electric (Canada)

    1968-07-15

    The computer program EPITHET was used to calculate the fast neutron flux (>1 MeV) in several reactor pressure tubes and experimental facilities in order to compare the fast neutron flux in the different cases and to provide a self-consistent set of flux values which may be used to relate creep strain to fast neutron flux . The facilities considered are shown below together with the calculated fast neutron flux (>1 MeV). Fast flux 10{sup 13} n/cm{sup 2}s: NPD 1.14, Douglas Point 2.66, Pickering 2.89, Gentilly 2.35, SGHWR 3.65, NRU U-1 and U-2 3.25'' pressure tube - 19 element fuel 3.05, NRU U-1 and U-2 4.07'' pressure tube - 28 element fuel 3.18, NRU U-1 and U-2 4.07'' pressure tube - 18 element fuel 2.90, NRX X-5 0.88, PRTR Mk I fuel 2.81, PRTR HPD fuel 3.52, WR-1 2.73, Mk IV creep machine (NRX) 0.85, Mk VI creep machine (NRU) 2.04, Biaxial creep insert (NRU U-49) 2.61.

  20. The development of lower enrichment fuels for Canadian research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Feraday, M A; Belanger, L; Grolway, C M [AECL, Atomic Energy of Canada Limited, Chalk River, ON (Canada); Foo, M T [CRNL, Combustion Engineering Superheater Ltd., Moncton, NB (Canada)

    1983-08-01

    As part of the world wide move to proliferation resistant fuels, new fuels which use reduced enrichment uranium are being developed for use in the NRX and NRU reactors. A fuel consisting of particles of a USiAl alloy dispersed in an Al matrix has been selected for development along with Al-37 wt% U alloy and Al-U{sub 3}O{sub 8} cermet as backup fuels. This report outlines the progress made in the development of the Al-USiAl and Al-37 wt% U. Results show that good quality extruded rods containing either fuel can be made with techniques similar to those used to fabricate the current NRX and NRU fuels. However, the new fuels will be more expensive to make. Although the oxidation behaviour of the Al-USiAl is not as good as that of the Al-U alloys, its corrosion behaviour in high temperature water does not seem much worse. The oxidation and aqueous corrosion of A-37 wt% U are not much different from those of the Al-U alloys currently used. (author)

  1. Detection of gaseous fission products in water - a method of monitoring fuel sheathing failures

    Energy Technology Data Exchange (ETDEWEB)

    Tunnicliffe, P. R.; Whittier, A. C.

    1959-05-15

    The gaseous activities stripped from samples of effluent coolant from the NRU fuel elements tested in the central thimble of the NRX reactor (NRU loop) and from the NRX main effluent have been investigated. The activities obtained from the NRU loop can be attributed to gaseous fission products only. Design data have been obtained for a 'Gaseous Fission Product Monitor' to be installed for use with the NRU reactor. It is expected that this monitor will have high sensitivity to activity indicative of an incipient fuel element sheath failure. No qualitative determination of the various gaseous activities obtained from the NRX effluent has been made. A strong component of 25 {+-}1 seconds half-life is not consistent with O-19. Limited information concerning sheath failures in NRX was obtained. Of six failures observed in parallel with the installed delayed neutron monitors, three of these gave pre-warnings and in each case the gaseous fission product monitor showed a substantially greater sensitivity. An experiment in which small samples of uranium, inserted into the NRX reactor, could be exposed at will to a stream of water showed the behaviour of the two types of monitors to be similar. However, a number of signals were detected only by the gaseous fission product monitor. These can be attributed to its sensitivity to relatively long lived fission products. (author)

  2. An analysis of water reactor burnup data with the METHUSELAH II code

    International Nuclear Information System (INIS)

    Floyd, M.; Hicks, D.

    1964-10-01

    The METHUSELAH II code has been used to predict long term reactivity and isotopic changes in the YANKEE, Dresden and NRX reactors. In general it is shown that there is a satisfactory measure of agreement and the first core lives of YANKEE and Dresden appear well predicted. However there are discrepancies in the isotopic composition of the plutonium formed which appear to be correlated with the degree of hardness of the reactor spectrum. It is demonstrated that plausible changes in nuclear data could reduce the discrepancies. (author)

  3. Reactor loops at Chalk River

    International Nuclear Information System (INIS)

    Sochaski, R.O.

    1962-07-01

    This report describes broadly the nine in-reactor loops, and their components, located in and around the NRX and NRU reactors at Chalk River. First an introduction and general description is given of the loops and their function, supplemented with a table outlining some loop specifications and nine simplified flow sheets, one for each individual loop. The report then proceeds to classify each loop into two categories, the 'main loop circuit' and the 'auxiliary circuit', and descriptions are given of each circuit's components in turn. These components, in part, are comprised of the main loop pumps, the test section, loop heaters, loop coolers, delayed-neutron monitors, surge tank, Dowtherm coolers, loop piping. Here again photographs, drawings and tables are included to provide a clearer understanding of the descriptive literature and to include, in tables, some specifications of the more important components in each loop. (author)

  4. Effects of heat and pressure on the swelling of irradiated uranium. Part I

    Energy Technology Data Exchange (ETDEWEB)

    Churcman, A. T.; Barnes, R. S.; Cottrell, A. H.

    1956-09-15

    Small pieces of a natural uranium fuel bar from the NRX reactor which had been irradiated to 0.3 - 0.4% burn up have been heat treated either in vacuo or at high pressure and changes in their density measured.

  5. An estimate of the reactivity of assemblies of NRX fuel elements in light water

    International Nuclear Information System (INIS)

    Jarvis, R.G.

    1960-03-01

    This report contains calculations on the criticality of assemblies of NRX fuel elements in light water. The elements are dealt with in three sections, 'X rods' of natural uranium, enriched elements of U 235 /A1 alloy and enriched elements of Pu/Al alloy. Values of k ∞ and B 2 are provided for two fuel concentrations for each of the two enriched types and for a range of irradiations of the X rods. The calculations for the X rods provide maximum and minimum values of k ∞ . The maximum values for some lattices are a few per cent above unity. Unfortunately, the present experimental evidence does not prove that it is impossible to achieve values of k ∞ greater than unity in lattices of natural uranium in light water. Hence for safety predictions maximum values have been used. The resulting restrictions are not very severe. It is possible to make critical assemblies of the enriched elements, Part (5) contains a set of recommended minimum spacings such that elements of all kinds may safely be mixed in a stack together. There are also predictions of the minimum critical numbers of complete elements or elements cut into slugs. (author)

  6. Decommissioning of a small reactor (BR3 reactor, Belgium)

    International Nuclear Information System (INIS)

    Dadoumont, J.; Massaut, V.; Klein, M.; Demeulemeester, Y.

    2002-01-01

    Since 1989, SCK-CEN has been dismantling its PWR reactor BR3 (Belgian Reactor No. 3). After gaining a great deal of experience in remote dismantling of highly radioactive components during the actual dismantling of the two sets of internals, the BR3 team completed the cutting of its reactor pressure vessel (RPV). During the feasibility phase of the RPV dismantling, a decision was made to cut it under water in the refuelling pool of the plant, after having removed it from its cavity. The RPV was cut into segments using a milling cutter and a bandsaw machine. These mechanical techniques have shown their ability for this kind of operations. Prior to the segmentation, the thermal insulation situated around the RPV was remotely removed and disposed of. The paper will describe all these operations. The BR3 decommissioning activities also include the dismantling of contaminated loops and equipment. After a careful sorting of the pieces, optimized management routes are selected in order to minimize the final amount of radioactive waste to be disposed of. Some development of different methods of decontamination were carried out: abrasive blasting (or sand blasting), chemical decontamination (Oxidizing-Reducing process using Cerium). The main goal of the decontamination program is to recycle most of the metallic materials either in the nuclear world or in the industrial world by reaching the respective recycling or clearance level. Overall the decommissioning of the BR3 reactor has shown the feasibility of performing such a project in a safe and economical way. Moreover, BR3 has developed methodologies and decontamination processes to economically reduce the amount of radwaste produced. (author)

  7. JENDL-3.3 thermal reactor benchmark test

    International Nuclear Information System (INIS)

    Akie, Hiroshi

    2001-01-01

    Integral tests of JENDL-3.2 nuclear data library have been carried out by Reactor Integral Test WG of Japanese Nuclear Data Committee. The most important problem in the thermal reactor benchmark testing was the overestimation of the multiplication factor of the U fueled cores. With several revisions of the data of 235 U and the other nuclides, JENDL-3.3 data library gives a good estimation of multiplication factors both for U and Pu fueled thermal reactors. (author)

  8. Dynamics of TRIGA-3 Salazar Reactor

    International Nuclear Information System (INIS)

    Gallardo S, L.F.

    1990-01-01

    The theoretical study of temporal behavior of a nuclear reactor is of great importance, since it allows to know, in advance, the conditions to which a reactor is going to be submitted. The reliability of two computer codes (AIREK-JEN and PLANKIN) designed to reproduce the temporal behavior of nuclear reactors, generally power reactors, when they are applied to reproduce the dynamic behavior of TRIGA-3 Salazar Reactor is analyzed. In the first chapters, the fundamental equations that solve this computer codes are deduced, and also the main characteristics of TRIGA-3 Salazar Reactor and the necessary data to run the programs are presented; later the results obtained with the computer codes and the experimental results reported in the operational logbook of the reactor are compared, with the result that such computer codes are applicable to the temporal study of TRIGA-3 Salazar Reactor. (Author)

  9. Removal in a lump of JRR-3 nuclear reactor

    International Nuclear Information System (INIS)

    Ohnishi, Nobuaki; Suzuki, Masanori; Nagase, Tetsuo; Watanabe, Morinari.

    1989-01-01

    The research reactor JRR-3 in Japan Atomic Energy Research Institute is called 'Home made No.1 reactor' as all except fuel and heavy water as the moderator and coolant were manufactured in Japan. The JRR-3 attained the criticality in 1962, and the cumulative time of operation reached 47135.5 hours, and the cumulative power output reached 419073.5 MWh. It was stopped in 1983. During the period, it was utilized for beam experiment, irradiation of fuel and materials, RI production and others. In order to cope with the expansion of utilization and the advance of utilizing technology of the research reactor, the reconstruction works are in progress, and the criticality of the reconstructed reactor is expected in 1990. On the site where the old reactor is removed, the reactor of different type is installed, and the first large cold neutron source is equipped. In this report, as to the removal of the old reactor proper, the method of working and the results are described. Considering the period of working, the cost and the management of the removed reactor, in the case of the JRR-3, the method of carrying it out in a lump was adopted as the optimum removal method. The plan, procedure and results of the removal working are reported. (K.I.)

  10. The MAPLE-X concept dedicated to the production of radio-isotopes

    International Nuclear Information System (INIS)

    Heeds, W.

    1985-06-01

    MAPLE is a versatile new Canadian multi-purpose research reactor concept that meets the nuclear aspirations of developing countries. It is planned to convert the NRX reactor at Chalk River Nuclear Laboratories into MAPLE-X as a demonstration prototype of this concept and thereafter to dedicate its operation to the production of radio-isotopes. A description of MAPLE-X and details of molybdenum-99 production are given

  11. Description of the RA-3 research reactor as a model facility

    International Nuclear Information System (INIS)

    Vicens, Hugo E.; Quintana, Jorge A.

    2001-01-01

    The Argentine RA-3 reactor is described as a model facility for the information to be provided to the IAEA in accordance with the requirements of the Model Additional Protocol. RA-3 reactor was designed as a 5 MW swimming pool reactor, moderated and cooled with light water. Its fuel was 90% enriched uranium. The reactor started its operation in 1967, has been modified and improved in many components, including the core, that now is fueled with moderately enriched uranium

  12. Future development in heavy water reactors in Canada

    International Nuclear Information System (INIS)

    Donnelly, J.; Hart, R.G.

    1982-01-01

    1982 marks the 35th anniversary of the start-up of Canada's first research and test reactor, NRX. Its first power reactor has been operating successfully for the past 20 years. With 5,000 MWe of domestic capacity installed, Canada's major CANDU (Canada Deuterium, Uranium) nuclear program has a further 9,500 MWe under construction in Canada for completion by 1990 as well as committed offshore projects in Argentina, Korea and Romania. The CANDU operating record, by any measure of performance, has been outstanding. This performance is largely due to the discipline imposed on the development, design, construction and operation by two fundamental choices: natural uranium and heavy water. The impact of these two choices on availability, fuel utilization, safety and economics is discussed. Future plans call for building on those characteristics which have made CANDU so successful. When time for change comes, current assessments indicate that it will be possible to convert to more efficient advanced fuel cycles without major changes to the basic CANDU design. Primary attention is being focussed on thorium fuel cycles to ensure an abundant and continuing supply of low cost energy for the long term. The resource savings available from these fuel cycles in expanding systems are reviewed and compared with those available from LWR's and Fast Breeders. The results clearly illustrate the versatility of the CANDU reactor. It can benefit from enrichment plants or get along without them. It can complement LWR's or compete with them. It can complement Fast Breeder Reactors or compete with them as well. In the very long term CANDU's and Fast Breeders combined offer the potential of burning all the world's uranium and all the world's thorium. (author)

  13. The simplified P3 approach on a trigonal geometry in the nodal reactor code DYN3D

    International Nuclear Information System (INIS)

    Duerigen, S.; Fridman, E.

    2011-01-01

    DYN3D is a three-dimensional nodal diffusion code for steady-state and transient analyses of Light-Water Reactors with square and hexagonal fuel assembly geometries. Currently, several versions of the DYN3D code are available including a multi-group diffusion and a simplified P 3 (SP 3 ) neutron transport option. In this work, the multi-group SP 3 method based on trigonal-z geometry was developed. The method is applicable to the analysis of reactor cores with hexagonal fuel assemblies and allows flexible mesh refinement, which is of particular importance for WWER-type Pressurized Water Reactors as well as for innovative reactor concepts including block type High-Temperature Reactors and Sodium Fast Reactors. In this paper, the theoretical background for the trigonal SP 3 methodology is outlined and the results of a preliminary verification analysis are presented by means of a simplified WWER-440 core test example. The accordant cross sections and reference solutions were produced by the Monte Carlo code SERPENT. The DYN3D results are in good agreement with the reference solutions. The average deviation in the nodal power distribution is about 1%. (Authors)

  14. 3D CAD model of the subcritical nuclear reactor of IPN; Modelo CAD 3D del reactor nuclear subcritico del IPN

    Energy Technology Data Exchange (ETDEWEB)

    Pahuamba V, F. de J.; Delfin L, A.; Gomez T, A. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Ibarra R, G.; Del Valle G, E.; Sanchez R, A., E-mail: narehc@hotmail.com [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN, Edif. 9, Unidad Profesional Adolfo Lopez Mateos, San Pedro Zacatenco, 07738 Ciudad de Mexico (Mexico)

    2016-09-15

    The three-dimensional (3D) CAD model of the subcritical reactor Chicago model 9000 of Instituto Politecnico Nacional (IPN) allows obtaining a 3D view with the dimensions of each of its components, such as: natural uranium cylindrical rods, fuel elements, hexagonal reactor core arrangement, cylindrical stainless steel tank containing the core, fuel element support grids and reactor water cleaning system. As a starting point for the development of the model, the Chicago model 9000 subcritical reactor manual provided by the manufacturer was used, the measurement and verification of the components to adapt the geometric, physical and mechanical characteristics was carried out and materials standards were used to obtain a design that allows to elaborate a new manual according to the specifications. In addition, the 3D models of the building of the Advanced Physics Laboratory, neutron generator, cobalt source and the corridors connecting to the subcritical reactor facility were developed, allowing an animated ride, developed by computer-aided design software. The manual provided by the company Nuclear Chicago, dates from the year 1959 and presents diverse deviations in the design and dimensions of the reactor components. The model developed; in addition to supporting the development of the new manual represents a learning tool to visualize the reactor components. (Author)

  15. Integral test of JENDL-3.3 for fast reactors

    International Nuclear Information System (INIS)

    Chiba, Gou

    2003-01-01

    An integral test of JENDL-3.3 was performed for fast reactors. Various types of fast reactors were analyzed. Calculation values of the nuclear characteristics were greatly especially affected by the revisions of the cross sections of U-235 capture and elastic scattering reactions. The C/E values were improved for ZPPR cross where plutonium is mainly fueled, but not for BFS cores where uranium is mainly fueled. (author)

  16. Direct harvesting of Helium-3 (3He) from heavy water nuclear reactors

    International Nuclear Information System (INIS)

    Bentoumi, G.; Didsbury, R.; Jonkmans, G.; Rodrigo, L.; Sur, B.

    2013-01-01

    The thermal neutron activation of deuterium inside a heavy-water-moderated or -cooled nuclear reactor produces a build-up of tritium in the heavy water. The in situ decay of tritium can, for certain reactor types and operating conditions, produce potentially useable amounts of 3 He, which can be directly extracted via the heavy-water cover gas without first separating, collecting and storing tritium outside the reactor. It is estimated that the amount of 3 He available for recovery from the moderator cover gas of a 700 MWe class Pressurized Heavy Water Reactor (PHWR) ranges from 0.1 to 0.7 m 3 (STP) per annum, varying with the tritium activity buildup in the moderator. The harvesting of 3 He would generate approximately 12.7 m 3 (STP) of 3 He, worth more than $30M at current market rates, over a typical 25-year operating cycle of the PHWR. This paper discusses the production of 3 He in the moderator of a PHWR and its extraction from the 4 He moderator cover gas system using conventional methods. (author)

  17. D-3He fuel cycles for neutron lean reactors

    International Nuclear Information System (INIS)

    Kernbichler, W.; Miley, G.H.; Heindler, M.

    1989-01-01

    The intrinsic potential of D-3He as a reactor fuel is investigated for a large range of 3He to D density ratios. A steady-state zero-dimensional reactor model is developed in which much care is attributed to a proper treatment of fast fusion products. Useful ranges of reactor parameters as well as temperature-density windows for driven and ignited operation are identified. Various figures of merit are calculated, such as power densities, net power production, neutron production, tritium load and radiative power. These results suggest several optimistic conclusions about the performance of D-3He as a reactor fuel

  18. Dynamics of TRIGA-3 Salazar Reactor.; Dinamica del Reactor TRIGA Mark III del Centro Nuclear de Mexico.

    Energy Technology Data Exchange (ETDEWEB)

    Gallardo S, L F

    1991-12-31

    The theoretical study of temporal behavior of a nuclear reactor is of great importance, since it allows to know, in advance, the conditions to which a reactor is going to be submitted. The reliability of two computer codes (AIREK-JEN and PLANKIN) designed to reproduce the temporal behavior of nuclear reactors, generally power reactors, when they are applied to reproduce the dynamic behavior of TRIGA-3 Salazar Reactor is analyzed. In the first chapters, the fundamental equations that solve this computer codes are deduced, and also the main characteristics of TRIGA-3 Salazar Reactor and the necessary data to run the programs are presented; later the results obtained with the computer codes and the experimental results reported in the operational logbook of the reactor are compared, with the result that such computer codes are applicable to the temporal study of TRIGA-3 Salazar Reactor. (Author).

  19. Ignition access in a D-3He helical reactor

    International Nuclear Information System (INIS)

    Mitarai, Osamu

    2003-01-01

    Ignition access in a D- 3 He helical reactor is studied based on 0-dimensional particle and power balance equations for deuterium, tritium, helium-3, alpha ash, proton ash, electron density and temperature. The calculations are based on the following experimental facts observed in LHD. (author)

  20. Conceptual design of D-3He FRC reactor 'ARTEMIS'

    International Nuclear Information System (INIS)

    Momota, H.; Ishida, A.; Kohzaki, Y.

    1991-07-01

    A comprehensive design study of the D- 3 He fueled field-reversed configuration (FRC) reactor 'ARTEMIS' is carried out for the purpose of proving its attractive characteristics and clarifying the critical issues for a commercial fusion reactor. The FRC burning plasma is stabilized and sustained in a steady equilibrium by means of a preferential trapping of D- 3 He fusion-produced energetic protons. A novel direct energy converter for 15MeV protons is also presented. On the bases of a consistent scenario of the fusion plasma production and simple engineering, a compact and simple reactor concept is presented. The design of the D- 3 He FRC power plant definitely offers the most attractive prospect for energy development. It is environmentally acceptable in view of radio-activity and fuel resources; and the estimated cost of electricity is low compared to a light water reactor. Critical issues concerning physics or engineering for the development of the D- 3 He FRC reactor are clarified. (author)

  1. CNTNAP2 and NRXN1 are mutated in autosomal-recessive Pitt-Hopkins-like mental retardation and determine the level of a common synaptic protein in Drosophila

    DEFF Research Database (Denmark)

    Zweier, Christiane; de Jong, Eiko K; Zweier, Markus

    2009-01-01

    , phenotypically overlapping with Pitt-Hopkins syndrome. With a frequency of at least 1% in our cohort of 179 patients, recessive defects in CNTNAP2 appear to significantly contribute to severe MR. Whereas the established synaptic role of NRXN1 suggests that synaptic defects contribute to the associated...... protein can reorganize synaptic morphology and induce increased density of active zones, the synaptic domains of neurotransmitter release. Moreover, both Nrx-I and Nrx-IV determine the level of the presynaptic active-zone protein bruchpilot, indicating a possible common molecular mechanism in Nrx...

  2. A robot-automated work site for repair of the Chinon A3 reactor

    International Nuclear Information System (INIS)

    Raynal, A.

    1987-01-01

    In 1982, following degradation due to corrosion of low-carbon steel by carbon dioxide gas, the utility undertook to repair some of the support structures at Chinon A3. This involved consolidation and reinforcing thermocouples and gas monitor pipeworks supports. A welding process was selected and the use of robots became indispensable because of the large number of components to be replaced (200 per outage). Two robots, supplied with tool heads and replacement components from outside the reactor were used. The robots and their servers were coordinated by a central computer and monitored by a closed circuit television system. Each repair operation was performed after ''training'' on a full-scale mockup of the top of the reactor reconstructed from telemetry of the real reactor dimensions. Since becoming operational in June 1986, the robots have accumulated over 20 000 hours of operation and seventy parts have been welded to the reactor. A 3D CAD system has been adapted to simulate the robots and analyse long trajectories in order to reduce robot learning time [fr

  3. 3D simulation of CANDU reactor regulating system

    International Nuclear Information System (INIS)

    Venescu, B.; Zevedei, D.; Jurian, M.

    2013-01-01

    Present paper shows the evaluation of the performance of the 3-D modal synthesis based reactor kinetic model in a closed-loop environment in a MATLAB/SIMULINK based Reactor Regulating System (RRS) simulation platform. A notable advantage of the 3-D model is the level of details that it can reveal as compared to the coupled point kinetic model. Using the developed RRS simulation platform, the reactor internal behaviours can be revealed during load-following tests. The test results are also benchmarked against measurements from an existing (CANDU) power plant. It can be concluded that the 3-D reactor model produces more realistic view of the core neutron flux distribution, which is closer to the real plant measurements than that from a coupled point kinetic model. It is also shown that, through a vectorization process, the computational load of the 3-D model is comparable with that of the 14-zone coupled point kinetic model. Furthermore, the developed Graphical User Interface (GUI) software package for RRS implementation represents a user friendly and independent application environment for education training and industrial utilizations. (authors)

  4. GRIMH3: A new reactor calculation code at Savannah River Site

    International Nuclear Information System (INIS)

    Le, T.T.; Pevey, R.E.

    1993-01-01

    The GRIMHX reactor code currently in use at the Savannah River Site (SRS) was written at a time when computer processing speed and memory storage were very limited. Recently, a new reactor code (GRIMH3) was written to take advantage of the hardware improvements (vectorization and higher memory capacities) as well as the range of available computers at SRS (workstations and supercomputers). The GRIMH3 code computes the solution of the static multigroup neutron diffusion equation in one-, two-, and three-dimensional hexagonal geometry. Either direct or adjoint solutions can be computed for k eff searches, buckling searches, external neutron sources, power flattening searches, or power normalization factor calculations with 1, 6, 24, 54, or 96 points per hex. The GRIMHX reactor code currently in use at the Savannah River Site (SRS) was written at a time when computer processing speed and memory storage were very limited. Recently, a new reactor code (GRIMH3) was written to take advantage of the hardware improvements (vectorization and higher memory capacities) as well as the range of available computers at SRS (workstations and supercomputers). The GRIMH3 code computes the solution of the static multigroup neutron diffusion equation in one-, two-, and three-dimensional hexagonal geometry. Either direct or adjoint solutions can be computed for k eff searches, buckling searches, external neutron sources, power flattening searches, or power normalization factor calculations with 1, 6, 24, 54, or 96 points per hex

  5. Analyses for MARIA Research Reactor with RELAP/MOD3 code

    International Nuclear Information System (INIS)

    Szczurek, J.; Czerski, P.

    2004-01-01

    This paper deals with the application of the RELAP5/MOD3 code to the transient analyses for MARIA research reactor. Poland's MARIA Research Reactor is water and beryllium moderated, water-cooled reactor of a pool type with pressurized fuel channels containing concentric multi-tube assemblies of highly enriched uranium clad in aluminium. The RELAP5/MOD3 input data model includes the whole primary cooling circuit of the MARIA reactor. The model was qualified against the reactor data at steady state conditions and additionally against the existing reliable experimental data for a transient initiated by the reactor scram. The RELAP transient simulation was performed for loss of forced flow accidents including two scenarios with protected and unprotected (no scram) reactor core. Calculations allow estimating time margin for reactor scram initiation and reactivity feedbacks contribution to the results. (author)

  6. EL3 reactor description and safety analysis report

    International Nuclear Information System (INIS)

    1969-02-01

    The EL-3 reactor is an experimental pile. Heterogenous type reactor, water moderated and cooled it uses slightly enriched uranium oxide as fuel (4.5 percent) distributed in vertical cells that constitute the core (the maximum number of cells is 99). It is conceived to function at a maximal thermal power of 20 MW. It supplies a maximum thermal neutron flux of 10 14 neutrons/cm 2 /sec. It has several experimental devices. The EL-3 reactor is surrounded by auxiliary circuits of fluids, in a sealed containment, slightly depressed. The primary heavy water coolant circuit is completely included in this containment. Its cooling is made by the intermediary of a light water secondary circuit by atmospheric refrigerants. The ventilation circuits of the sealed containment and the reactor block do not release air outside, under nornal functioning, by a particularly studied chimney only after filtering and eventually dilution. The eventual contamination of the light water or air by active products is permanently monitored to allow the reactor shutdown and avoid the release in atmosphere of dangerous products. The EL-3 reactor, laying down in may 1955, has diverged in july 1957, made its first ascending in power in december 1957 and reached its complete power in april 1958. The positioning of actual fuel (snow crystal) was made during summer 1964. Reactor with an experimental aim, it is used for theoretical and technological studies by material irradiation in the experimental channels and the core cells, with possibilities to constitute independent loops (relative to the cooling fluids). Thirty vertical channels are devoted to the fabrication of artificial radioelements [fr

  7. Course of operators of the RA-3 reactor

    International Nuclear Information System (INIS)

    Caligiuri, G.A.

    1983-01-01

    Description of the fundamental principles of the nuclear reactors' control systems. The RA-3 reactor's control and measurement systems are principally described, without setting aside the basic criteria for the design of an appropriate instrumentation for the control of a nuclear reactor, as well as the theory on which the functioning of the several detectors and equipments used in a nuclear instrumentation are based. The main purpose of this course is that of serving, preferentially as a text, for the training of personnel which shall perform operation tasks in this reactor. The work includes three well-defined sections. The first two ones make an introduction to the subject, while the third one, extending to more than half-work, deals with the general description of the system in which the control and operation logic of RA-3 are included. (R.J.S) [es

  8. Early years of nuclear energy research in Canada

    International Nuclear Information System (INIS)

    Laurence, G.C.

    1980-01-01

    The first experimental attempts in Canada to obtain energy from uranium fission were carried out by the author in the Ottawa laboratories of the National Research Council from 1940 to 42. This program grew into a joint British-Canadian laboratory in Montreal. Work done at this laboratory, which moved to Chalk River in 1946, led to the construction of ZEEP (the first nuclear reactor to operate outside of the United States) NRX, and ultimately to the development of the CANDU power reactors. People involved in the work and events along the way are covered in detail. (LL)

  9. Station Blackout Analysis for a 3-Loop Westinghouse PWR Reactor Using Trace

    International Nuclear Information System (INIS)

    El-Sahlamy, N.M.

    2017-01-01

    One of the main concerns in the area of severe accidents in nuclear reactors is that of station blackout (SBO). The loss of offsite electrical power concurrent with the unavailability of the onsite emergency alternating current (AC) power system can result in loss of decay heat removal capability, leading to a potential core damage which may lead to undesirable consequences to the public and the environment. To cope with an SBO, nuclear reactors are provided with protection systems that automatically shut down the reactor, and with safety systems to remove the core residual heat. This paper provides a best estimate assessment of the SBO scenario in a 3-loop Westinghouse PWR reactor. The evaluation is performed using TRACE, a best estimate computer code for thermal-hydraulic calculations. Two sets of scenarios for SBO analyses are discussed in the current work. The first scenario is the short term SBO where it is assumed that in addition to the loss of AC power, there is no DC power; i.e., no batteries are available. In the second scenario, a long term SBO is considered. For this scenario, DC batteries are available for four hours. The aim of the current SBO analyses for the 3-loop pressurized water reactor presented in this paper is to focus on the effect of the availability of a DC power source to delay the time to core uncovers and heatup

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  11. Experiment for search for sterile neutrino at SM-3 reactor

    Science.gov (United States)

    Serebrov, A. P.; Ivochkin, V. G.; Samoylov, R. M.; Fomin, A. K.; Zinoviev, V. G.; Neustroev, P. V.; Golovtsov, V. L.; Gruzinsky, N. V.; Solovey, V. A.; Cherniy, A. V.; Zherebtsov, O. M.; Martemyanov, V. P.; Zinoev, V. G.; Tarasenkov, V. G.; Aleshin, V. I.; Petelin, A. L.; Pavlov, S. V.; Izhutov, A. L.; Sazontov, S. A.; Ryazanov, D. K.; Gromov, M. O.; Afanasiev, V. V.; Matrosov, L. N.; Matrosova, M. Yu.

    2016-11-01

    In connection with the question of possible existence of sterile neutrino the laboratory on the basis of SM-3 reactor was created to search for oscillations of reactor antineutrino. A prototype of a neutrino detector with scintillator volume of 400 l can be moved at the distance of 6-11 m from the reactor core. The measurements of background conditions have been made. It is shown that the main experimental problem is associated with cosmic radiation background. Test measurements of dependence of a reactor antineutrino flux on the distance from a reactor core have been made. The prospects of search for oscillations of reactor antineutrino at short distances are discussed.

  12. In-pile critical heat flux and post-dryout heat transfer measurements – A historical perspective

    Energy Technology Data Exchange (ETDEWEB)

    Groeneveld, D.C., E-mail: degroeneveld@gmail.com

    2017-06-15

    In the 1960s’ and 1970s’ Canada was a world leader in performing in-reactor heat transfer experiments on fuel bundles instrumented with miniature sheath thermocouples. Several Critical Heat Flux (CHF) and Post-CHF experiments were performed in Chalk River’s NRU and NRX reactors on water-cooled 3-, 18-, 19-, 21-, and 36-element fuel bundles. Most experiments were obtained at steady-state conditions, where the power was raised gradually from single-phase conditions up to the CHF and beyond. Occasionally, post-dryout temperatures up to 600 °C were maintained for several hours. In some tests, the fuel behaviour during loss-of-flow and blowdown transients was investigated – during these transients sheath temperatures could exceed 2000 °C. Because of the increasingly more stringent licensing requirements for in-pile heat transfer tests on instrumented fuel bundles, no in-pile CHF and post-dryout tests on fuel bundles have been performed anywhere in the world for the past 40 years. This paper provides details of these unique in-pile experiments and describes some of their heat transfer results.

  13. D-3He fueled FRC reactor 'ARTEMIS-L'

    International Nuclear Information System (INIS)

    Momota, Hiromu; Tomita, Yukihiro; Ishida, Akio; Kohzaki, Yasuji; Nakao, Yasuyuki; Nishikawa, Masabumi; Ohi, Shoichi; Ohnishi, Masami.

    1992-09-01

    A neutron-lean D- 3 He fueled field reversed configuration (FRC) fusion reactor is studied on the bases of former high-efficiency ARTEMIS design. Certain improvements such as effective axial contracting plasma heating and cusp-type direct energy converters as well as an empirical scale of the energy confinement are introduced. The resultant total neutron load onto the first wall of the plasma chamber is as low as 0.1 MW/m 2 , which enable the life of the first wall or the structural materials to be longer than the whole life of the reactor. The attractive characteristics of the neutron-lean reactor follow in the ARTEMIS design: it is socially acceptable in views of radioactivity and fuel resources, and the cost of electricity appears to be cheap compared with that from a light water reactor. Critical physics and engineering issues for performing the ARTEMIS-L reactor are clarified. (author)

  14. New version of the reactor dynamics code DYN3D for Sodium cooled Fast Reactor analyses

    Energy Technology Data Exchange (ETDEWEB)

    Nikitin, Evgeny [Ecole Polytechnique Federale de Lausanne (Switzerland); Helmholtz-Zentrum Dresden-Rossendorf (HZDR) e.V., Dresden (Germany); Fridman, Emil; Bilodid, Yuri; Kliem, Soeren [Helmholtz-Zentrum Dresden-Rossendorf (HZDR) e.V., Dresden (Germany)

    2017-07-15

    The reactor dynamics code DYN3D being developed at the Helmholtz-Zentrum Dresden-Rossendorf is currently under extension for Sodium cooled Fast Reactor analyses. This paper provides an overview on the new version of DYN3D to be used for SFR core calculations. The current article shortly describes the newly implemented thermal mechanical models, which can account for thermal expansion effects of the reactor core. Furthermore, the methodology used in Sodium cooled Fast Reactor analyses to generate homogenized few-group cross sections is summarized. The conducted and planned verification and validation studies are briefly presented. Related publications containing more detailed descriptions are outlined for the completeness of this overview.

  15. Integral test of JENDL-3.3 on fast reactors

    International Nuclear Information System (INIS)

    Chiba, Gou; Hazama, Taira

    2003-05-01

    An integral test has been carried out to evaluate a performance of evaluated nuclear data library JENDL-3.3, which was newly released, in a view of applying neutronics analyses of fast reactors. Japan Nuclear Cycle Development Institute has a large amount of data of critical assembly experiments (ZPPR, BFS, MOZART and FCA) and power reactor tests (JOYO). The database was utilized in this test. In plutonium loaded cores, an improvement was observed about 0.3% ε k in criticality and 5% in the non-leakage term of sodium void reactivity by a revision form JENDL-3.2 to -3.3. These results shoed that the revision is valid in plutonium loaded cores. In uranium loaded cores, dependence of C/E values on control rod position became smaller in control rod worth in ZPPR cores. On the other hand, C/E values became worse both in criticality (0.6%εk) and in sodium void reactivity (30%) in BFS cores. The main cause was a revision of uranium-235 capture cross section, and it could not be concluded whether the revision is valid or not in uranium loaded cores. It is necessary to carry out a validation test at other independent critical experiments in which uranium fuel is used. (author)

  16. Safety analysis of loss of flow transients in a typical research reactor by RELAP5/MOD3.3

    International Nuclear Information System (INIS)

    Di Maro, B.; Pierro, F.; Adorni, M.; Bousbia Salah, A.; D'Auria, F.

    2003-01-01

    The main aim of the following study is to assess the RELAP5/MOD3.3 code capability in simulating transient dynamic behaviour in nuclear research reactors. For this purpose typical loss of flow transient in a representative MTR (Metal Test Reactor) fuel type Research Reactor is considered. The transient herein considered is a sudden pump trip followed by the opening of a safety valve in order to allow passive decay heat removal by natural convection. During such transient the coolant flow decay, originally downward, leads to a flow reversal and the cooling process of the core passes from forced, mixed and finally to natural circulation. This fact makes it suitable for evaluating the new features of RELAP5 to simulate such specific operating conditions. The instantaneous reactor power is derived through the point kinetic calculation, both protected and unprotected cases are considered (with and without Scram). The results obtained from this analysis were also compared with previous results obtained by old version RELAP5/MOD2 code. (author)

  17. Gas Reactor International Cooperative program. Pebble bed reactor plant: screening evaluation. Volume 3. Appendix A. Equipment list

    International Nuclear Information System (INIS)

    1979-11-01

    This report consists of three volumes which describe the design concepts and screening evaluation for a 3000 MW(t) Pebble Bed Reactor Multiplex Plant (PBR-MX). The Multiplex plant produces both electricity and transportable chemical energy via the thermochemical pipeline (TCP). The evaluation was limited to a direct cycle plant which has the steam generators and steam reformers in the primary circuit. Volume 1 reports the overall plant and reactor system and was prepared by the General Electric Company. Core scoping studies were performed which evaluated the effects of annular and cylindrical core configurations, radial blanket zones, burnup, and ball heavy metal loadings. The reactor system, including the PCRV, was investigated for both the annular and cylindrical core configurations. Volume 3 is an Appendix containing the equipment list for the plant and was also prepared by United Engineers and Constructors, Inc. It tabulates the major components of the plant and describes each in terms of quantity, type, orientation, etc., to provide a basis for cost estimation

  18. A Small-Animal Irradiation Facility for Neutron Capture Therapy Research at the RA-3 Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Emiliano Pozzi; David W. Nigg; Marcelo Miller; Silvia I. Thorp; Amanda E. Schwint; Elisa M. Heber; Veronica A. Trivillin; Leandro Zarza; Guillermo Estryk

    2007-11-01

    The National Atomic Energy Commission of Argentina (CNEA) has constructed a thermal neutron source for use in Boron Neutron Capture Therapy (BNCT) applications at the RA-3 research reactor facility located in Buenos Aires. The Idaho National Laboratory (INL) and CNEA have jointly conducted some initial neutronic characterization measurements for one particular configuration of this source. The RA-3 reactor (Figure 1) is an open pool type reactor, with 20% enriched uranium plate-type fuel and light water coolant. A graphite thermal column is situated on one side of the reactor as shown. A tunnel penetrating the graphite structure enables the insertion of samples while the reactor is in normal operation. Samples up to 14 cm height and 15 cm width are accommodated.

  19. Program ELM: A tool for rapid thermal-hydraulic analysis of solid-core nuclear rocket fuel elements

    International Nuclear Information System (INIS)

    Walton, J.T.

    1992-11-01

    This report reviews the state of the art of thermal-hydraulic analysis codes and presents a new code, Program ELM, for analysis of fuel elements. ELM is a concise computational tool for modeling the steady-state thermal-hydraulics of propellant flow through fuel element coolant channels in a nuclear thermal rocket reactor with axial coolant passages. The program was developed as a tool to swiftly evaluate various heat transfer coefficient and friction factor correlations generated for turbulent pipe flow with heat addition which have been used in previous programs. Thus, a consistent comparison of these correlations was performed, as well as a comparison with data from the NRX reactor experiments from the Nuclear Engine for Rocket Vehicle Applications (NERVA) project. This report describes the ELM Program algorithm, input/output, and validation efforts and provides a listing of the code

  20. 3D CAD model of the subcritical nuclear reactor of IPN

    International Nuclear Information System (INIS)

    Pahuamba V, F. de J.; Delfin L, A.; Gomez T, A.; Ibarra R, G.; Del Valle G, E.; Sanchez R, A.

    2016-09-01

    The three-dimensional (3D) CAD model of the subcritical reactor Chicago model 9000 of Instituto Politecnico Nacional (IPN) allows obtaining a 3D view with the dimensions of each of its components, such as: natural uranium cylindrical rods, fuel elements, hexagonal reactor core arrangement, cylindrical stainless steel tank containing the core, fuel element support grids and reactor water cleaning system. As a starting point for the development of the model, the Chicago model 9000 subcritical reactor manual provided by the manufacturer was used, the measurement and verification of the components to adapt the geometric, physical and mechanical characteristics was carried out and materials standards were used to obtain a design that allows to elaborate a new manual according to the specifications. In addition, the 3D models of the building of the Advanced Physics Laboratory, neutron generator, cobalt source and the corridors connecting to the subcritical reactor facility were developed, allowing an animated ride, developed by computer-aided design software. The manual provided by the company Nuclear Chicago, dates from the year 1959 and presents diverse deviations in the design and dimensions of the reactor components. The model developed; in addition to supporting the development of the new manual represents a learning tool to visualize the reactor components. (Author)

  1. 3 Investment Scenarios for Fast Reactors

    International Nuclear Information System (INIS)

    Shoai Tehrani, Bianka; Da Costa, Pascal

    2013-01-01

    Results: • 4 families of scenarios: – In each of them, 3 options for national nuclear policy → 12 scenarios; – 3 favorable to FRs: - “climate constraint” with strong pro-nuclear policy - “climate constraint” with moderate pro-nuclear policy - “totally green” with strong pro-nuclear policy. • Business As Usual is not favorable to Fast Reactors; Fast reactors deployment: - Needs strong climate policy - Is viable in case of important renewable progress as long as climate policy is strong. International perspective: • Results are valid for Europe, other drivers being likely to be more important in other countries : high growth and demand (Asia); • With strong contrasts between European countries. Further research: • Finer modeling of drivers with unclear influence (clustered and excluded variables): Influence of weak signals

  2. Investigations related to a one-piece removal of the reactor block in the frame of the JRR-3 reconstruction program

    International Nuclear Information System (INIS)

    Onishi, N.; Kanenari, A.; Futamura, Y.; Sakurai, H.; Suzuki, S.; Nagase, T.; Iwatani, A.; Otsubo, F.

    1987-01-01

    In the Japan Atomic Energy Research Institute (JAERI), an outdated research reactor (Japan Research Reactor No.3; JRR-3) was removed to a storage facility between October 14th and November 7th, 1986. The removal of the 2250-ton reactor block (10 x 10 x 10 m) was performed as a part of a program to replace the JRR-3's core (10-MW thermal) with an upgraded research reactor core. The heavy water and fuel elements were taken out from the JRR-3 before removal work began. The reactor block was raised about 3.7 meters, using a 12-cubic meter steel frame and a center-hole jack system. The reactor block was then transported horizontally about 34 meters on steel rails, using four 100-ton jacks, to a storage facility. Finally, the reactor block was lowered 14 meters into the storage facility. After the reactor block was stored, a new 20-MW thermal, light-water moderated and cooled JRR-3 core will be built, with criticality targeted for 1989

  3. 3 D flow computations under a reactor vessel closure head

    International Nuclear Information System (INIS)

    Daubert, O.; Bonnin, O.; Hofmann, F.; Hecker, M.

    1995-12-01

    The flow under a vessel cover of a pressurised water reactor is investigated by using several computations and a physical model. The case presented here is turbulent, isothermal and incompressible. Computations are made with N3S code using a k-epsilon model. Comparisons between numerical and experimental results are on the whole satisfying. Some local improvements are expected either with more sophisticated turbulence models or with mesh refinements automatically computed by using the adaptive meshing technique which has been just implemented in N3S for 3D cases. (authors). 6 refs., 7 figs

  4. Study on the reactivity behavior partially loaded reactor cores using SIMULATE-3

    International Nuclear Information System (INIS)

    Holzer, Robert; Zeitz, Andreas; Grimminger, Werner; Lubczyk, Tobias

    2009-01-01

    The reactor core design for the NPP Gundremmingen unit B and C is performed since several years using the validated 3D reactor core calculation program SIMULATE-3. The authors describe a special application of the program to study the reactivity for different partial core loadings. Based on the comparison with results of the program CASMO-4 the program SIMULATE-3 was validated for the calculation of partially loaded reactor cores. For the planned reactor operation in NPP Gundremmingen using new MOX fuel elements the reactivity behavior was studied with respect to the KTA-Code requirements.

  5. Atmospherically dispersed radiocarbon at the Chalk River Laboratories

    International Nuclear Information System (INIS)

    Milton, G.M.; Brown, R.M.; Repta, C.J.W.; Selkirk, C.J.

    1996-01-01

    A small percentage of the total radiocarbon produced by the NRX and NRU experimental reactors at the Chalk River Laboratories has been vented from the main reactor stack and atmospherically dispersed across the site. Surveys conducted in 1982-83 and 1993-94 have shown that atmospheric levels more than 50 m from the stack are never greater than 600 Bq.kg -1 carbon above the natural background level, falling to near-global atmospheric levels at the site boundaries roughly 7 km away. A dispersion factor > 1.2 x 10 6 m 3 .s -1 at ∼ 0.75 km distance from the point of emission is calculated on the basis of recent in-stack monitoring. Analysis of growth rings in on-site trees has provided an opportunity to search for correlations of 14 C output summer power production and/or moderator losses. (author). 16 refs., 14 tabs., 11 figs

  6. Development of a 3-D flow analysis computer program for integral reactor

    International Nuclear Information System (INIS)

    Youn, H. Y.; Lee, K. H.; Kim, H. K.; Whang, Y. D.; Kim, H. C.

    2003-01-01

    A 3-D computational fluid dynamics program TASS-3D is being developed for the flow analysis of primary coolant system consists of complex geometries such as SMART. A pre/post processor also is being developed to reduce the pre/post processing works such as a computational grid generation, set-up the analysis conditions and analysis of the calculated results. TASS-3D solver employs a non-orthogonal coordinate system and FVM based on the non-staggered grid system. The program includes the various models to simulate the physical phenomena expected to be occurred in the integral reactor and will be coupled with core dynamics code, core T/H code and the secondary system code modules. Currently, the application of TASS-3D is limited to the single phase of liquid, but the code will be further developed including 2-phase phenomena expected for the normal operation and the various transients of the integrator reactor in the next stage

  7. 3D computer visualization and animation of CANDU reactor core

    International Nuclear Information System (INIS)

    Qian, T.; Echlin, M.; Tonner, P.; Sur, B.

    1999-01-01

    Three-dimensional (3D) computer visualization and animation models of typical CANDU reactor cores (Darlington, Point Lepreau) have been developed using world-wide-web (WWW) browser based tools: JavaScript, hyper-text-markup language (HTML) and virtual reality modeling language (VRML). The 3D models provide three-dimensional views of internal control and monitoring structures in the reactor core, such as fuel channels, flux detectors, liquid zone controllers, zone boundaries, shutoff rods, poison injection tubes, ion chambers. Animations have been developed based on real in-core flux detector responses and rod position data from reactor shutdown. The animations show flux changing inside the reactor core with the drop of shutoff rods and/or the injection of liquid poison. The 3D models also provide hypertext links to documents giving specifications and historical data for particular components. Data in HTML format (or other format such as PDF, etc.) can be shown in text, tables, plots, drawings, etc., and further links to other sources of data can also be embedded. This paper summarizes the use of these WWW browser based tools, and describes the resulting 3D reactor core static and dynamic models. Potential applications of the models are discussed. (author)

  8. RELAP5/MOD 3.3 analysis of Reactor Coolant Pump Trip event at NPP Krsko

    International Nuclear Information System (INIS)

    Bencik, V.; Debrecin, N.; Foretic, D.

    2003-01-01

    In the paper the results of the RELAP5/MOD 3.3 analysis of the Reactor Coolant Pump (RCP) Trip event at NPP Krsko are presented. The event was initiated by an operator action aimed to prevent the RCP 2 bearing damage. The action consisted of a power reduction, that lasted for 50 minutes, followed by a reactor and a subsequent RCP 2 trip when the reactor power was reduced to 28 %. Two minutes after reactor trip, the Main Steam Isolation Valves (MSIV) were isolated and the steam dump flow was closed. On the secondary side the Steam Generator (SG) pressure rose until SG 1 Safety Valve (SV) 1 opened. The realistic RELAP5/MOD 3.3 analysis has been performed in order to model the particular plant behavior caused by operator actions. The comparison of the RELAP5/MOD 3.3 results with the measurement for the power reduction transient has shown small differences for the major parameters (nuclear power, average temperature, secondary pressure). The main trends and physical phenomena following the RCP Trip event were well reproduced in the analysis. The parameters that have the major influence on transient results have been identified. In the paper the influence of SG 1 relief and SV valves on transient results was investigated more closely. (author)

  9. Informal presentations by fuel fabricators and others [contributed by W. Ross, U.S. NRC

    International Nuclear Information System (INIS)

    Ross, W.

    1993-01-01

    Al-28 wt.% U alloy and its ductility certainly will not increase with irradiation. The effect of this brittleness on behavior of the long NRX/NRU rods during thermal cycling in reactor would have to be investigated. Therefore from a fabrication point of view, it may be possible to make the rods from Al-40 wt.% U but it is obvious that an extensive fabrication, safety, and irradiation study would be required before a definitive answer could be given. Development of Al-50 wt.% U alloy for such rods would be even more difficult, the probability of success smaller, and the development program somewhat larger. The development of a completely new design driver fuel for the reactors using Zircaloy clad powder packed UO 2 or dispersion type fuels would require an even larger and more expensive program. At 20% enrichment level, the current Al-U designs of driver fuel could not be used even by increasing the number of fuel rods and/or by removing experimental facilities. A new fuel design would have to be tested to burnups of 60-70% for NRX and over 80% for NRU before it could be considered to be acceptable. The defect performance of the higher uranium alloys would also have to be checked out. In summary, the NRX and NRU are high performance, high flux research reactors having a very heavy experimental and radioisotope load. With the present designs of driver fuel and reactor loadings the use of 50% enriched uranium would be possible only if the brittle Al-40 to 50 wt.% U alloys can be successfully developed for high burnups. Extensive fabrication, safety, and physics work and irradiation of intact and defective elements would have to be done for each of the fuel designs considered before a definitive answer could be given. If these alloys could not be successfully developed, new designs of driver fuel would be required. The fuel developments would be in three phases: fuel alloy development, pin development, and irradiation testing and would probably take 3 to years. A change

  10. PR-EDB: Power Reactor Embrittlement Database Version 3

    International Nuclear Information System (INIS)

    Wang, Jy-An John; Subramani, Ranjit

    2008-01-01

    The aging and degradation of light-water reactor pressure vessels is of particular concern because of their relevance to plant integrity and the magnitude of the expected irradiation embrittlement. The radiation embrittlement of reactor pressure vessel materials depends on many factors, such as neutron fluence, flux, and energy spectrum, irradiation temperature, and preirradiation material history and chemical compositions. These factors must be considered to reliably predict pressure vessel embrittlement and to ensure the safe operation of the reactor. Large amounts of data from surveillance capsules are needed to develop a generally applicable damage prediction model that can be used for industry standards and regulatory guides. Furthermore, the investigations of regulatory issues such as vessel integrity over plant life, vessel failure, and sufficiency of current codes, Standard Review Plans (SRPs), and Guides for license renewal can be greatly expedited by the use of a well-designed computerized database. The Power Reactor Embrittlement Database (PR-EDB) is such a comprehensive collection of data for U.S. designed commercial nuclear reactors. The current version of the PR-EDB lists the test results of 104 heat-affected-zone (HAZ) materials, 115 weld materials, and 141 base materials, including 103 plates, 35 forgings, and 3 correlation monitor materials that were irradiated in 321 capsules from 106 commercial power reactors. The data files are given in dBASE format and can be accessed with any personal computer using the Windows operating system. 'User-friendly' utility programs have been written to investigate radiation embrittlement using this database. Utility programs allow the user to retrieve, select and manipulate specific data, display data to the screen or printer, and fit and plot Charpy impact data. The PR-EDB Version 3.0 upgrades Version 2.0. The package was developed based on the Microsoft .NET framework technology and uses Microsoft Access for

  11. PR-EDB: Power Reactor Embrittlement Database - Version 3

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [ORNL; Subramani, Ranjit [ORNL

    2008-03-01

    The aging and degradation of light-water reactor pressure vessels is of particular concern because of their relevance to plant integrity and the magnitude of the expected irradiation embrittlement. The radiation embrittlement of reactor pressure vessel materials depends on many factors, such as neutron fluence, flux, and energy spectrum, irradiation temperature, and preirradiation material history and chemical compositions. These factors must be considered to reliably predict pressure vessel embrittlement and to ensure the safe operation of the reactor. Large amounts of data from surveillance capsules are needed to develop a generally applicable damage prediction model that can be used for industry standards and regulatory guides. Furthermore, the investigations of regulatory issues such as vessel integrity over plant life, vessel failure, and sufficiency of current codes, Standard Review Plans (SRPs), and Guides for license renewal can be greatly expedited by the use of a well-designed computerized database. The Power Reactor Embrittlement Database (PR-EDB) is such a comprehensive collection of data for U.S. designed commercial nuclear reactors. The current version of the PR-EDB lists the test results of 104 heat-affected-zone (HAZ) materials, 115 weld materials, and 141 base materials, including 103 plates, 35 forgings, and 3 correlation monitor materials that were irradiated in 321 capsules from 106 commercial power reactors. The data files are given in dBASE format and can be accessed with any personal computer using the Windows operating system. "User-friendly" utility programs have been written to investigate radiation embrittlement using this database. Utility programs allow the user to retrieve, select and manipulate specific data, display data to the screen or printer, and fit and plot Charpy impact data. The PR-EDB Version 3.0 upgrades Version 2.0. The package was developed based on the Microsoft .NET framework technology and uses Microsoft Access for

  12. Nuclear research reactor 0.5 to 3 MW

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1992-05-15

    This nuclear reactor has been designed for radioisotope production, basic and applied research in reactor physics and nuclear engineering, neutron-beam experimentation, irradiation of various materials and training of scientific and technical personnel. It is located in the 'Production Area' of the Nuclear Technology Center. It is equipped with the necessary facilities for large-scale production of radioisotopes to be used in medicine as well as for other scientific and industrial purposes. In addition, it has a Neutronography Facility and the required equipment to perform Neutron-Activation Analysis. It is an open pool-type reactor, moderated and cooled with light water, fuelled with 20% enriched uranium. Its reflector are graphite and water. It has plate-type fuel elements clad in aluminium. The reactor core is located near the bottom of the demineralized water pool. It includes fuel elements, reflector and sample-holding devices for materials to be irradiated. This kind of configuration, which is widely used in research reactors, provides a high degree of safety since it prevents the core from becoming exposed under any circumstance and does not require any cooling system during reactor shutdown. Power output is between 0.5 to 3 MW{sub TH}, with a minimum thermal neutron flux of approx, 10{sup 13} n/cm{sup 2}{center_dot}sec, at irradiation zone almost with no modifications. Heat extraction is achieved by means of a cooling circuit which comprises two circulation pumps and a plate-type heat exchanger. Final heat dissipation to the atmosphere is performed through another cooling circuit which includes two circulation pumps and a cooling tower. Reactor control is accomplished with five neutron-absorbing rods positioned by means of especially designed elements and governed by the reactor's instrumentation and control system. Should an abnormal situation arise, gravity causes the rods to fall automatically, thus extinguishing the nuclear reaction. The reactor

  13. Nuclear research reactor 0.5 to 3 MW

    International Nuclear Information System (INIS)

    1992-05-01

    This nuclear reactor has been designed for radioisotope production, basic and applied research in reactor physics and nuclear engineering, neutron-beam experimentation, irradiation of various materials and training of scientific and technical personnel. It is located in the 'Production Area' of the Nuclear Technology Center. It is equipped with the necessary facilities for large-scale production of radioisotopes to be used in medicine as well as for other scientific and industrial purposes. In addition, it has a Neutronography Facility and the required equipment to perform Neutron-Activation Analysis. It is an open pool-type reactor, moderated and cooled with light water, fuelled with 20% enriched uranium. Its reflector are graphite and water. It has plate-type fuel elements clad in aluminium. The reactor core is located near the bottom of the demineralized water pool. It includes fuel elements, reflector and sample-holding devices for materials to be irradiated. This kind of configuration, which is widely used in research reactors, provides a high degree of safety since it prevents the core from becoming exposed under any circumstance and does not require any cooling system during reactor shutdown. Power output is between 0.5 to 3 MW TH , with a minimum thermal neutron flux of approx, 10 13 n/cm 2 ·sec, at irradiation zone almost with no modifications. Heat extraction is achieved by means of a cooling circuit which comprises two circulation pumps and a plate-type heat exchanger. Final heat dissipation to the atmosphere is performed through another cooling circuit which includes two circulation pumps and a cooling tower. Reactor control is accomplished with five neutron-absorbing rods positioned by means of especially designed elements and governed by the reactor's instrumentation and control system. Should an abnormal situation arise, gravity causes the rods to fall automatically, thus extinguishing the nuclear reaction. The reactor building has a ventilation

  14. Reactor Dosimetry Applications Using RAPTOR-M3G:. a New Parallel 3-D Radiation Transport Code

    Science.gov (United States)

    Longoni, Gianluca; Anderson, Stanwood L.

    2009-08-01

    The numerical solution of the Linearized Boltzmann Equation (LBE) via the Discrete Ordinates method (SN) requires extensive computational resources for large 3-D neutron and gamma transport applications due to the concurrent discretization of the angular, spatial, and energy domains. This paper will discuss the development RAPTOR-M3G (RApid Parallel Transport Of Radiation - Multiple 3D Geometries), a new 3-D parallel radiation transport code, and its application to the calculation of ex-vessel neutron dosimetry responses in the cavity of a commercial 2-loop Pressurized Water Reactor (PWR). RAPTOR-M3G is based domain decomposition algorithms, where the spatial and angular domains are allocated and processed on multi-processor computer architectures. As compared to traditional single-processor applications, this approach reduces the computational load as well as the memory requirement per processor, yielding an efficient solution methodology for large 3-D problems. Measured neutron dosimetry responses in the reactor cavity air gap will be compared to the RAPTOR-M3G predictions. This paper is organized as follows: Section 1 discusses the RAPTOR-M3G methodology; Section 2 describes the 2-loop PWR model and the numerical results obtained. Section 3 addresses the parallel performance of the code, and Section 4 concludes this paper with final remarks and future work.

  15. Characteristics of D(-3)He fueled FRC reactor: ARTEMIS-L

    Science.gov (United States)

    Momota, H.; Motojima, O.; Okamoto, M.; Sudo, S.; Tomita, Y.; Yamaguchi, S.; Iiyoshi, A.; Onozuka, M.; Ohnishi, M.; Uenosono, C.

    1993-11-01

    The characteristics of D(-3)He fueled commercial fusion reactor ARTEMIS-L are discussed. By using favorable characteristics of a field-reversed configuration, the fusion plasma of ARTEMIS-L becomes compact and its veta-value is extremely high. Consequently, it is possible to construct an economical fusion power plant based on this concept. The life of the structural materials is found during the full reactor life (30 years) and the safety of the reactor is intrinsic to D(-3)He fuels. The amount of disposed materials is rather small and the level of the intruder dose is so low that the plant appears to be acceptable in regards to the environment.

  16. Benchmark tests of JENDL-3.2 for thermal and fast reactors

    International Nuclear Information System (INIS)

    Takano, Hideki

    1995-01-01

    Benchmark calculations for a variety of thermal and fast reactors have been performed by using the newly evaluated JENDL-3 Version-2 (JENDL-3.2) file. In the thermal reactor calculations for the uranium and plutonium fueled cores of TRX and TCA, the k eff and lattice parameters were well predicted. The fast reactor calculations for ZPPR-9 and FCA assemblies showed that the k eff , reactivity worth of Doppler, sodium void and control rod, and reaction rate distribution were in a very good agreement with the experiments. (author)

  17. Development of a 3-dimensional calculation model of the Danish research reactor DR3 to analyse a proposal to a new core design called ring-core

    Energy Technology Data Exchange (ETDEWEB)

    Nonboel, E

    1985-07-01

    A 3-dimensional calculation model of the Danish research reactor DR3 has been developed. Demands of a more effective utilization of the reactor and its facilities has required a more detailed calculation tool than applied so far. A great deal of attention has been devoted to the treatment of the coarse control arms. The model has been tested against measurements with satisfying results. Furthermore the model has been used to analyse a proposal to a new core design called ring-core where 4 central fuel elements are replaced by 4 dummy elements to increase the thermal flux in the center of the reactor. (author)

  18. The design of a fuel element for the RA-3 reactor (Ezeiza Atomic Center)

    International Nuclear Information System (INIS)

    Agueda, Horacio C.; Estevez, Esteban; Gerding, Jose R.; Markiewicz, Mario E.

    2003-01-01

    Some features of the mechanical design of the low enrichment fuel element for the RA-3 reactor are described, with emphasis in those aspects of the original design that have been modified considering the experience acquired in the design of other fuel elements. The proposed modification is based fundamentally on the replacement of all welded joints by screwed joints, which facilitates the manufacture of the fuel element, avoiding the distortions produced by the welds used at present and contributing to the fulfillment of the foreseen tolerances. A basic characteristic of this design is a careful manufacture of the fuel element's structural components in order to assure an assembling of the fuel element that fulfills the tolerances intrinsically required. The fuel is designed for the RA-3 reactor and uses U 3 O 8 or U 3 Si 2 as carrying phase of the fissile material with an enrichment of 19.70% of 235 U. The design verification was performed by analytical and numerical methods, and is supported by testing of materials in laboratory, hydrodynamics tests and performance evaluations of the fuel elements in the RA-3 reactor. (author)

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  20. Analysis of a total flow blockage of a Fuel Assembly in a typical MTR Research Reactor by RELAP5/MOD3.3

    International Nuclear Information System (INIS)

    Adorni, M.; Salah, A.B.; Di Maro, B.; Pierro, F.; D'Auria, F.; Hamidouche, T.

    2004-01-01

    The lack of full understanding of complex mechanisms connected with the interaction between thermal-hydraulics and neutronics still challenge the design and the operation of nuclear reactors by the adoption of conservative safety limits. The recent availability of powerful computer and computational techniques together with the continuing increase in operational experience imposes the revisiting of those areas and eventually the identification of design/safety requirements that can be relaxed [1]. Currently, the enlarged commercial exploitation of nuclear Research Reactors (RR) has increased the consideration to their corresponding safety issues. Almost all of the safety analyses have so far been performed using conservative computational tools [2]. Nowadays, the application of Best-Estimate (BE) methods constitutes a real necessity in order to increase their commercial productivity. In this framework, an attempt is made to apply the BE technique to perform a safety evaluation under research reactors operational conditions. In fact, this technique has been largely verified and validated for power reactors using coupled system thermal-hydraulic and three-dimensional neutron kinetics [1]. For this purpose, as typical representative of research reactors, the IAEA 10 MW MTR Research Reactors problem [3] is considered. The system thermal-hydraulic RELAP5 [4] code was developed to simulate transient scenarios in Power reactors such PWR, BWR, VVER, etc. However, only limited work was performed to access the applicability of the code to Research Reactors operating conditions (low pressure, mass flow rates, power, etc) [5]. Previous works performed in this field are reported in [5], [6] and [7]. In this framework, total and partial blockage of a single Fuel Assembly cooling channel are investigated. As a first attempt the calculations are performed by applying the BE thermal-hydraulic system code RELAP5 alone using its point kinetic model to derive the instantaneous core

  1. Sensory regulation of neuroligins and neurexin I in the honeybee brain.

    Directory of Open Access Journals (Sweden)

    Sunita Biswas

    2010-02-01

    Full Text Available Neurexins and neuroligins, which have recently been associated with neurological disorders such as autism in humans, are highly conserved adhesive proteins found on synaptic membranes of neurons. These binding partners produce a trans-synaptic bridge that facilitates maturation and specification of synapses. It is believed that there exists an optimal spatio-temporal code of neurexin and neuroligin interactions that guide synapse formation in the postnatal developing brain. Therefore, we investigated whether neuroligins and neurexin are differentially regulated by sensory input using a behavioural model system with an advanced capacity for sensory processing, learning and memory, the honeybee.Whole brain expression levels of neuroligin 1-5 (NLG1-5 and neurexin I (NrxI were estimated by qRT-PCR analysis in three different behavioural paradigms: sensory deprivation, associative scent learning, and lateralised sensory input. Sensory deprived bees had a lower level of NLG1 expression, but a generally increased level of NLG2-5 and NrxI expression compared to hive bees. Bees that had undergone associative scent training had significantly increased levels of NrxI, NLG1 and NLG3 expression compared to untrained control bees. Bees that had lateralised sensory input after antennal amputation showed a specific increase in NLG1 expression compared to control bees, which only happened over time.Our results suggest that (1 there is a lack of synaptic pruning during sensory deprivation; (2 NLG1 expression increases with sensory stimulation; (3 concomitant changes in gene expression suggests NrxI interacts with all neuroligins; (4 there is evidence for synaptic compensation after lateralised injury.

  2. Summary of the 3rd workshop on the reduced-moderation water reactor

    International Nuclear Information System (INIS)

    Ishikawa, Nobuyuki; Nakatsuka, Tohru; Iwamura, Takamichi

    2000-06-01

    The research activities of a Reduced-Moderation Water Reactor (RMWR) are being performed for a development of the next generation water-cooled reactor. A workshop on the RMWR was held on March 3rd 2000 aiming to exchange information between JAERI and other organizations such as universities, laboratories, utilities and vendors. This report summarizes the contents of lectures and discussions on the workshop. The 1st workshop was held on March 1998 focusing on the review of the research activities and future research plan. The succeeding 2nd workshop was held on March 1999 focusing on the topics of the plutonium utilization in water-cooled reactors. The 3rd workshop was held on March 3rd 2000, which was attended by 77 participants. The workshop began with a lecture titled 'Recent Situation Related to Reduced-Moderation Water Reactor (RMWR)', followed by 'Program on MOX Fuel Utilization in Light Water Reactors' which is the mainstream scenario of plutonium utilization by utilities, and 'Feasibility Studies on Commercialized Fast Breeder Reactor Cycle System' mainly conducted by Japan Nuclear Cycle Development Institute (JNC). Also, following lectures were given as the recent research activities in JAERI: 'Progress in Design Study on Reduced-Moderation Water Reactors', 'Long-Term Scenarios of Power Reactors and Fuel Cycle Development and the Role of Reduced Moderation Water Reactors', 'Experimental and Analytical Study on Thermal Hydraulics' and Reactor Physics Experiment Plan using TCA'. At the end of the workshop, a general discussion was performed about the research and development of the RMWR. This report includes the original papers presented at the workshop and summaries of the questions and answers for each lecture and general discussion, as well as presentation viewgraphs, program and participant list as appendixes. The 7 of the presented papers are indexed individually. (J.P.N.)

  3. Summary of the 3rd workshop on the reduced-moderation water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ishikawa, Nobuyuki; Nakatsuka, Tohru; Iwamura, Takamichi [eds.

    2000-06-01

    The research activities of a Reduced-Moderation Water Reactor (RMWR) are being performed for a development of the next generation water-cooled reactor. A workshop on the RMWR was held on March 3rd 2000 aiming to exchange information between JAERI and other organizations such as universities, laboratories, utilities and vendors. This report summarizes the contents of lectures and discussions on the workshop. The 1st workshop was held on March 1998 focusing on the review of the research activities and future research plan. The succeeding 2nd workshop was held on March 1999 focusing on the topics of the plutonium utilization in water-cooled reactors. The 3rd workshop was held on March 3rd 2000, which was attended by 77 participants. The workshop began with a lecture titled 'Recent Situation Related to Reduced-Moderation Water Reactor (RMWR)', followed by 'Program on MOX Fuel Utilization in Light Water Reactors' which is the mainstream scenario of plutonium utilization by utilities, and 'Feasibility Studies on Commercialized Fast Breeder Reactor Cycle System' mainly conducted by Japan Nuclear Cycle Development Institute (JNC). Also, following lectures were given as the recent research activities in JAERI: 'Progress in Design Study on Reduced-Moderation Water Reactors', 'Long-Term Scenarios of Power Reactors and Fuel Cycle Development and the Role of Reduced Moderation Water Reactors', 'Experimental and Analytical Study on Thermal Hydraulics' and Reactor Physics Experiment Plan using TCA'. At the end of the workshop, a general discussion was performed about the research and development of the RMWR. This report includes the original papers presented at the workshop and summaries of the questions and answers for each lecture and general discussion, as well as presentation viewgraphs, program and participant list as appendixes. The 7 of the presented papers are indexed individually. (J.P.N.)

  4. Experimental and kinetic modeling study of 3-methylheptane in a jet-stirred reactor

    KAUST Repository

    Karsenty, Florent

    2012-08-16

    Improving the combustion of conventional and alternative fuels in practical applications requires the fundamental understanding of large hydrocarbon combustion chemistry. The focus of the present study is on a high-molecular-weight branched alkane, namely, 3-methylheptane, oxidized in a jet-stirred reactor. This fuel, along with 2-methylheptane, 2,5-dimethylhexane, and n-octane, are candidate surrogate components for conventional diesel fuels derived from petroleum, synthetic Fischer-Tropsch diesel and jet fuels derived from coal, natural gas, and/or biomass, and renewable diesel and jet fuels derived from the thermochemical treatment of bioderived fats and oils. This study presents new experimental results along with a low- and high-temperature chemical kinetic model for the oxidation of 3-methylheptane. The proposed model is validated against these new experimental data from a jet-stirred reactor operated at 10 atm, over the temperature range of 530-1220 K, and for equivalence ratios of 0.5, 1, and 2. Significant effort is placed on the understanding of the effects of methyl substitution on important combustion properties, such as fuel reactivity and species formation. It was found that 3-methylheptane reacts more slowly than 2-methylheptane at both low and high temperatures in the jet-stirred reactor. © 2012 American Chemical Society.

  5. Characteristics of D-3He fueled frc reactor: ARTEMIS-L

    International Nuclear Information System (INIS)

    Momota, H.; Motojima, O.; Okamoto, M.; Sudo, S.; Tomita, Y.; Yamaguchi, S.; Iiyoshi, A.; Onozuka, M.; Ohnishi, M.; Uenosono, C.

    1993-11-01

    The paper introduces briefly the scenario and discuss the attractive characteristics of D-3He fueled commercial fusion reactor ARTEMIS-L. By using favorable characteristics of a field-reversed configuration, the fusion plasma of ARTEMIS-L is compact and its beta-value is extremely high. One find consequently a possibility of constructing an economical fusion power power plant on this prospect. The life of the structural materials is sound during the full reactor life (30 years) and the safety of the reactor is intrinsic to D-3He fuels. The amount of disposed materials is rather small and the level of these intruder dose is so low that the plant appears to be acceptable in view of the environment. (author)

  6. A next-generation reactor concept: The Integral Fast Reactor (IFR)

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y.I.

    1992-01-01

    The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

  7. A next-generation reactor concept: The Integral Fast Reactor (IFR)

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y.I.

    1992-07-01

    The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

  8. A next-generation reactor concept: The Integral Fast Reactor (IFR)

    International Nuclear Information System (INIS)

    Chang, Y.I.

    1992-01-01

    The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing

  9. 3D CFD for chemical transport profiles in a rotating disk CVD reactor

    Science.gov (United States)

    Han, Jong-Hyun; Yoon, Do-Young

    2010-06-01

    The RDCVD (Rotating Disk Chemical Vapor Deposition) technique is an appropriate method for uniform deposition of grains, such as compound semiconductior materials. The substrate temperature and rotation speed are the major factors, which determine the thickness uniformity of the deposited films. This paper investigates 3D CFD (3 Dimensional Computational Fluid Dynamics) simulation results of flow and heat transfer in a reactor of RDCVD using Fluent. In order to establish the reducibility of buoyancy effect on deposition quality, the chemical transport profile upon the disk heated is examined successfully in 3D domain for different rotating speeds. The resulting vortex flows due the simultaneous buoyance and centrifuge are discussed qualitatively in the 3D virtual system of a RDCVD reactor. 3D CFD is even more effective to describe the internal vortex flows due to the competitive inlet, buoyancy and centrifuge flows, which cannot be realized in the general 2D (2 Dimensional) CFD.[Figure not available: see fulltext.

  10. Use of plate fuel elements for the RA3 reactor

    International Nuclear Information System (INIS)

    Parodi, C.; Parkanski, D.; Higa, M.; Marajofsky, A.

    1992-01-01

    The RA3 reactor is a pool reactor, redesigned for 5 MW dissipation. Nineteen plates are used in each fuel element. The utilization of 20% enriched U, gives the possibility of the development of rod type fuel with Al/U 3 O 8 cermets. The thermohydraulic and neutronic conditions are studied in this work in order to satisfy the stipulated power. In addition, the fabrication conditions of Al/U 3 O 8 and Al/U 3 O 8 /Zr H 2 cermets with densities within the limits imposed by the thermohydraulics and neutronics conditions are studied. (author)

  11. Simulation of a reactor FBR with hexagonal-Z geometry using the code PARCS 3.1; Simulacion de un reactor FBR con geometria hexagonal-Z usando el codigo PARCS 3.1

    Energy Technology Data Exchange (ETDEWEB)

    Reyes F, M. C.; Del Valle G, E. [IPN, Escuela Superior de Fisica y Matematicas, Av. Instituto Politecnico Nacional s/n, U.P. Adolfo Lopez Mateos, Edificio 9, Col. San Pedro Zacatenco, 07738 Mexico D. F. (Mexico); Filio L, C., E-mail: rf.melisa@gmail.com [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Jose Ma. Barragan No. 779, Col. Narvarte, 03020 Mexico D. F. (Mexico)

    2013-10-15

    The nuclear reactor core type FBR (Fast Breeder Reactor) was modeled in three dimensions of hexagonal-Z geometry using the code PARCS (Purdue Advanced Reactor Core Simulator) version 3.1 developed by Purdue University researchers. To carry out the modeling of the mentioned reactor was taken the corresponding information to one of the described benchmarks in the document NEACRP-L-330 (3-D Neutron Transport Benchmarks, 1991); fundamentally the corresponding to the geometric data and the cross sections. Being a quick reactor of breeding, known as the Knk-II, for which are considered 4 energy groups without dispersions up. The reactor core is formed by prismatic elements of hexagonal transversal cut where part of them only corresponds to nuclear fuel assemblies. This has four reflector rings and 6 identical control elements that together with the active part of the core is configured with 8 different types of elements.With the extracted information of the mentioned document the entrance file was prepared for PARCS 3.1 only considering a sixth part of the core due to the symmetry that presents their configuration. The NEACRP-L-330 shows a wide range of results reported by those who collaborated in its elaboration using different solution techniques that go from the Monte Carlo method to the approaches S{sub 2} and P{sub 1}. Of all the results were selected those obtained with the code HEXNOD, to which were carried out a comparison of the effective multiplication factor, being smaller differences to the 300 pcm, for three different scenarios: a) with the control bars extracted totally, b) with the semi-inserted control bars and c) with the control bars inserted completely and two different axial meshes, a thick mesh with 14 slices and another fine with 38, that which implies that the results can be considered very similar among if same. Radial maps and axial profiles are included, as much of the power as of the neutrons flow. (Author)

  12. Estimation of reactor pool water temperature after shutdown in JRR-3M

    International Nuclear Information System (INIS)

    Yagi, Masahiro; Sato, Mitsugu; Kakefuda, Kazuhiro

    1999-01-01

    The reactor pool water temperature increasing by the decay heat was estimated by calculation. The reactor pool water temperature was calculated by increased enthalpy that was estimated by the reactor decay heat, the heat released from the reactor biological shielding concrete, reactor pool water surface, the heat conduction from the canal and the core inlet piping. These results of calculation were compared with the past measured data. As the results of estimation, after the JRR-3M shutdown, the calculated reactor pool temperature first increased sharply. This is because the decay heat was the major contribution. And then, rate of increased reactor pool temperature decreased. This is because the ratio of heat released from reactor biological shielding concrete and core inlet piping to the decay heat increased. Besides, the calculated reactor pool water temperature agreed with the past measured data in consequence of correcting the decay heat and the released heat. The corrected coefficient k 1 of decay heat was 0.74 - 0.80. And the corrected coefficient k 2 of heat released from the reactor biological shielding concrete was 3.5 - 4.5. (author)

  13. Development of a 3D-Multigroup program to simulate anomalous diffusion phenomena in the nuclear reactors

    International Nuclear Information System (INIS)

    Maleki Moghaddam, Nader; Afarideh, Hossein; Espinosa-Paredes, Gilberto

    2015-01-01

    Highlights: • The new version of neutron diffusion equation for simulating anomalous diffusion is presented. • Application of fractional calculus in the nuclear reactor is revealed. • A 3D-Multigroup program is developed based on the fractional operators. • The super-diffusion and sub-diffusion phenomena are modeled in the nuclear reactors core. - Abstract: The diffusion process is categorized in three parts, normal diffusion, super-diffusion and sub-diffusion. The classical neutron diffusion equation is used to model normal diffusion. A new scheme of derivatives is required to model anomalous diffusion phenomena. The fractional space derivatives are employed to model anomalous diffusion processes where a plume of particles spreads at an inconsistent rate with the classical Brownian motion model. In the fractional diffusion equation, the fractional Laplacians are used; therefore the statistical jump length of neutrons is unrestricted. It is clear that the fractional Laplacians are capable to model the anomalous phenomena in nuclear reactors. We have developed a NFDE-3D (neutron fractional diffusion equation) as a core calculation code to model normal and anomalous diffusion phenomena. The NFDE-3D is validated against the LMW-LWR reactor. The results demonstrate that reactors exhibit complex behavior versus order of the fractional derivatives which depends on the competition between neutron absorption and super-diffusion phenomenon

  14. Modelling of MOCVD Reactor: New 3D Approach

    Science.gov (United States)

    Raj, E.; Lisik, Z.; Niedzielski, P.; Ruta, L.; Turczynski, M.; Wang, X.; Waag, A.

    2014-04-01

    The paper presents comparison of two different 3D models of vertical, rotating disc MOCVD reactor used for 3D GaN structure growth. The first one is based on the reactor symmetry, while the second, novel one incorporates only single line of showerhead nozzles. It is shown that both of them can be applied interchangeably regarding the phenomena taking place within the processing area. Moreover, the importance of boundary conditions regarding proper modelling of showerhead cooling and the significance of thermal radiation on temperature field within the modelled structure are presented and analysed. The last phenomenon is erroneously neglected in most of the hitherto studies.

  15. Modelling of MOCVD reactor: new 3D approach

    International Nuclear Information System (INIS)

    Raj, E; Lisik, Z; Niedzielski, P; Ruta, L; Turczynski, M; Wang, X; Waag, A

    2014-01-01

    The paper presents comparison of two different 3D models of vertical, rotating disc MOCVD reactor used for 3D GaN structure growth. The first one is based on the reactor symmetry, while the second, novel one incorporates only single line of showerhead nozzles. It is shown that both of them can be applied interchangeably regarding the phenomena taking place within the processing area. Moreover, the importance of boundary conditions regarding proper modelling of showerhead cooling and the significance of thermal radiation on temperature field within the modelled structure are presented and analysed. The last phenomenon is erroneously neglected in most of the hitherto studies.

  16. Extension of the reactor dynamics code MGT-3D for pebblebed and blocktype high-temperature-reactors

    International Nuclear Information System (INIS)

    Shi, Dunfu

    2015-01-01

    The High Temperature Gas cooled Reactor (HTGR) is an improved, gas cooled nuclear reactor. It was chosen as one of the candidates of generation IV nuclear plants [1]. The reactor can be shut down automatically because of the negative reactivity feedback due to the temperature's increasing in designed accidents. It is graphite moderated and Helium cooled. The residual heat can be transferred out of the reactor core by inactive ways as conduction, convection, and thermal radiation during the accident. In such a way, a fuel temperature does not go beyond a limit at which major fission product release begins. In this thesis, the coupled neutronics and fluid mechanics code MGT-3D used for the steady state and time-dependent simulation of HTGRs, is enhanced and validated [2]. The fluid mechanics part is validated by SANA experiments in steady state cases as well as transient cases. The fuel temperature calculation is optimized by solving the heat conduction equation of the coated particles. It is applied in the steady state and transient simulation of PBMR, and the results are compared to the simulation with the old overheating model. New approaches to calculate the temperature profile of the fuel element of block-type HTGRs, and the calculation of the homogeneous conductivity of composite materials are introduced. With these new developments, MGT-3D is able to simulate block-type HTGRs as well. This extended MGT-3D is used to simulate a cuboid ceramic block heating experiment in the NACOK-II facility. The extended MGT-3D is also applied to LOFC and DLOFC simulation of GT-MHR. It is a fluid mechanics calculation with a given heat source. This calculation result of MGT-3D is verified with the calculation results of other codes. The design of the Japanese HTTR is introduced. The deterministic simulation of the LOFC experiment of HTTR is conducted with the Monte-Carlo code Serpent and MGT-3D, which is the LOFC Project organized by OECD/NEA [3]. With Serpent the burnup

  17. Tritium production, management and its impact on safety for a D-3He fusion reactor

    International Nuclear Information System (INIS)

    Sze, D.K.; Herring, S.; Sawan, M.

    1991-11-01

    About three percent of the fusion energy produced by a D- 3 He reactor is in the form of neutrons. Those neutrons are generated by D-D and D-T reactions, with the tritium produced by the D-D fusion. The neutrons will react with structural steel, deuterium, 3 He and shielding material to produce tritium. About half of the tritium generated by the D-D reaction will not burn in the plasma and will exit as a part of the plasma exhaust. Thus, there is enough tritium produced in a D- 3 He reactor and careful management will be required. The tritium produced in the shield and plasma can be managed with an acceptable effect on cost and safety. 3 refs., 2 figs., 3 tabs

  18. Operating reactors licensing actions summary. Vol. 3, No. 3

    International Nuclear Information System (INIS)

    1983-04-01

    The operating reactors licensing actions summary is designed to provide the management of the Nuclear Regularory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Management and Program Analysis. This summary report is published primarily for internal NRC use in managing the operating reactors licensing actions program

  19. Development of the fast reactor group constant set JFS-3-J3.2R based on the JENDL-3.2

    CERN Document Server

    Chiba, G

    2002-01-01

    It is reported that the fast reactor group constant set JFS-3-J3.2 based on the newest evaluated nuclear data library JENDL3.2 has a serious error in the process of applying the weighting function. As the error affects greatly nuclear characteristics, and a corrected version of the reactor constant set, JFS-3-J3.2R, was developed, as well as lumped FP cross sections. The use of JFS-3-J3.2R improves the results of analyses especially on sample Doppler reactivity and reaction rate in the blanket region in comparison with those obtained using the JFS-3-J3.2.

  20. Large-signal, dynamic simulation of the slowpoke-3 nuclear heating reactor

    International Nuclear Information System (INIS)

    Tseng, C.M.; Lepp, R.M.

    1983-07-01

    A 2 MWt nuclear reactor, called SLOWPOKE-3, is being developed at the Chalk River Nuclear Laboratories (CRNL). This reactor, which is cooled by natural circulation, is designed to produce hot water for commercial space heating and perhaps generate some electricity in remote locations where the costs of alternate forms of energy are high. A large-signal, dynamic simulation of this reactor, without closed-loop control, was developed and implemented on a hybrid computer, using the basic equations of conservation of mass, energy and momentum. The natural circulation of downcomer flow in the pool was simulated using a special filter, capable of modelling various flow conditions. The simulation was then used to study the intermediate and long-term transient response of SLOWPOKE-3 to large disturbances, such as loss of heat sink, loss of regulation, daily load following, and overcooling of the reactor coolant. Results of the simulation show that none of these disturbances produce hazardous transients

  1. Development of 3D CFD simulation method in nuclear reactor safety analysis

    International Nuclear Information System (INIS)

    Rosli Darmawan; Mariah Adam

    2012-01-01

    One of the most prevailing issues in the operation of nuclear reactor is the safety of the system. Worldwide publicity on a few nuclear accidents as well as the notorious Hiroshima and Nagasaki bombing have always brought about public fear on anything related to nuclear. Most findings on the nuclear reactor accidents are closely related to the reactor cooling system. Thus, the understanding of the behaviour of reactor cooling system is very important to ensure the development and improvement on safety can be continuously done. Throughout the development of nuclear reactor technology, investigation and analysis on reactor safety have gone through several phases. In the early days, analytical and experimental methods were employed. For the last three decades 1D system level codes were widely used. The continuous development of nuclear reactor technology has brought about more complex system and processes of nuclear reactor operation. More detailed dimensional simulation codes are needed to assess these new reactors. This paper discusses the development of 3D CFD usage in nuclear reactor safety analysis worldwide. A brief review on the usage of CFD at Malaysia's Reactor TRIGA PUSPATI is also presented. (author)

  2. Simulation of a reactor FBR with hexagonal-Z geometry using the code PARCS 3.1

    International Nuclear Information System (INIS)

    Reyes F, M. C.; Del Valle G, E.; Filio L, C.

    2013-10-01

    The nuclear reactor core type FBR (Fast Breeder Reactor) was modeled in three dimensions of hexagonal-Z geometry using the code PARCS (Purdue Advanced Reactor Core Simulator) version 3.1 developed by Purdue University researchers. To carry out the modeling of the mentioned reactor was taken the corresponding information to one of the described benchmarks in the document NEACRP-L-330 (3-D Neutron Transport Benchmarks, 1991); fundamentally the corresponding to the geometric data and the cross sections. Being a quick reactor of breeding, known as the Knk-II, for which are considered 4 energy groups without dispersions up. The reactor core is formed by prismatic elements of hexagonal transversal cut where part of them only corresponds to nuclear fuel assemblies. This has four reflector rings and 6 identical control elements that together with the active part of the core is configured with 8 different types of elements.With the extracted information of the mentioned document the entrance file was prepared for PARCS 3.1 only considering a sixth part of the core due to the symmetry that presents their configuration. The NEACRP-L-330 shows a wide range of results reported by those who collaborated in its elaboration using different solution techniques that go from the Monte Carlo method to the approaches S 2 and P 1 . Of all the results were selected those obtained with the code HEXNOD, to which were carried out a comparison of the effective multiplication factor, being smaller differences to the 300 pcm, for three different scenarios: a) with the control bars extracted totally, b) with the semi-inserted control bars and c) with the control bars inserted completely and two different axial meshes, a thick mesh with 14 slices and another fine with 38, that which implies that the results can be considered very similar among if same. Radial maps and axial profiles are included, as much of the power as of the neutrons flow. (Author)

  3. Integral test of JENDL-3.3 for thermal reactors

    International Nuclear Information System (INIS)

    Okumura, Keisuke; Mori, Takamasa

    2003-01-01

    Criticality benchmark testing was carried out for 59 experiments in various thermal reactors using a continues-energy Monte Carlo code MVP and its different libraries generated from JENDL-3.2, JENDL-3.3, JEF-2.2 and ENDF/B-VI (R8). From the benchmark results, we can say JENDL-3.3 generally gives better k eff values compared with other nuclear data libraries. However, further modification of JENDL-3.3 is expected to solve the following problems: 1) systematic underestimation of k eff depending on 235 U enrichment for the cores with low (less than 3wt.%) enriched uranium fueled cores, 2) dependence of C/E value of k eff on neutron spectrum and plutonium composition for MOX fueled cores. These are common problems for all of the nuclear data libraries used in this study. (author)

  4. EL-3 dismantling of an experimental reactor

    International Nuclear Information System (INIS)

    1989-01-01

    The EL3 experimental reactor has been definitively stopped in march 1979. Its decommissioning has been pronounced in the end of 1982. This article is consecrated at decontamination and dismantling works necessited by its passage at the dismantling level 2 [fr

  5. Analysis of a possible experimental assessment of a prototype fuel element containing burnable poison in the RA-3 reactor

    International Nuclear Information System (INIS)

    Lerner, Ana Maria; Madariaga, Marcelo

    2002-01-01

    The Argentine RA-3 research reactor (5 MW) is presently operated with LEU fuel by the National Atomic Energy Commission (CNEA). It belongs to the group of nuclear installations controlled, from the radiological and nuclear safety point of view, by the Nuclear Regulatory Authority (ARN). A new type of fuel elements containing burnable absorbers, with similar enrichment as the standard fuel elements but greater fissile contents, has recently been proposed for a new Argentine reactor design (RRR). In this framework the ARN considers interesting, if technically possible, the performance of an experiment in the RA-3 reactor. The experiment might enable, for such fuel element containing burnable poison, the verification of its neutronic behaviour under irradiation as well as a validation of the calculation line by comparison to measured values. It should be desirable that such experiment could reproduce as much as possible those conditions estimated for the RRR reactor, still under design in Argentina, having Silicide fuel elements with burnable poison, in the shape of cadmium wires in their structure. We here analyse a possible experiment consisting in the loading of a prototype fuel element with burnable poison in a normally loaded RA-3 core configuration. It would essentially be a standard RA-3 fuel element, having cadmium wires in its frame. This experiment would enable the verification of the prototype behaviour under irradiation, its operation limits and conditions, and particularly, the reactivity safety margins established in Argentine Standards, both calculated and measured. The main part of the experiment would imply some 200 full power days of operation at 5 MW, which would be drastically reduced if the reactor power is increased to 10 MW, as foreseen. We also show that under the proposed conditions, the experiment would not represent a significant penalty to the reactor normal operation. (author)

  6. Fusion reactor control study. Volume 3. Tandem mirror reactors. Final report

    International Nuclear Information System (INIS)

    Chang, F.R.; DeCanio, F.; Fisher, J.L.; Madden, P.A.

    1982-03-01

    A study of the control requirements of the Tandem Mirror Reactor concept is reported. The study describes the development of a control simulator that is based upon a spatially averaged physics code of the reactor concept. The simulator portrays the evolution of the plasma through the complete reactor operating cycle; it includes models of the control and measurement system, thus allowing the exploration of various strategies for reactor control. Startup, shutdown, and control during the quasi-steady-state power producing phase were explored. Configurations are described which use a variety of control effectors including modulation of the refueling rate, beam current, and electron cyclotron resonance heating. Multivariable design techniques were used to design the control laws and compensators for the feedback controllers and presume the practical measurement of only a subset of the plasma and machine variables. Performance of the various controllers is explored using the nonlinear control simulator. Derivative control strategies using new or developed sensors and effectors appropriate to a power reactor environment are postulated, based upon the results of the control configurations tested. Research and development requirements for these controls are delineated

  7. Systems analysis of the CANDU 3 Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wolfgong, J.R.; Linn, M.A.; Wright, A.L.; Olszewski, M.; Fontana, M.H. [Oak Ridge National Lab., TN (United States)

    1993-07-01

    This report presents the results of a systems failure analysis study of the CANDU 3 reactor design; the study was performed for the US Nuclear Regulatory Commission. As part of the study a review of the CANDU 3 design documentation was performed, a plant assessment methodology was developed, representative plant initiating events were identified for detailed analysis, and a plant assessment was performed. The results of the plant assessment included classification of the CANDU 3 event sequences that were analyzed, determination of CANDU 3 systems that are ``significant to safety,`` and identification of key operator actions for the analyzed events.

  8. A 3D transport-based core analysis code for research reactors with unstructured geometry

    International Nuclear Information System (INIS)

    Zhang, Tengfei; Wu, Hongchun; Zheng, Youqi; Cao, Liangzhi; Li, Yunzhao

    2013-01-01

    Highlights: • A core analysis code package based on 3D neutron transport calculation in complex geometry is developed. • The fine considerations on flux mapping, control rod effects and isotope depletion are modeled. • The code is proved to be with high accuracy and capable of handling flexible operational cases for research reactors. - Abstract: As an effort to enhance the accuracy in simulating the operations of research reactors, a 3D transport core analysis code system named REFT was developed. HELIOS is employed due to the flexibility of describing complex geometry. A 3D triangular nodal S N method transport solver, DNTR, endows the package the capability of modeling cores with unstructured geometry assemblies. A series of dedicated methods were introduced to meet the requirements of research reactor simulations. Afterwards, to make it more user friendly, a graphical user interface was also developed for REFT. In order to validate the developed code system, the calculated results were compared with the experimental results. Both the numerical and experimental results are in close agreement with each other, with the relative errors of k eff being less than 0.5%. Results for depletion calculations were also verified by comparing them with the experimental data and acceptable consistency was observed in results

  9. A new MTR fuel for a new MTR reactor: UMo for the Jules Horowitz reactor

    International Nuclear Information System (INIS)

    Guigon, B.; Vacelet, H.; Dornbusch, D.

    2000-01-01

    Within some years, the Jules Horowitz Reactor will be the only working experimental reactor (material and fuel testing reactor) in France. It will have to provide facilities for a wide range of needs from activation analysis to power reactor fuel qualification. In this paper the main characteristics of the Jules Horowitz Reactor are presented. Safety criteria are explained. Finally, merits and disadvantages of UMo compared to the standard U 3 Si 2 fuel are discussed. (author)

  10. Studies on the liquid fluoride thorium reactor: Comparative neutronics analysis of MCNP6 code with SRAC95 reactor analysis code based on FUJI-U3-(0)

    Energy Technology Data Exchange (ETDEWEB)

    Jaradat, S.Q., E-mail: sqjxv3@mst.edu; Alajo, A.B., E-mail: alajoa@mst.edu

    2017-04-01

    Highlights: • The verification for FUJI-U3-(0)—a molten salt reactor—was performed. • The MCNP6 was used to study the reactor physics characteristics for FUJI-U3 type. • The results from the MCNP6 were comparable with the ones obtained from literature. - Abstract: The verification for FUJI-U3-(0)—a molten salt reactor—was performed. The reactor used LiF-BeF2-ThF4-UF4 as the mixed liquid fuel salt, and the core was graphite moderated. The MCNP6 code was used to study the reactor physics characteristics for the FUJI-U3-(0) reactor. Results for reactor physics characteristic of the FUJI-U3-(0) exist in literature, which were used as reference. The reference results were obtained using SRAC95 (a reactor analysis code) coupled with ORIGEN2 (a depletion code). Some modifications were made in the reconstruction of the FUJI-U3-(0) reactor in MCNP due to unavailability of more detailed description of the reactor core. The assumptions resulted in two representative models of the reactor. The results from the MCNP6 models were compared with the reference results obtained from literature. The results were comparable with each other, but with some notable differences. The differences are because of the approximations that were done on the SRAC95 model of the FUJI-U3 to simplify the simulation. Based on the results, it is concluded that MCNP6 code predicts well the overall simulation of neutronics analysis to the previous simulation works using SRAC95 code.

  11. Characteristics of D-{sup 3}He fueled frc reactor: ARTEMIS-L

    Energy Technology Data Exchange (ETDEWEB)

    Momota, H.; Motojima, O.; Okamoto, M.; Sudo, S.; Tomita, Y.; Yamaguchi, S.; Iiyoshi, A.; Onozuka, M.; Ohnishi, M.; Uenosono, C.

    1993-11-01

    The paper introduces briefly the scenario and discuss the attractive characteristics of D-3He fueled commercial fusion reactor ARTEMIS-L. By using favorable characteristics of a field-reversed configuration, the fusion plasma of ARTEMIS-L is compact and its beta-value is extremely high. One find consequently a possibility of constructing an economical fusion power power plant on this prospect. The life of the structural materials is sound during the full reactor life (30 years) and the safety of the reactor is intrinsic to D-3He fuels. The amount of disposed materials is rather small and the level of these intruder dose is so low that the plant appears to be acceptable in view of the environment. (author).

  12. RELAP5-3D code validation of RBMK-1500 reactor reactivity measurement transients

    International Nuclear Information System (INIS)

    Kaliatka, Algirdas; Bubelis, Evaldas; Uspuras, Eugenijus

    2003-01-01

    This paper deals with the modeling of transients taking place during the measurements of the void and fast power reactivity coefficients performed at Ignalina NPP. The simulation of these transients was performed using RELAP5-3D code model of RBMK-1500 reactor. At the Ignalina NPP void and fast power reactivity coefficients are measured on a regular basis and, based on the total reactor power, reactivity, control and protection system control rods positions and the main circulation circuit parameter changes during the experiments, the actual values of these reactivity coefficients are determined. Following the simulation of the two above mentioned transients with RELAP5-3D code, a conclusion was made that the obtained calculation results demonstrate reasonable agreement with Ignalina NPP measured data. Behaviors of the separate MCC thermal-hydraulic parameters as well as physical processes are predicted reasonably well to the real processes, occurring in the primary circuit of RBMK-1500 reactor. The calculated reactivity and the total reactor core power behavior in time are also in reasonable agreement with the measured plant data. Despite of the small differences, RELAP5-3D code predicts reactivity and the total reactor core power behavior during the transients in a reasonable manner. Reasonable agreement of the measured and the calculated total reactor power change in time demonstrates the correct modeling of the neutronic processes taking place in RBMK-1500 reactor core

  13. Development of a version of the reactor dynamics code DYN3D applicable for High Temperature Reactors; Entwicklung einer Version des Reaktordynamikcodes DYN3D fuer Hochtemperaturreaktoren. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Rohde, Ulrich; Apanasevich, Pavel; Baier, Silvio; Duerigen, Susan; Fridman, Emil; Grahn, Alexander; Kliem, Soeren; Merk, Bruno

    2012-07-15

    Based on the reactor dynamics code DYN3D for the simulation of transient processes in Light Water Reactors, a code version DYN3D-HTR for application to graphitemoderated, gas-cooled block-type high temperature reactors has been developed. This development comprises: - the methodical improvement of the 3D steady-state neutron flux calculation for the hexagonal geometry of the HTR fuel element blocks - the development of methods for the generation of homogenised cross section data taking into account the double heterogeneity of the fuel element block structure - the implementation of a 3D model for heat conduction and heat transport in the graphite matrix. The nodal method for neutron flux calculation based on SP3 transport approximation was extended to hexagonal fuel element geometry, where the hexagons are subdivided into triangles, thus the method had finally to be derived for triangular geometry. In triangular geometry, a subsequent subdivision of the hexagonal elements can be considered, and therefore, the effect of systematic mesh refinement can be studied. The algorithm was verified by comparison with Monte Carlo reference solutions, on the node-wise level, as well as also on the pin-wise level. New procedures were developed for the homogenization of the double-heterogeneous fuel element structures. One the one hand, the so-called Reactivity equivalent Physical Transformation (RPT), the two-step homogenization method based on 2D deterministic lattice calculations, was extended to cells with different temperatures of the materials. On the other hand, the progress in development of Monte Carlo methods for spectral calculations, in particular the development of the code SERPENT, opened a new, fully consistent 3D approach, where all details of the structures on fuel particle, fuel compact and fuel block level can be taken into account within one step. Moreover, a 3D heat conduction and heat transport model was integrated into DYN3D to be able to simulate radial

  14. A new MTR fuel for a new MTR reactor: UMo for the Jules Horowitz reactor

    Energy Technology Data Exchange (ETDEWEB)

    Guigon, B. [CEA Cadarache, Dir. de l' Energie Nucleaire DEN, Reacteur Jules Horowitz, 13 - Saint-Paul-lez-Durance (France); Vacelet, H. [Compagnie pour l' Etude et la Realisation de Combustibles Atomiques, CERCA, Etablissement de Romans, 26 (France); Dornbusch, D. [Technicatome, Service d' Architecture Generale, 13 - Aix-en-Provence (France)

    2003-07-01

    Within some years, the Jules Horowitz Reactor will be the only working experimental reactor (material and fuel testing reactor) in France. It will have to provide facilities for a wide range of needs: from activation analysis to power reactor fuel qualification. In this paper will be presented the main characteristics of the Jules Horowitz Reactor: its total power, neutron flux, fuel element... Safety criteria will be explained. Finally merits and disadvantages of UMo compared to the standard U{sub 3}Si{sub 2} fuel will be discussed. (authors)

  15. A packed bed membrane reactor for the oxidative dehydrogenation of propane on a Ga2O3 / MoO3 based catalyst

    NARCIS (Netherlands)

    Kotanjac, Ž.S.; Sint Annaland, van M.; Kuipers, J.A.M.

    2010-01-01

    Oxidative dehydrogenation of propane has been studied over a Ga2O3/MoO3 based catalyst. Using a differentially operated packed bed reactor with premixed oxygen and propane feed, the kinetic parameters for the main reaction and the consecutive and parallel reactions were experimentally determined. It

  16. Fusion blankets for catalyzed D--D and D--He3 reactors

    International Nuclear Information System (INIS)

    Fillo, J.A.; Powell, J.R.

    1977-01-01

    Blanket designs are presented for catalyzed D-D (Cat-D) and D-He 3 fusion reactors. Because of relatively low neutron wall loads and the flexibility due to non-tritium breeding, blankets potentially should operate for reactor life-times of approximately 30 years. Unscheduled replacement of failed blanket modules should be relatively rapid, due to very low residual activity, by operators working either through access ports in the shield (option 1) or directly in the plasma chamber (option 2). Cat-D blanket designs are presented for high (approximately 30%) and low (approximately 12%) β noncircular Tokamak reactors. The blankets are thick graphite screens, operating at high temperature to anneal radiation damage; the deposited neutron and gamma energy is thermally radiated along internal cavities and conducted to a bank of internal SiC coolant tubes (approximately 4 cm. ID) containing high pressure helium. In the D-He 3 Tokamak reactor design, the blanket consists of multiple layers (e.g., three) of thin (approximately 10 cm.) high strength aluminum (e.g., SAP), modular plates, cooled by organic terphynyl coolant

  17. Fusion blankets for catalyzed D--D and D--3He reactors

    International Nuclear Information System (INIS)

    Fillo, J.A.; Powell, J.R.

    1977-01-01

    Blanket designs are presented for catalyzed D-D (Cat-D) and D-He 3 fusion reactors. Because of relatively low neutron wall loads and the flexibility due to non-tritium breeding, blankets potentially should operate for reactor life-times of approximately 30 years. Unscheduled replacement of failed blanket modules should be relatively rapid, due to very low residual activity, by operators working either through access ports in the shield (option 1) or directly in the plasma chamber (option 2). Cat-D blanket designs are presented for high (approximately 30%) and low (approximately 12%) β non-circular Tokamak reactors. The blankets are thick graphite screens, operating at high temperature to anneal radiation damage; the deposited neutron and gamma energy is thermally radiated along internal cavities and conducted to a bank of internal SiC coolant tubes (approximately 4 cm. ID) containing high pressure helium. In the D-He 3 Tokamak reactor design, the blanket consists of multiple layers (e.g., three) of thin (approximately 10 cm.) high strength aluminum (e.g., SAP), modular plates, cooled by organic terphenyl coolant

  18. Digital computer control on Canadian nuclear power plants -experience to date and the future outlook

    International Nuclear Information System (INIS)

    Pearson, A.

    1977-10-01

    This paper discusses the performance of the digital computer control system at Pickering through the years 1973 to 1976. This evaluation is based on a study of the Pickering Generating Station operating records. The paper goes on to explore future computer architectures and the advantages that could accrue from a distributed system approach. Also outlined are the steps being taken to develop these ideas further in the context of two Chalk River projects - REDNET, an advanced data acquisition system being installed to process information from engineering experiments in NRX and NRU reactors, and CRIP, a prototype communications network using cable television technology. (author)

  19. Irradiated graphite studies prior to decommissioning of G1, G2 and G3 reactors

    International Nuclear Information System (INIS)

    Bonal, J.P.; Vistoli, J.Ph.; Combes, C.

    2005-01-01

    G1 (46 MW th ), G2 (250 MW th ) and G3 (250 MW th ) are the first French plutonium production reactors owned by CEA (Commissariat a l'Energie Atomique). They started to be operated in 1956 (G1), 1959 (G2) and 1960 (G3); their final shutdown occurred in 1968, 1980 and 1984 respectively. Each reactor used about 1200 tons of graphite as moderator, moreover in G2 and G3, a 95 tons graphite wall is used to shield the rear side concrete from neutron irradiation. G1 is an air cooled reactor operated at a graphite temperature ranging from 30 C to 230 C; G2 and G3 are CO 2 cooled reactors and during operation the graphite temperature is higher (140 C to 400 C). These reactors are now partly decommissioned, but the graphite stacks are still inside the reactors. The graphite core radioactivity has decreased enough so that a full decommissioning stage may be considered. Conceming this decommissioning, the studies reported here are: (i) stored energy in graphite, (ii) graphite radioactivity measurements, (iii) leaching of radionuclide ( 14 C, 36 Cl, 63 Ni, 60 Co, 3 H) from graphite, (iv) chlorine diffusion through graphite. (authors)

  20. RA reactor exploitation, task 3.08/01

    International Nuclear Information System (INIS)

    Zecevic, V.

    1963-01-01

    During 1963 the RA reactor was operated for 1852 hours at mean power of 5.7 MW (total power production was 10716 MWh). Reactor was used for irradiation according to the demand of 356 users, and 15 experiments. The reason for decreased operation in comparison with the previous year was repair of all the reactor equipment and decontamination of the heavy water system. This report contains detailed data about reactor power, reactivity changes and fuel burnup. Mean monthly usage of the reactor experimental channels as well as samples which were irradiated are part of this report

  1. Denitrification performance of Pseudomonas denitrificans in a fluidized-bed biofilm reactor and in a stirred tank reactor

    Energy Technology Data Exchange (ETDEWEB)

    Cattaneo, C.; Nicolella, C.; Rovatti, M. [Department of Chemical and Process Engineering, Faculty of Engineering, University of Genoa, Via Opera Pia 15, 16145 Genoa (Italy)

    2003-04-09

    Denitrification of a synthetic wastewater containing nitrates and methanol as carbon source was carried out in two systems - a fluidized-bed biofilm reactor (FBBR) and a stirred tank reactor (STR) - using Pseudomonas denitrificans over a period of five months. Nitrogen loading was varied during operation of both reactors to assess differences in the response to transient conditions. Experimental data were analyzed to obtain a comparison of denitrification kinetics in biofilm and suspended growth reactors. The comparison showed that the volumetric degradation capacity in the FBBR (5.36 kg {sub N} . m{sup -3} . d{sup -1}) was higher than in the STR, due to higher biomass concentration (10 kg {sub BM} . m{sup -3} vs 1.2 kg {sub BM} m{sup -3}). (Abstract Copyright [2003], Wiley Periodicals, Inc.)

  2. Design of a New Research Reactor: Preliminary Conceptual Design (3rd Year)

    International Nuclear Information System (INIS)

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

    2006-01-01

    A research reactor design is a kind of integral engineering project and a process to obtain a concrete shape through several years of concept development, conceptual design, basic design and detail design. So it requires close cooperation in various areas as well as lots of manpower and cost. The overall process at each stage may be said to be similar except for some stage-specific works. In 2005 as last year of a concept development stage, investigations on the various concepts of the fuel, reactor structure and systems which can meet the requirements established. The requirements for the process systems and I and C systems have also been embodied. The major tasks planned at the early of 2005 have been performed for each area of reactor design as follows: Establishment of the fuel and reactor core concept, and the core analysis, Preliminary thermal-hydraulic and safety analyses for the conceptual cores, Establishment and improvement of analysis system, Concept developments of the reactor structures and major systems, Test and test plan to verify the developed concepts, International cooperation to establish the foundations for exporting a research reactor

  3. 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

  4. Simulation in 3 dimensions of a cycle 18 months for an BWR type reactor using the Nod3D program; Simulacion en 3 dimensiones de un ciclo de 18 meses para un reactor BWR usando el programa Nod3D

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez, N.; Alonso, G. [ININ, A.P. 18-1027, 11801 Mexico D.F. (Mexico)]. E-mail: nhm@nuclear.inin.mx; Valle, E. del [IPN, ESFM, 07738 Mexico D.F. (Mexico)

    2004-07-01

    The development of own codes that you/they allow the simulation in 3 dimensions of the nucleus of a reactor and be of easy maintenance, without the consequent payment of expensive use licenses, it can be a factor that propitiates the technological independence. In the Department of Nuclear Engineering (DIN) of the Superior School of Physics and Mathematics (ESFM) of the National Polytechnic Institute (IPN) a denominated program Nod3D has been developed with the one that one can simulate the operation of a reactor BWR in 3 dimensions calculating the effective multiplication factor (kJJ3, as well as the distribution of the flow neutronic and of the axial and radial profiles of the power, inside a means of well-known characteristics solving the equations of diffusion of neutrons numerically in stationary state and geometry XYZ using the mathematical nodal method RTN0 (Raviart-Thomas-Nedelec of index zero). One of the limitations of the program Nod3D is that it doesn't allow to consider the burnt of the fuel in an independent way considering feedback, this makes it in an implicit way considering the effective sections in each step of burnt and these sections are obtained of the code Core Master LEND. However even given this limitation, the results obtained in the simulation of a cycle of typical operation of a reactor of the type BWR are similar to those reported by the code Core Master LENDS. The results of the keJ - that were obtained with the program Nod3D they were compared with the results of the code Core Master LEND, presenting a difference smaller than 0.2% (200 pcm), and in the case of the axial profile of power, the maxim differs it was of 2.5%. (Author)

  5. Mortality among long-term Chalk River employees

    International Nuclear Information System (INIS)

    Werner, M.M.; Myers, D.K.

    1986-12-01

    Mortality among Chalk River Nuclear Laboratory (CRNL) employees who died during employment or after retirement has been updated to 1985 December 31. Data in tabular form are presented for overall mortality for male and female employees, for the participants in the clean-up for the NRX and NRU reactor accidents and for a group of CRNL staff with lifetime accumulative doses in excess of 0.2 Sv. Data are also presented on the different types of cancer causing death among male employees. No statistically significant increases in cancer deaths were found in any of the groups analyzed. 25 refs

  6. The first critical experiment with a new type of fuel assemblies IRT-3M on the training reactor VR-I

    International Nuclear Information System (INIS)

    Matejka, Karel; Sklenka, Lubomir

    1997-01-01

    The paper 'The first critical experiment with a new type of fuel assemblies IRT-3M on training reactor VR-1 presents basic information about the replacement of fuel on the reactor VR-1 run on FJFI CVUT in Prague. In spring 1997 the IRT-2M fuel type used till then was replaced by the IRT-3M type. When the fuel was replaced, no change in its enrichment was made, i.e. its level remained as 36% 235 U. The replacement itself was carried out in tight co-operation with the Nuclear Research Institute Rez plc., as related to the operation of the research reactor LVR-15. The fuel replacement on the VR-I reactor is a part of the international program RERTR (Reduced Enrichment for Research and Test Reactors) in which the Czech Republic participates. (author)

  7. Reversed field pinch reactor study 3

    International Nuclear Information System (INIS)

    Hollis, A.A.; Mitchell, J.T.D.

    1977-12-01

    This report, the third of a series on the Reversed Field Pinch Reactor, describes a preliminary concept of the engineering design and layout of this pulsed toroidal reactor, which uses the stable plasma behaviour first observed in ZETA. The basic parameters of the 600 MW(e) reactor are taken from a companion study by Hancox and Spears. The plasma volume is 1.75m minor radius and 16m major radius surrounded by a 1.8m blanket-shield region - with the blanket divided into 14 removable segments for servicing. The magnetic confinement system consists of 28 toroidal field coils situated just outside the blanket and inside the poloidal and vertical field coils and all coils have normal copper conductors. The requirement to incorporate a conducting shell at the front of the blanket to provide a short-time plasma stability has a marked effect on the design. It sets the size of the blanket segment and the scale of the servicing operations, limits the breeding gain and complicates the blanket cooling and its integration with the heat engine. An extensive study will be required to confirm the overall reactor potential of the concept. (author)

  8. Safety in the ARIES-III D-3He tokamak reactor design

    International Nuclear Information System (INIS)

    Herring, J.S.; Dolan, T.J.

    1992-01-01

    This paper reports on the ARIES-III reactor study, an extensive examination of the viability of a D- 3 He-fueled commercial tokamak powder reactor. Because neutrons are produced only through side reactions (D+D- 3 HE+N; and D+D-T+p followed by D+T- 4 He+n), the reactor has the significant advantages of reduced activation of the first wall and shield, low afterheat and Class A or C low level waste disposal. Since no tritium is required for operation, no lithium-containing breeding blanket is necessary. A ferritic steel shield behind the first wall protects the magnets from gamma and neutron heating and from radiation damage. The authors explored the potential for isotopically tailoring the 4 mm tungsten layer on the divertor in order to reduce the offsite doses should a tungsten aerosol be released from the reactor after an accident. The authors also modeled a loss-of-cooling accident (LOCA) in which the organic coolant was burning in order to estimate the amount of radionuclides released from the first wall. Because the maximum temperature is low, degree C, release fractions are small. The authors analyzed the disposition of the 20 g/day of tritium that is produced by D-D reactions and removed by the vacuum pumps

  9. Reactor core for LMFBR type reactors

    International Nuclear Information System (INIS)

    Masumi, Ryoji; Azekura, Kazuo; Kurihara, Kunitoshi; Bando, Masaru; Watari, Yoshio.

    1987-01-01

    Purpose: To reduce the power distribution fluctuations and obtain flat and stable power distribution throughout the operation period in an LMFBR type reactor. Constitution: In the inner reactor core region and the outer reactor core region surrounding the same, the thickness of the inner region is made smaller than the axial height of the reactor core region and the radial width thereof is made smaller than that of the reactor core region and the volume thereof is made to 30 - 50 % for the reactor core region. Further, the amount of the fuel material per unit volume in the inner region is made to 70 - 90 % of that in the outer region. The difference in the neutron infinite multiplication factor between the inner region and the outer region is substantially constant irrespective of the burnup degree and the power distribution fluctuation can be reduced to about 2/3, by which the effect of thermal striping to the reactor core upper mechanisms can be moderated. Further, the maximum linear power during operation can be reduced by 3 %, by which the thermal margin in the reactor core is increased and the reactor core fuels can be saved by 3 %. (Kamimura, M.)

  10. The Flamanville 3 EPR reactor; Le reacteur EPR Flamanville 3

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2007-07-01

    On April 10. 2007, the french government authorized EDF to create on the site of Flamanville ( La Manche) a nuclear base installation containing a pressurized water EPR type reactor. This nuclear reactor, conceived by AREVA NP and EDF, is the first copy of a generation susceptible to replace later, at least partly, the French nuclear reactors at present in operation.Within the framework of its mission of technical support of the Authority of Nuclear Safety ( A.S.N.), the I.R.S.N. widely contributed successively: to define the general objectives of safety assigned to this new generation of pressurized water nuclear reactors; to analyze the options of safety proposed by EDF for the EPR project; To deepen, upstream to the authorization of creation, the evaluation of the step of safety and the measures of conception retained by EDF that have to allow to respect the objectives of safety which were notified to it. (N.C.)

  11. Applicability of RELAP5-3D for Thermal-Hydraulic Analyses of a Sodium-Cooled Actinide Burner Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    C. B. Davis

    2006-07-01

    The Actinide Burner Test Reactor (ABTR) is envisioned as a sodium-cooled, fast reactor that will burn the actinides generated in light water reactors to reduce nuclear waste and ease proliferation concerns. The RELAP5-3D computer code is being considered as the thermal-hydraulic system code to support the development of the ABTR. An evaluation was performed to determine the applicability of RELAP5-3D for the analysis of a sodium-cooled fast reactor. The applicability evaluation consisted of several steps, including identifying the important transients and phenomena expected in the ABTR, identifying the models and correlations that affect the code’s calculation of the important phenomena, and evaluating the applicability of the important models and correlations for calculating the important phenomena expected in the ABTR. The applicability evaluation identified code improvements and additional models needed to simulate the ABTR. The accuracy of the calculated thermodynamic and transport properties for sodium was also evaluated.

  12. Simulation in 3 dimensions of a cycle 18 months for an BWR type reactor using the Nod3D program

    International Nuclear Information System (INIS)

    Hernandez, N.; Alonso, G.; Valle, E. del

    2004-01-01

    The development of own codes that you/they allow the simulation in 3 dimensions of the nucleus of a reactor and be of easy maintenance, without the consequent payment of expensive use licenses, it can be a factor that propitiates the technological independence. In the Department of Nuclear Engineering (DIN) of the Superior School of Physics and Mathematics (ESFM) of the National Polytechnic Institute (IPN) a denominated program Nod3D has been developed with the one that one can simulate the operation of a reactor BWR in 3 dimensions calculating the effective multiplication factor (kJJ3, as well as the distribution of the flow neutronic and of the axial and radial profiles of the power, inside a means of well-known characteristics solving the equations of diffusion of neutrons numerically in stationary state and geometry XYZ using the mathematical nodal method RTN0 (Raviart-Thomas-Nedelec of index zero). One of the limitations of the program Nod3D is that it doesn't allow to consider the burnt of the fuel in an independent way considering feedback, this makes it in an implicit way considering the effective sections in each step of burnt and these sections are obtained of the code Core Master LEND. However even given this limitation, the results obtained in the simulation of a cycle of typical operation of a reactor of the type BWR are similar to those reported by the code Core Master LENDS. The results of the keJ - that were obtained with the program Nod3D they were compared with the results of the code Core Master LEND, presenting a difference smaller than 0.2% (200 pcm), and in the case of the axial profile of power, the maxim differs it was of 2.5%. (Author)

  13. Development of telerobotic systems for reactor decommissioning, (3)

    International Nuclear Information System (INIS)

    Usui, Hozumi; Fujii, Yoshio; Shinohara, Yoshikuni

    1991-01-01

    This paper describes the telerobotic system for reactor decommissioning in the scope of engineering demonstration of dismantling radioactive reactor internals of an experimental boiling water power reactor JPDR. The total system consists of a telerobotic manipulator system equipped with a multi-functional amphibious slave manipulator with a load capacity of 25 daN, a chain-driven transport system, and a computer-assisted monitoring and control system. Preceding to the application of the telerobotic system to actual dismantling operation, a mockup test was performed of dismantling the simulated reactor internals of actual-size by the method of underwater plasma arc cutting in order to study the performance of the telerobotic system in a realistic environment. The system was then successfully applied to dismantling the actual reactor internals according to the JPDR decommissioning program. (author)

  14. The World's Reactors no. 70 - Forsmark 3, BWR-75

    International Nuclear Information System (INIS)

    Anon.

    1976-01-01

    A large pull-out wall chart is presented showing a coloured cut-away diagram of the Forsmark 3 station. It is accompanied by 2 small sketches one showing the layout of station buildings and the other the inside of the reactor vessel. Parameters are listed. (U.K.)

  15. A solution to level 3 dismantling of gas-cooled reactors: Graphite incineration

    International Nuclear Information System (INIS)

    Dubourg, M.

    1993-01-01

    This paper presents an approach developed to solve the specific decommissioning problems of the G2 and G3 gas cooled reactors at Marcoule and the strategy applied with emphasis in incinerating the graphite core components, using a fluidized-bed incinerator developed jointly between the CEA and FRAMATOME. The incineration option was selected over subsurface storage for technical and economic reasons. Studies have shown that gaseous incineration releases are environmentally acceptable

  16. Photocatalytic reactors for treating water pollution with solar illumination, Part 3: a simplified analysis for recirculating reactors

    Energy Technology Data Exchange (ETDEWEB)

    Sagawe, G.; Bahnemann, D. [Hannover Univ. (Germany). Inst. fuer Technische Chemie; Brandi, R.J.; Cassano, A.E. [Universidad Nacional de Litoral, Santa Fe (Argentina). Inst. de Desarrollo Tecnologico para la Imdustria Quimica

    2004-11-01

    A solar photoreactor operated in the batch, recirculating mode is analyzed in terms of very simple observable variables such as the impinging photon flux, the incident area, the initial concentration, the flow rate, the reactor volume and a property defined as the Observed Photonic Efficiency. The proposed equipment is made of a tubular reactor, a tank, a pump and the connecting pipes. The analysis is formulated in terms of the photon input corresponding to an equivalent batch system that is derived as a new reaction coordinate for photoreactions. Employing several plausible approximations, the pollutant concentration evolution in the tank is cast in terms of very simple analytical solutions. Process photonic efficiencies are defined for the system operation and calculated with respect to the maximum achievable yield corresponding to the differential operation of the solar recirculating reactor. (Author)

  17. RA Reactor applications, Annex A

    International Nuclear Information System (INIS)

    Martinc, R.; Cupac, S.; Stanic, A.

    1990-01-01

    RA reactor was not operated during the past five years due to the renewal and reconstruction of the reactor systems, which in underway. In the period from 1986-1990, reactor was operated only 144 MWh in 1986, for the need of testing the reactor systems and possibility of irradiating 125 I. Reactor will not be operated in 1991 because of the exchange of complete instrumentation which is planned to be finished by the end of 1991. It is expected to start operation in May 1992. That is why this annex includes the plan of reactor operation for period of nine months starting from from the moment of start-up. It is planned to operate the reactor at 0.02 MW power first three months, to increase the power gradually and reach 3.5 MW after 8 months of operation. It is foreseen to operate the reactor at 4.7 MW from the tenth month on [sr

  18. Modular 3D printed lab-on-a-chip bio-reactor for the biochemical energy cascade of microorganisms

    Science.gov (United States)

    Podwin, Agnieszka; Dziuban, Jan A.

    2017-10-01

    The paper presents the sandwiched polymer 3D printed lab-on-a-chip bio-reactor for the biochemical energy cascade of microorganisms. Euglenas and yeast were separately and simultaneously cultured for 10 d in the chip. As a result of the experiments, euglenas, light-initialized and nourished by CO2—a product of ethanol fermentation handled by yeast—generated oxygen, based on the photosynthesis process. The presence of oxygen in the bio-reactor was confirmed by the colorimetric method—a bicarbonate (pH) indicator. Preliminary studies towards the obtainment of an effective source of oxygen are promising and further research should be done to enable the utility of the bio-reactor in, for instance, microbial fuel cells.

  19. Modular 3D printed lab-on-a-chip bio-reactor for the biochemical energy cascade of microorganisms

    International Nuclear Information System (INIS)

    Podwin, Agnieszka; Dziuban, Jan A

    2017-01-01

    The paper presents the sandwiched polymer 3D printed lab-on-a-chip bio-reactor for the biochemical energy cascade of microorganisms. Euglenas and yeast were separately and simultaneously cultured for 10 d in the chip. As a result of the experiments, euglenas, light-initialized and nourished by CO 2 —a product of ethanol fermentation handled by yeast—generated oxygen, based on the photosynthesis process. The presence of oxygen in the bio-reactor was confirmed by the colorimetric method—a bicarbonate (pH) indicator. Preliminary studies towards the obtainment of an effective source of oxygen are promising and further research should be done to enable the utility of the bio-reactor in, for instance, microbial fuel cells. (paper)

  20. Operation experience with the 3 MW TRIGA Mark-II research reactor of Bangladesh

    International Nuclear Information System (INIS)

    Islam, M.S.; Haque, M.M.; Salam, M.A.; Rahman, M.M.; Khandokar, M.R.I.; Sardar, M.A.; Saha, P.K.; Haque, A.; Malek Sonar, M.A.; Uddin, M.M.; Hossain, S.M.S.; Zulquarnain, M.A.

    2004-01-01

    The 3 MW TRIGA Mark-II research reactor of Bangladesh Atomic Energy Commission (BAEC) has been operating since September 14, 1986. The reactor is used for radioisotope production ( 131 I, 99m Tc, 46 Sc), various R and D activities and manpower training. The reactor has been operated successfully since it's commissioning with the exception of a few reportable incidents. Of these, the decay tank leakage incident of 1997 is considered to be the most significant one. As a result of this incident, reactor operation at full power under forced-convection mode remained suspended for about 4 years. During that time, the reactor was operated at a power level of 250 kW so as to carry out experiments that require lower neutron flux. This was made possible by establishing a temporary by pass connection across the decay tank using local technology. The other incident was the contamination of the Dry Central Thimble (DCT) that took place in March 2002 when a pyrex vial containing 50 g of TeO 2 powder got melted inside the DCT. The vial was melted due to high heat generation on its surface while the reactor was operated for 8 hours at 3 MW for trial production of Iodine-131 ( 131 I). A Wet Central Thimble (WCT) was used to replace the damaged DCT in June 2002 such that the reactor operation could be resumed. The WCT was again replaced by a new DCT in June 2003 such that radioisotope production could be continued. A total of 873 irradiation requests (IRs) have been catered for different reactor uses. Out of these, 114 IRs were for radioisotope (RI) production and 759 IRs for different experiments. The total amount of RI produced stands at about 2100 GBq. The total amount of burn-up-fuel is about 6158 MWh. Efforts are on to undertake an ADP project so as to convert the analog console and I and C system of the reactor into digital one. The paper summarizes the reactor operation experiences focusing on troubleshooting, rectification, modification, RI production, various R and D

  1. Distributed control and instrumentation systems for future nuclear power plants

    International Nuclear Information System (INIS)

    Yan, G.; L'Archeveque, J.V.R.

    1976-01-01

    The centralized dual computer system philosophy has evolved as the key concept underlying the highly successful application of direct digital control in CANDU power reactors. After more than a decade, this basis philosophy bears re-examination in the light of advances in system concepts--notably distributed architectures. A number of related experimental programs, all aimed at exploring the prospects of applying distributed systems in Canadian nuclear power plants are discussed. It was realized from the outset that the successful application of distributed systems depends on the availability of a highly reliable, high capacity, low cost communications medium. Accordingly, an experimental facility has been established and experiments have been defined to address such problem areas as interprocess communications, distributed data base design and man/machine interfaces. The design of a first application to be installed at the NRU/NRX research reactors is progressing well

  2. EL3 reactor description and safety analysis report; Pile EL3, rapport descriptif et de surete

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1969-02-01

    The EL-3 reactor is an experimental pile. Heterogenous type reactor, water moderated and cooled it uses slightly enriched uranium oxide as fuel (4.5 percent) distributed in vertical cells that constitute the core (the maximum number of cells is 99). It is conceived to function at a maximal thermal power of 20 MW. It supplies a maximum thermal neutron flux of 10{sup 14} neutrons/cm{sup 2}/sec. It has several experimental devices. The EL-3 reactor is surrounded by auxiliary circuits of fluids, in a sealed containment, slightly depressed. The primary heavy water coolant circuit is completely included in this containment. Its cooling is made by the intermediary of a light water secondary circuit by atmospheric refrigerants. The ventilation circuits of the sealed containment and the reactor block do not release air outside, under nornal functioning, by a particularly studied chimney only after filtering and eventually dilution. The eventual contamination of the light water or air by active products is permanently monitored to allow the reactor shutdown and avoid the release in atmosphere of dangerous products. The EL-3 reactor, laying down in may 1955, has diverged in july 1957, made its first ascending in power in december 1957 and reached its complete power in april 1958. The positioning of actual fuel (snow crystal) was made during summer 1964. Reactor with an experimental aim, it is used for theoretical and technological studies by material irradiation in the experimental channels and the core cells, with possibilities to constitute independent loops (relative to the cooling fluids). Thirty vertical channels are devoted to the fabrication of artificial radioelements. [French] La pile EL-3 est une pile experimentale. Du type heterogene, moderee et refroidie a l'eau lourde elle utilise comme combustible de l'oxygene d'uranium faiblement enrichi (4,5 p.cent) reparti en cellules verticales qui constituent le coeur (le nombre maximal de cellules est de, 99). Elle est

  3. EL3 reactor description and safety analysis report; Pile EL3, rapport descriptif et de surete

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1969-02-01

    The EL-3 reactor is an experimental pile. Heterogenous type reactor, water moderated and cooled it uses slightly enriched uranium oxide as fuel (4.5 percent) distributed in vertical cells that constitute the core (the maximum number of cells is 99). It is conceived to function at a maximal thermal power of 20 MW. It supplies a maximum thermal neutron flux of 10{sup 14} neutrons/cm{sup 2}/sec. It has several experimental devices. The EL-3 reactor is surrounded by auxiliary circuits of fluids, in a sealed containment, slightly depressed. The primary heavy water coolant circuit is completely included in this containment. Its cooling is made by the intermediary of a light water secondary circuit by atmospheric refrigerants. The ventilation circuits of the sealed containment and the reactor block do not release air outside, under nornal functioning, by a particularly studied chimney only after filtering and eventually dilution. The eventual contamination of the light water or air by active products is permanently monitored to allow the reactor shutdown and avoid the release in atmosphere of dangerous products. The EL-3 reactor, laying down in may 1955, has diverged in july 1957, made its first ascending in power in december 1957 and reached its complete power in april 1958. The positioning of actual fuel (snow crystal) was made during summer 1964. Reactor with an experimental aim, it is used for theoretical and technological studies by material irradiation in the experimental channels and the core cells, with possibilities to constitute independent loops (relative to the cooling fluids). Thirty vertical channels are devoted to the fabrication of artificial radioelements. [French] La pile EL-3 est une pile experimentale. Du type heterogene, moderee et refroidie a l'eau lourde elle utilise comme combustible de l'oxygene d'uranium faiblement enrichi (4,5 p.cent) reparti en cellules verticales qui constituent le coeur (le nombre maximal de cellules est de, 99

  4. Gas-cooled reactor thermal-hydraulics using CAST3M and CRONOS2 codes

    International Nuclear Information System (INIS)

    Studer, E.; Coulon, N.; Stietel, A.; Damian, F.; Golfier, H.; Raepsaet, X.

    2003-01-01

    The CEA R and D program on advanced Gas Cooled Reactors (GCR) relies on different concepts: modular High Temperature Reactor (HTR), its evolution dedicated to hydrogen production (Very High Temperature Reactor) and Gas Cooled Fast Reactors (GCFR). Some key safety questions are related to decay heat removal during potential accident. This is strongly connected to passive natural convection (including gas injection of Helium, CO 2 , Nitrogen or Argon) or forced convection using active safety systems (gas blowers, heat exchangers). To support this effort, thermal-hydraulics computer codes will be necessary tools to design, enhance the performance and ensure a high safety level of the different reactors. Accurate and efficient modeling of heat transfer by conduction, convection or thermal radiation as well as energy storage are necessary requirements to obtain a high level of confidence in the thermal-hydraulic simulations. To achieve that goal a thorough validation process has to ve conducted. CEA's CAST3M code dedicated to GCR thermal-hydraulics has been validated against different test cases: academic interaction between natural convection and thermal radiation, small scale in-house THERCE experiments and large scale High Temperature Test Reactor benchmarks such as HTTR-VC benchmark. Coupling with neutronics is also an important modeling aspect for the determination of neutronic parameters such as neutronic coefficient (Doppler, moderator,...), critical position of control rods...CEA's CAST3M and CRONOS2 computer codes allow this coupling and a first example of coupled thermal-hydraulics/neutronics calculations has been performed. Comparison with experimental data will be the next step with High Temperature Test Reactor experimental results at nominal power

  5. 3-DB, 3-D Multigroup Diffusion, X-Y-Z, R-Theta-Z, Triangular-Z Geometry, Fast Reactor Burnup

    International Nuclear Information System (INIS)

    Hardie, R.W.; Little, W.W. Jr.; Mroz, W.

    1974-01-01

    1 - Description of problem or function: 3DB is a three-dimensional (x-y-z, r-theta-z, triangular-z) multigroup diffusion code for use in detailed fast-reactor criticality and burnup analysis. The code can be used to - (a) compute k eff and perform criticality searches on time absorption, reactor composition, and reactor dimensions by means of either a flux or an adjoint model, (b) compute material burnup using a flexible material shuffling scheme, and (c) compute flux distributions for an arbitrary extraneous source. 2 - Method of solution: Eigenvalues are computed by standard source- iteration techniques. Group re-balancing and successive over-relaxation with line inversion are used to accelerate convergence. Adjoint solutions are obtained by inverting the input data and redefining the source terms. Material burnup is by reactor zone. The burnup rate is determined by the zone and energy-averaged cross sections which are recomputed after each time-step. The isotopic chains, which can contain any number of isotopes are formed by the user. The code does not contain built- in or internal chains. 3 - Restrictions on the complexity of the problem: Since variable dimensioning is employed, no simple bounds can be stated

  6. Energy conversion options for ARIES-III - A conceptual D-3He tokamak reactor

    International Nuclear Information System (INIS)

    Santarius, J.F.; Blanchard, J.P.; Emmert, G.A.; Sviatoslavsky, I.N.; Wittenberg, L.J.; Ghoneim, N.M.; Hasan, M.Z.; Mau, T.K.; Greenspan, E.; Herring, J.S.; Kernbichler, W.; Klein, A.C.; Miley, G.H.; Miller, R.L.; Peng, Y.K.M.

    1989-01-01

    The potential for highly efficient conversion of fusion power to electricity provides one motivation for investigating D- 3 He fusion reactors. This stems from: (1) the large fraction of D- 3 He power produced in the forms of charged particles and synchrotron radiation which are amenable to direct conversion, and (2) the low neutron fluence and lack of tritium breeding constraints, which increase design flexibility. The design team for a conceptual D- 3 He tokamak reactor, ARIES-III, has investigated numerous energy conversion options at a scoping level in attempting to realize high efficiency. The energy conversion systems have been studied in the context of their use on one or more of three versions of a D- 3 He tokamak: a first stability regime device, a second stability regime device, and a spherical torus. The set of energy conversion options investigated includes bootstrap current conversion, compression-expansion cycles, direct electrodynamic conversion, electrostatic direct conversion, internal electric generator, liquid metal heat engine blanket, liquid metal MHD, plasma MHD, radiation boiler, scrape-off layer thermoelectric, synchrotron radiation conversion by rectennas, synchrotron radiation conversion by thermal cycles, thermionic/AMTEC/thermal systems, and traveling wave conversion. The original set of options is briefly discussed, and those selected for further study are described in more detail. The four selected are liquid metal MHD, plasma MHD, rectenna conversion, and direct electrodynamic conversion. Thermionic energy conversion is being considered, and some options may require a thermal cycle in parallel or series. 17 refs., 3 figs., 1 tab

  7. Decommissioning of the BR3 pressurized-water reactor

    International Nuclear Information System (INIS)

    Massaut, V.

    1996-01-01

    The dismantling and the decommissioning of nuclear installations at the end of their life-cycle is a new challenge to the nuclear industry. Different techniques and procedures for the dismantling of a nuclear power plant on an existing installation, the BR-3 pressurized-water reactor, are described. The scientific programme, objectives, achievements in this research area at the Belgian Nuclear Research Centre SCK-CEN for 1995 are summarized

  8. The 4th surveillance testing for Kori unit 3 reactor vessel materials

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-10-01

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 4th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejeon after the capsule was transported from Kori site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Kori unit 3 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules U, V, X and W are 4.983E+18, 1.641E+19, 3.158E+19, and 4.469E+19n/cm{sup 2}, respectively. The bias factor, the ratio of calculation/measurement, was 0.840 for the 1st through 4th testing and the calculational uncertainty, 12% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.362E+19n/cm{sup 2} based on the end of 12th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 40, 48 and 56EFPY would reach 3.481E+19, 4.209E+19, 5.144E+19 and 5.974E+19n/cm{sup 2} based on the current calculation. The result through this analysis for Kori unit 3 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life. 48 refs., 35 figs., 41 tabs. (Author)

  9. Reactor core of FBR type reactor

    International Nuclear Information System (INIS)

    Hayashi, Hideyuki; Ichimiya, Masakazu.

    1994-01-01

    A reactor core is a homogeneous reactor core divided into two regions of an inner reactor core region at the center and an outer reactor core region surrounding the outside of the inner reactor core region. In this case, the inner reactor core region has a lower plutonium enrichment degree and less amount of neutron leakage in the radial direction, and the outer reactor core region has higher plutonium enrichment degree and greater amount of neutron leakage in the radial direction. Moderator materials containing hydrogen are added only to the inner reactor core fuels in the inner reactor core region. Pins loaded with the fuels with addition of the moderator materials are inserted at a ratio of from 3 to 10% of the total number of the fuel pins. The moderator materials containing hydrogen comprise zirconium hydride, titanium hydride, or calcium hydride. With such a constitution, fluctuation of the power distribution in the radial direction along with burning is suppressed. In addition, an absolute value of the Doppler coefficient can be increased, and a temperature coefficient of coolants can be reduced. (I.N.)

  10. Comparison of 2D and 3D Neutron Transport Analyses on Yonggwang Unit 3 Reactor

    International Nuclear Information System (INIS)

    Maeng, Aoung Jae; Kim, Byoung Chul; Lim, Mi Joung; Kim, Kyung Sik; Jeon, Young Kyou; Yoo, Choon Sung

    2012-01-01

    10 CFR Part 50 Appendix H requires periodical surveillance program in the reactor vessel (RV) belt line region of light water nuclear power plant to check vessel integrity resulting from the exposure to neutron irradiation and thermal environment. Exact exposure analysis of the neutron fluence based on right modeling and simulations is the most important in the evaluation. Traditional 2 dimensional (D) and 1D synthesis methodologies have been widely applied to evaluate the fast neutron (E > 1.0 MeV) fluence exposure to RV. However, 2D and 1D methodologies have not provided accurate fast neutron fluence evaluation at elevations far above or below the active core region. RAPTOR-M3G (RApid Parallel Transport Of Radiation - Multiple 3D Geometries) program for 3D geometries calculation was therefore developed both by Westinghouse Electronic Company, USA and Korea Reactor Integrity Surveillance Technology (KRIST) for the analysis of In-Vessel Surveillance Test and Ex-Vessel Neutron Dosimetry (EVND). Especially EVND which is installed at active core height between biological shielding material and concrete also evaluates axial neutron fluence by placing three dosimetries each at Top, Middle and Bottom part of the angle representing maximum neutron fluence. The EVND programs have been applied to the Korea Nuclear Plants. The objective of this study is therefore to compare the 3D and the 2D Neutron Transport Calculations and Analyses on the Yonggwang unit 3 Reactor as an example

  11. VIPRE-01: a thermal-hydraulic analysis code for reactor cores. Volume 3. Programmer's manual. Final report

    International Nuclear Information System (INIS)

    Stewart, C.W.; Koontz, A.S.; Cuta, J.M.; Montgomery, S.D.

    1983-05-01

    VIPRE (Versatile Internals and Component Program for Reactors; EPRI) has been developed for nuclear power utility thermal-hydraulic analysis applications. It is designed to help evaluate nuclear-reactor-core safety limits including minimum departure from nucleate boiling ratio (MDNBR), critical power ratio (CPR), fuel and clad temperatures, and coolant state in normal operation and assumed accident conditions. This is Volume 3, the Programmer's Manual. It explains the codes' structures and the computer interfaces

  12. A comparison of the electrochemical recovery of palladium using a parallel flat plate flow-by reactor and a rotating cylinder electrode reactor

    International Nuclear Information System (INIS)

    Terrazas-Rodriguez, J.E.; Gutierrez-Granados, S.; Alatorre-Ordaz, M.A.; Ponce de Leon, C.; Walsh, F.C.

    2011-01-01

    The production of catalytic converters generates large amounts of waste water containing Pd 2+ , Rh 3+ and Nd 3+ ions. The electrochemical treatment of these solutions offers an economic and effective alternative to recover the precious metals in comparison with other traditional metal recovery technologies. The separation of palladium from this mixture of metal ions by catalytic deposition was carried out using a rotating cylinder electrode reactor (RCER) and a parallel plate reactor (FM01-LC) with the same cathode area (64 cm 2 ) and electrolyte volume (300 cm 3 ). The study was carried out at mean linear flow velocities of 1.27 -1 (120 e /v -1 (7390 2+ ions in the parallel plate electrode reactor was 35% while the recovery of 97% of Pd 2+ in the RCER was 62%. The volumetric energy consumption during the electrolysis was 0.56 kW h m -3 and 2.1 kW h m -3 for the RCER and the FM01-LC reactors, respectively. Using a three-dimensional stainless steel electrode in the FM01-LC laboratory reactor, 99% of palladium ions were recovered after 30 min of electrolysis while in the RCER, 120 min were necessary.

  13. Computational Analysis of Nuclear Safety Parameters of 3 MW TRIGA Mark-II Research Reactor Based on Evaluated Nuclear Data Libraries JENDL-3.3 and ENDF/B-VII.0

    International Nuclear Information System (INIS)

    Khan, Jahirul Haque

    2013-01-01

    The objective of this study is to explain the main nuclear safety parameters of 3 MW TRIGA Mark-II Research Reactor at AERE, Savar, Dhaka, Bangladesh from the viewpoint of reactor safety and also reactor operator. The most important nuclear reactor physics safety parameters are power distribution, power peaking factors, shutdown margin, control rod worth, excess reactivity and fuel temperature reactivity coefficient. These parameters are calculated using the chain of the computer codes the SRAC-PIJ for cell calculation based on neutron transport theory and the SRAC-CITATION for core calculation based on neutron diffusion equation. To achieve this objective the TRIGA model is developed by the 3-D diffusion code SRAC-CITATION based on the group constants that come from the collision probability transport code SRAC-PIJ. In this study the evaluated nuclear data libraries JENDL-3.3 and ENDF/B-VII.0 are used. The calculated most important reactor physics parameters are compared to the safety analysis report (SAR) values as well as earlier published MCNP results (numerically benchmark). It was found that the calculated results show a good agreement between the said libraries. Besides, in most cases the calculated results reveal a reasonable agreement with the SAR values (by General Atomic) as well as the MCNP results. In addition, this analysis can be used as the inputs for thermal-hydraulic calculations of the TRIGA fresh core in the steady state and pulse mode operation. Because of power peaking factors, power distributions and temperature reactivity coefficients are the most important reactor safety parameters for normal operation and transient safety analysis in research as well as in power reactors. They form the basis for technical specifications and limitations for reactor operation such as loading pattern limitations for pulse operation (in TRIGA). Therefore, this analysis will be very important to develop the nuclear safety parameters data of 3 MW TRIGA Mark

  14. Uranium-fuel thermal reactor benchmark testing of CENDL-3

    International Nuclear Information System (INIS)

    Liu Ping

    2001-01-01

    CENDL-3, the new version of China Evaluated Nuclear Data Library are being processed, and distributed for thermal reactor benchmark analysis recently. The processing was carried out using the NJOY nuclear data processing system. The calculations and analyses of uranium-fuel thermal assemblies TRX-1,2, BAPL-1,2,3, ZEEP-1,2,3 were done with lattice code WIMSD5A. The results were compared with the experimental results, the results of the '1986'WIMS library and the results based on ENDF/B-VI. (author)

  15. Hybrid Reactor Simulation and 3-D Information Display of BWR Out-of-Phase Oscillation

    International Nuclear Information System (INIS)

    Edwards, Robert; Huang, Zhengyu

    2001-01-01

    The real-time hybrid reactor simulation (HRS) capability of the Penn State TRIGA reactor has been expanded for boiling water reactor (BWR) out-of-phase behavior. During BWR out-of-phase oscillation half of the core can significantly oscillate out of phase with the other half, while the average power reported by the neutronic instrumentation may show a much lower amplitude for the oscillations. A description of the new HRS is given; three computers are employed to handle all the computations required, including real-time data processing and graph generation. BWR out-of-phase oscillation was successfully simulated. By adjusting the reactivity feedback gains from boiling channels to the TRIGA reactor and to the first harmonic mode power simulation, limit cycle can be generated with both reactor power and the simulated first harmonic power. A 3-D display of spatial power distributions of fundamental mode, first harmonic, and total powers over the reactor cross section is shown

  16. Application of RELAP5-3D code for thermal analysis of the ADS reactor core

    International Nuclear Information System (INIS)

    Fernandes, Gustavo Henrique Nazareno

    2018-01-01

    Nuclear power is essential to supply global energy demand. Therefore, in order to use nuclear fuel more efficiently, more efficient nuclear reactors technologies researches have been intensified, such as hybrid systems, composed of particle accelerators coupled into nuclear reactors. In order to add knowledge to such studies, an innovative reactor design was considered where the RELAP5-3D thermal-hydraulic analysis code was used to perform a thermal analysis of the core, either in stationary operation or in situations transitory. The addition of new kind of coolants, such as, liquid salts, among them Flibe, lead, lead-bismuth, sodium, lithium-bismuth and lithium-lead was an important advance in this version of the code, making possible to do the thermal simulation of reactors that use these types of coolants. The reactor, object of study in this work, is an innovative reactor, due to its ability to operate in association with an Accelerator Driven System (ADS), considered a predecessor system of the next generation of nuclear reactors (GEN IV). The reactor selected was the MYRRHA (Multi-purpose Hybrid Research Reactor for High tech Applications) due to the availability of data to perform the simulation. In the modeling of the reactor with the code RELAP5-3D, the core was simulated using nodules with 1, 7, 15 and 51 thermohydraulic channels and eutectic lead-bismuth (LBE) as coolant. The parameters, such as, pressure, mass flow and coolant and heat structure temperature were analyzed. In addition, the thermal behavior of the core was evaluated by varying the type of coolant (sodium) in substitution for the LBE of the original design using the model with 7 thermohydraulic channels. The results of the steady-state calculations were compared with data from the literature and the proposed models were verified certifying the ability of the RELAP5-3D code to simulate this innovative reactor. After this step, it was analysed cases of transients with loss of coolant flow

  17. Optimization of reactor power by taking into consideration temperature increase in a reactor pumped 3He-Xe laser

    International Nuclear Information System (INIS)

    Cetin, Fuesun

    2009-01-01

    In nuclear pumped lasers, gas parameters are optimized in a manner such that output power is increased for constructing a high power laser. Since output power increases with the increase of energy deposited in the gas, high output power requires high pumping power. However, the high energy loading results in elevated gas temperature. Temperature increase of this magnitude can detrimentally influence the laser gain and efficiency, since it negatively impacts several important laser kinetic.processes. This fact may cause laser output to abruptly terminate before the peak of the pump pulse [1-3]. A nuclear pumped laser using a volumetric energy source through the 3 He(n, p) 3 H reaction has here been considered. It is assumed that TRIGA Mark II Reactor at Istanbul Technical University is used for nuclear pumping as the neutron source. In the previous papers, the optimal parameters for improving both output power and optical homogeneity were determined [4-5]. Spatial and temporal variations of gas temperature during pumping pulse for maximum peak power (1200 MW) were determined for various operating pressures in Ref. [6]. It was seen that gas temperature reaches up to 1000 0 K near the peak of the pumping pulse for the initial pressures of 1-4 atm. This means that laser output may terminate before the peak of the pump pulse due to overheating of laser gas. Under these conditions, a question arises about a further optimisation taking into consideration gas temperature. This question has been examined in this study. Experimental results (Batyrbekov et al, 1989) showed that temperature rise up to 650 C had no influence on Xe laser characteristics [ 7]. Therefore, It has here been assumed that the lasing will terminate when gas temperature reaches 1000 0 K for a Xe-laser with 3 He buffer gas. Under these conditions optimum reactor power is investigated by taking into consideration lasing duration also. (orig.)

  18. A history of ZED-2

    International Nuclear Information System (INIS)

    Jones, R.

    2010-01-01

    The ZED-2 Reactor at Chalk River Laboratories was 50 years old this fall. First criticality occurred in September 1960. ZED-2 is perhaps not very well known in the Canadian Nuclear Industry, certainly not as well known as the various CANDU power reactors or the research reactors NRU and NRX. Part of the reason for this I suspect is that when casually judging the importance of reactors the first parameters that spring to mind are power generated (for power reactors) or neutron flux (for research reactors), bigger being 'better in both cases. By these standards ZED-2 does indeed appear puny: the maximum allowed power is 200W and the corresponding flux about 10 8 to 10 9 neutrons cm -2 s -1 , both numbers being about a factor of 500,000 smaller than the corresponding values for NRU. So, what is it all about? How is it that such an apparently insignificant reactor has operated for 50 years?, longer than any other Canadian reactor except NRU and the McMaster Reactor. What is it used for? What contributions has it made to the Canadian industry? Maybe one might also ask for how long is it going to continue? Well, that's what this talk is about, although I think I will leave the final question to wiser heads than mine. ZED-2 is a descendant of famous progenitors: starting with Enrico Farm's first critical pile of graphite and uranium (created at the University of Chicago in 1942) through Canada's ZEEP (first reactor to go critical outside the USA) that went critical in 1945. These early critical facilities were first about proof of principle that a self sustaining nuclear chain reaction could be established and controlled in a reasonable sized facility and second, in the longer term, developing understanding of the underlying reactor physics and the development of theories and methods to accurately predict the important properties of critical assemblies generally. (author)

  19. 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)

  20. 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)

  1. Simulation of A Main Steam Line Break Accident Using the Coupled 'System Thermal-Hydraulics, 3D reactor Kinetics, and Hot Channel' Analysis Capability of MARS 3.0

    International Nuclear Information System (INIS)

    Jeong, Jae Jun; Chung, Bub Dong

    2005-09-01

    For realistic analysis of thermal-hydraulics (T-H) transients in light water reactors, KAERI has developed the best-estimate T-H system code, MARS. The code has been improved from the consolidated version of the RELAP5/MOD3 and COBRA-TF codes. Then, the MARS code was coupled with a three-dimensional (3-D) reactor kinetics code, MASTER. This coupled calculation feature, in conjunction with the existing hot channel analysis capabilities of the MARS and MASTER codes, allows for more realistic simulations of nuclear system transients. In this work, a main steam line break (MSLB) accident is simulated using the coupled 'system T-H, 3-D reactor kinetics, and hot channel analysis' feature of the MARS code. Two coupled calculations are performed for demonstration. First, a coupled calculation of the 'system T-H and 3-D reactor kinetics' with a refined core T-H nodalization is carried out to obtain global core power and local departure from nucleate boiling (DNB) ratio (DNBR) behaviors. Next, for a more accurate DNBR prediction, another coupled calculation with subchannel meshes for the hot channels is performed. The results of the coupled calculations are very reasonable and consistent so that these can be used to remove the excessive conservatism in the conventional safety analysis

  2. Calculation of low-energy reactor neutrino spectra reactor for reactor neutrino experiments

    Energy Technology Data Exchange (ETDEWEB)

    Riyana, Eka Sapta; Suda, Shoya; Ishibashi, Kenji; Matsuura, Hideaki [Dept. of Applied Quantum Physics and Nuclear Engineering, Kyushu University, Kyushu (Japan); Katakura, Junichi [Dept. of Nuclear System Safety Engineering, Nagaoka University of Technology, Nagaoka (Japan)

    2016-06-15

    Nuclear reactors produce a great number of antielectron neutrinos mainly from beta-decay chains of fission products. Such neutrinos have energies mostly in MeV range. We are interested in neutrinos in a region of keV, since they may take part in special weak interactions. We calculate reactor antineutrino spectra especially in the low energy region. In this work we present neutrino spectrum from a typical pressurized water reactor (PWR) reactor core. To calculate neutrino spectra, we need information about all generated nuclides that emit neutrinos. They are mainly fission fragments, reaction products and trans-uranium nuclides that undergo negative beta decay. Information in relation to trans-uranium nuclide compositions and its evolution in time (burn-up process) were provided by a reactor code MVP-BURN. We used typical PWR parameter input for MVP-BURN code and assumed the reactor to be operated continuously for 1 year (12 months) in a steady thermal power (3.4 GWth). The PWR has three fuel compositions of 2.0, 3.5 and 4.1 wt% {sup 235}U contents. For preliminary calculation we adopted a standard burn-up chain model provided by MVP-BURN. The chain model treated 21 heavy nuclides and 50 fission products. The MVB-BURN code utilized JENDL 3.3 as nuclear data library. We confirm that the antielectron neutrino flux in the low energy region increases with burn-up of nuclear fuel. The antielectron-neutrino spectrum in low energy region is influenced by beta emitter nuclides with low Q value in beta decay (e.g. {sup 241}Pu) which is influenced by burp-up level: Low energy antielectron-neutrino spectra or emission rates increase when beta emitters with low Q value in beta decay accumulate. Our result shows the flux of low energy reactor neutrinos increases with burn-up of nuclear fuel.

  3. Fusion reactors as a future energy source

    International Nuclear Information System (INIS)

    Seifritz, W.

    A detailed update of fusion research concepts is given. Discussions are given for the following areas: (1) the magnetic confinement principle, (2) UWMAK I: conceptual design for a fusion reactor, (3) the inertial confinement principle, (4) the laser fusion power plant, (5) electron-induced fusion, (6) the long-term development potential of fusion reactors, (7) the symbiosis between fusion and fission reactors, (8) fuel supply for fusion reactors, (9) safety and environmental impact, and (10) accidents, and (11) waste removal and storage

  4. Design and implementation of the control system for the new console of TRIGA-3-Salazar Reactor

    International Nuclear Information System (INIS)

    Gonzalez M, J.L.

    1994-01-01

    TRIGA-3-Salazar Reactor was set in operation in 1968 and the aging of its components has cause the increasing in the maintenance. In the presence of this, it becomes necessary to replace the reactor console using new technologies, considering the incorporation of a personal computer. The aim of this work is the design and construction of the equipment interfaces as well as the digital computer program for the automation and control of the TRIGA-3-Salazar Reactor by means of a personal computer. (Author)

  5. Research on intelligent monitor for 3D power distribution of reactor core

    International Nuclear Information System (INIS)

    Xia, Hong; Li, Bin; Liu, Jianxin

    2014-01-01

    Highlights: • Core power distribution of ex-core measurement system has been reconstructed. • Building up an artificial intelligence model for 3-D core power distribution. • Error of the experiments has been reduced to 0.76%. • Methods for improving the accuracy of the model have been obtained. - Abstract: A real-time monitor for 3D reactor power distribution is critical for nuclear safety and high efficiency of NPP’s operation as well as for optimizing the control system, especially when the nuclear power plant (NPP) works at a certain power level or it works in load following operation. This paper was based on analyzing the monitor for 3D reactor power distribution technologies used in modern NPPs. Furthermore, considering the latest research outcomes, the paper proposed a method based on using an ex-core neutron detector system and a neural network to set up a real time monitor system for reactor’s 3D power distribution supervision. The results of the experiments performed on a reactor simulation machine illustrated that the new monitor system worked very well for a certain burn-up range during the fuel cycle. In addition, this new model could reduce the errors associated with the fitting of the distribution effectively, and several optimization methods were also obtained to improve the accuracy of the simulation model

  6. On the major DYN3D developments for fast reactor design and transient analysis

    International Nuclear Information System (INIS)

    Merk, B.; Kliem, S.

    2013-01-01

    Due to the French project ASTRID, the European CP-ESFR project, and the MYRRHA/FASTEF project, the research work on fast reactors has got a new push in Europe. Additionally to this European projects a strong project is growing in Russia based on the lead cooled fast reactor design BREST. Following this trend, the Institute of Resource Ecology at the Helmholtz-Zentrum Dresden-Rossendorf has decided to start several projects dedicated to fast reactor technology, among them the extension of the well validated LWR core simulator DYN3D. The new developments, first validation results, and the next strategic steps for the adaption of the code for the improved simulation of fast reactor cores are presented. (orig.)

  7. On the major DYN3D developments for fast reactor design and transient analysis

    Energy Technology Data Exchange (ETDEWEB)

    Merk, B.; Kliem, S. [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany). Reactor Safety Div.

    2013-07-01

    Due to the French project ASTRID, the European CP-ESFR project, and the MYRRHA/FASTEF project, the research work on fast reactors has got a new push in Europe. Additionally to this European projects a strong project is growing in Russia based on the lead cooled fast reactor design BREST. Following this trend, the Institute of Resource Ecology at the Helmholtz-Zentrum Dresden-Rossendorf has decided to start several projects dedicated to fast reactor technology, among them the extension of the well validated LWR core simulator DYN3D. The new developments, first validation results, and the next strategic steps for the adaption of the code for the improved simulation of fast reactor cores are presented. (orig.)

  8. Study of seismic responses of Candu-3 reactor building using isolator bearings

    International Nuclear Information System (INIS)

    Biswas, J.K.

    1992-01-01

    Seismic isolator bearings are known to increase reliability, reduce cost and increase the potential sitings for nuclear power plants located in regions of high seismicity. High seismic activities in Canada occur mainly in the western coast, the Grand Banks and regions of Quebec along the St. Lawrence river. In Canada, nuclear power plants are located in Ontario, Quebec and New Brunswick where the seismicity levels are low to moderate. Consequently, seismic isolator bearings have not been used in the existing nuclear power plants in Canada. The present paper examines the effect of using seismic isolator bearings in the design for the new CANDU3 which would be suitable for regions having high seismicity. The CANDU3 Nuclear Power Plant is rated at 450 MW of net output power and is a smaller version of its predecessor CANDU6 successfully operating in Canada and abroad. The design of CANDU3 is being developed by AECL CANDU. Advanced technologies for design, construction and plant operation have been utilized. During the conceptual development of the CANDU3 design, various design options including the use of isolator bearings were considered. The present paper presents an overview of seismic isolation technology and summarizes the analytical work for predicting the seismic behavior of the CANDU3 reactor building. A lumped-parameter dynamic model for the reactor building is used for the analysis. The characteristics of the bearings are utilized in the analysis work. The time-history modal analysis has been used to compute the seismic responses. Seismic responses of the reactor building with and without isolator bearings are compared. The isolator bearings are found to reduce the accelerations of the reactor building. As a result, a lower level of seismic qualification for components and systems would be required. The use of these bearings however increases rigid body seismic displacements of the structure requiring special considerations in the layout and interfaces for

  9. G2 and G3 reactors design; Description des reacteurs G2 et G3

    Energy Technology Data Exchange (ETDEWEB)

    Herreng,; Ertaud,; Pasquet, [Societe Alsacienne de Constructions Mecaniques (France)

    1958-07-01

    'FRANCE ATOME' Manufacturers Party has been entrusted with the G2 and G3 reactors engineering by the french A.E.C., for the first-five-year french project. Although these reactors are essentially plutonium generators, everyone has been linked with a power station which is supposed to supply with 40 MW, 'Electricite de France' has taken the liability upon itself. The reactor core includes most of G1 reactor parts (central gap excluded): horizontal channels, graphite parallelepipedic bricks stacking, steel thermal shield. The cooling is provided with CO{sub 2} under a 15 atmospheres pressure. This pressure is kept steady in a press-stressed concrete packing-case which is a cylinder horizontally shaped. Steel strips tightened encircle the concrete cylinder; itself protected by sole-plates. The cylinder bottom has brought about unusual problems which have been solved by the choice of an hemispheric shape. Packing-case tightness is provided by a 30 mm iron-plate connected with the inner wall of concrete. One of the reactor's special characteristics is the possibility of loading and unloading while operating. On loading side, barrel locks, each weighting 50 tons, allow new cans, at a pressure of 15 atmospheres, to pass. The cans process almost in a steady way through the channel, and finally drop down through bent spouts, then through spiral toboggans into a new lock. The cooling CO{sub 2} flow is provided with 3 turbo-bellows, these are actuated by average pressure-steam, obtained from exchangers. Every reactor supplies 4 exchangers which have been very difficult to build and to set up. The secondary cycle is standard and contains 3 stages (pressure 10,3: 2 and 0,5 kg/cm{sup 2}). Steam can be condensed in the event of a group turbo-generator stopping, with no modifion for the normal operating conditions of the reactor. Auxiliary circuits have to assure the continuous purifying of cooling CO{sub 2}, its storage and drain. 49 boron carbide rods are used to control the

  10. Incorporating higher order WINKLER springs with 3-D finite element model of a reactor building for seismic SSI analysis

    International Nuclear Information System (INIS)

    Ermutlu, H.E.

    1993-01-01

    In order to fulfill the seismic safety requirements, in the frame of seismic requalification activities for NPP Muehleberg, Switzerland, detailed seismic analysis performed on the Reactor Building and the results are presented previously. The primary objective of the present investigation is to assess the seismic safety of the reinforced concrete structures of reactor building. To achieve this objective requires a rather detailed 3-D finite element modeling for the outer shell structures, the drywell, the reactor pools, the floor decks and finally, the basemat. This already is a complicated task, which enforces need for simplifications in modelling the reactor internals and the foundation soil. Accordingly, all internal parts are modelled by vertical sticks and the Soil Structure Interaction (SSI) effects are represented by sets of transitional and higher order rotational WINKLER springs, i.e. avoiding complicated finite element SSI analysis. As a matter of fact, the availability of the results of recent investigations carried out on the reactor building using diversive finite element SSI analysis methods allow to calibrate the WINKLER springs, ensuring that the overall SSI behaviour of the reactor building is maintained

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

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  12. Preparations for the shipment of RA-3 reactor irradiated fuel

    International Nuclear Information System (INIS)

    Goldschmidt, Adrian; Novara, Oscar; Lafuente, Jose

    2002-01-01

    During the last quarter of 2000, in the Radioactive Waste Management Area of the Argentine National Commission of Atomic Energy (CNEA), located at Ezeiza Atomic Center (CAE), activities associated to the shipment of 207 MTR spent fuels containing high enrichment uranium were carried out within the Foreign Research Reactor/Domestic Research Reactor Receipt Program launched by the US Department of Energy (DOE). The MTR spent fuel shipped to Savannah River Site (SRS) was fabricated in Argentina with 90% enriched uranium of US origin and it was utilized in the operation of the research and radioisotope production reactor RA-3 from 1968 until 1987. After a cooling period at the reactor, the spent fuel was transferred to the Central Storage Facility (CSF) located in the waste management area of CAE for interim storage. The spent fuel (SF) inventory consisted of 166 standard assemblies (SA) and 41 control assemblies (CA). Basically, the activities performed were the fuel conditioning operations inside the storage facility (remote transference of the assemblies to the operation pool, fuel cropping, fuel re-identification, loading in transport baskets, etc.) conducted by CNEA. The loading of the filled baskets in the transport casks (NAC-LWT) by means of intermediate transfer systems and loaded casks final preparations were conducted by NAC personnel (DOE's contractor) with the support of CNEA personnel. (author)

  13. Simulation of the operational monitoring of a BWR with Simulate-3; Simulacion del seguimiento operacional de un reactor BWR con Simulate-3

    Energy Technology Data Exchange (ETDEWEB)

    Jimenez F, J. O.; Martin del Campo M, C.; Fuentes M, L.; Francois L, J. L., E-mail: ace.jo.cu@gmail.com [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico)

    2015-09-15

    This work was developed in order to describe the methodology for calculating the fuel burned of nuclear power reactors throughout the duration of their operating cycle and for each fuel reload. In other words, simulate and give monitoring to the main operation parameters of sequential way along its operation cycles. For this particular case, the operational monitoring of five consecutive cycles of a reactor was realized using the information reported by their processes computer. The simulation was performed with the Simulate-3 software and the results were compared with those of the process computer. The goal is to get the fuel burned, cycle after cycle for obtain the state conditions of the reactor needed for the fuel reload analyses, stability studies and transients analysis, and the development of a methodology that allows to manage and resolve similar cases for future fuel cycles of the nuclear power plant and explore the various options offered by the simulator. (Author)

  14. Application of Raptor-M3G to reactor dosimetry problems on massively parallel architectures - 026

    International Nuclear Information System (INIS)

    Longoni, G.

    2010-01-01

    The solution of complex 3-D radiation transport problems requires significant resources both in terms of computation time and memory availability. Therefore, parallel algorithms and multi-processor architectures are required to solve efficiently large 3-D radiation transport problems. This paper presents the application of RAPTOR-M3G (Rapid Parallel Transport Of Radiation - Multiple 3D Geometries) to reactor dosimetry problems. RAPTOR-M3G is a newly developed parallel computer code designed to solve the discrete ordinates (SN) equations on multi-processor computer architectures. This paper presents the results for a reactor dosimetry problem using a 3-D model of a commercial 2-loop pressurized water reactor (PWR). The accuracy and performance of RAPTOR-M3G will be analyzed and the numerical results obtained from the calculation will be compared directly to measurements of the neutron field in the reactor cavity air gap. The parallel performance of RAPTOR-M3G on massively parallel architectures, where the number of computing nodes is in the order of hundreds, will be analyzed up to four hundred processors. The performance results will be presented based on two supercomputing architectures: the POPLE supercomputer operated by the Pittsburgh Supercomputing Center and the Westinghouse computer cluster. The Westinghouse computer cluster is equipped with a standard Ethernet network connection and an InfiniBand R interconnects capable of a bandwidth in excess of 20 GBit/sec. Therefore, the impact of the network architecture on RAPTOR-M3G performance will be analyzed as well. (authors)

  15. 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

  16. 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)

  17. Investigation of flow stabilization in a compact reactor vessel of a FBR. Flow visualization in a reactor vessel

    International Nuclear Information System (INIS)

    Sato, Hiroyuki; Igarashi, Minoru; Kimura, Nobuyuki; Kamide, Hideki

    2002-01-01

    In the feasibility studies of Commercialized Fast Breeder Reactor Cycle System, a compact reactor vessel is considered from economical improvement point of a sodium cooled loop type fast reactor. The flow field was visualized by water experiment for a reactor vessel with 'a column type UIS (Upper Internal Structure)', which has a slit for fuel handling mechanism and is useful for a compact fast reactor. In this research, the 1/20 scale test equipment using water was made to understand coolant flow through a slit of a column type UIS' and fundamental behavior of reactor upper plenum flow. In the flow visualization tests, tracer particles were added in the water, and illuminated by the slit-shaped pulse laser. The flow visualization image was taken with a CCD camera. We obtained fluid velocity vectors from the visualization image using the Particle Imaging Velocimetry (PIV). The results are as follows. 1. Most of coolant flow through a slit of 'column type UIS' arrived the dip plate directly. In the opposite side of a slit, most of coolant flowed toward reactor vessel wall before it arrived the dip plate. 2. The PIV was useful to measure the flow field in the reactor vessel. The obtained velocity field was consistent with the flow visualization result. 3. The jet through the UIS slit was dependent on the UIS geometry. There is a possibility to control the jet by the UIS geometry. (author)

  18. Evaluation of tritium production rate in a gas-cooled reactor with continuous tritium recovery system for fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Matsuura, Hideaki, E-mail: mat@nucl.kyushu-u.ac.jp [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Fukuoka 819-0395 (Japan); Nakaya, Hiroyuki; Nakao, Yasuyuki [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Fukuoka 819-0395 (Japan); Shimakawa, Satoshi; Goto, Minoru; Nakagawa, Shigeaki [Japan Atomic Energy Agency, 4002 Oarai, Ibaraki 311-1393 (Japan); Nishikawa, Masabumi [Graduate School of Engineering Science, Kyushu University, 6-10-1 Hakozaki, Fukuoka 812-8581 (Japan)

    2013-10-15

    Highlights: • The performance of a gas-cooled reactor as a tritium production system was studied. • A continuous tritium recovery using helium gas was considered. • Gas-cooled reactors with 3 GW output in all can produce ∼6 kg of tritium in a year • Performance of the system was examined for Li{sub 4}SiO{sub 4}, Li{sub 2}TiO{sub 3} and LiAlO{sub 2} compounds. -- Abstract: The performance of a high-temperature gas-cooled reactor as a tritium production with continuous tritium recovery system is examined. A gas turbine high-temperature reactor of 300-MWe (600 MW) nominal capacity (GTHTR300) is assumed as the calculation target, and using the continuous-energy Monte Carlo transport code MVP-BURN, burn-up simulations for the three-dimensional entire-core region of the GTHTR300 were performed. A Li loading pattern for the continuous tritium recovery system in the gas-cooled reactor is presented. It is shown that module gas-cooled reactors with a total thermal output power of 3 GW in all can produce ∼6 kg of tritium maximum in a year.

  19. The ARIES-III D-3He tokamak reactor

    International Nuclear Information System (INIS)

    Bathke, C.G.; Werley, K.A.; Miller, R.L.; Krakowski, R.A.; Santarius, J.F.

    1992-01-01

    The multi-institutional ARIES study has generated a conceptual design of another tokamak fusion reactor in a series that varies the assumed advances in technology and physics. The ARIES-III design uses a D- 3 He fuel cycle and requires advances in technology and physics for economical attractiveness. The optimal design was characterized through systems analyses for eventual conceptual engineering design. In this paper, results from the systems analysis are summarized, and a comparison with the high-field, D-T fueled ARIES-I is included

  20. Licensed reactor nuclear safety criteria applicable to DOE reactors

    International Nuclear Information System (INIS)

    1991-04-01

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

  1. 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

  2. A study of reactor neutrino monitoring at the experimental fast reactor JOYO

    International Nuclear Information System (INIS)

    Furuta, H.; Fukuda, Y.; Hara, T.; Haruna, T.; Ishihara, N.; Ishitsuka, M.; Ito, C.; Katsumata, M.; Kawasaki, T.; Konno, T.; Kuze, M.; Maeda, J.; Matsubara, T.; Miyata, H.; Nagasaka, Y.; Nitta, K.; Sakamoto, Y.; Suekane, F.; Sumiyoshi, T.; Tabata, H.

    2012-01-01

    We carried out a study of neutrino detection at the experimental fast reactor JOYO using a 0.76 tons gadolinium loaded liquid scintillator detector. The detector was set up on the ground level at 24.3 m from the JOYO reactor core of 140 MW thermal power. The measured neutrino event rate from reactor on-off comparison was 1.11±1.24(stat.)±0.46(syst.) events/day. Although the statistical significance of the measurement was not enough, backgrounds in such a compact detector at the ground level were studied in detail and MC simulations were found to describe the data well. A study for improvement of the detector for future such experiments is also shown.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-01-01

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

  4. Neutrino-4 experiment on the search for a sterile neutrino at the SM-3 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Serebrov, A. P., E-mail: serebrov@pnpi.spb.ru; Ivochkin, V. G.; Samoylov, R. M.; Fomin, A. K.; Zinoviev, V. G.; Neustroev, P. V.; Golovtsov, V. L.; Gruzinsky, N. V.; Solovey, V. A.; Chernyi, A. V.; Zherebtsov, O. M. [National Research Centre “Kurchatov Institute,”, Konstantinov Petersburg Nuclear Physics Institute (Russian Federation); Martemyanov, V. P.; Tsinoev, V. G.; Tarasenkov, V. G.; Aleshin, V. I. [National Research Centre “Kurchatov Institute,” (Russian Federation); Petelin, A. L.; Pavlov, S. V.; Izhutov, A. L.; Sazontov, S. A.; Ryazanov, D. K. [State Scientific Centre Research Institute of Atomic Reactors (Russian Federation); and others

    2015-10-15

    In view of the possibility of the existence of a sterile neutrino, test measurements of the dependence of the reactor antineutrino flux on the distance from the reactor core has been performed on SM-2 reactor with the Neutrino-2 detector model in the range of 6–11 m. Prospects of the search for reactor antineutrinos at short distances have been discussed.

  5. 3. International conference on catalysis in membrane reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-09-01

    The 3. International Conference on Catalysis in Membrane Reactors, Copenhagen, Denmark, is a continuation of the previous conferences held in Villeurbanne 1994 and Moscow 1996 and will deal with the rapid developments taking place within membranes with emphasis on membrane catalysis. The approx. 80 contributions in form of plenary lectures and posters discuss hydrogen production, methane reforming into syngas, selectivity and specificity of various membranes etc. The conference is organised by the Danish Catalytic Society under the Danish Society for Chemical Engineering. (EG)

  6. Gas pollutant cleaning by a membrane reactor

    Directory of Open Access Journals (Sweden)

    Kaldis Sotiris

    2006-01-01

    Full Text Available An alternative technology for the removal of gas pollutants at the integrated gasification combined cycle process for power generation is the use of a catalytic membrane reactor. In the present study, ammonia decomposition in a catalytic reactor, with a simultaneous removal of hydrogen through a ceramic membrane, was investigated. A Ni/Al2O3 catalyst was prepared by the dry and wet impregnation method and characterized by the inductively coupled plasma method, scanning electron microscopy, X-ray diffraction, and N2 adsorption before and after activation. Commercially available a-Al2O3 membranes were also characterized and the permeabilities and permselectivities of H2, N2, and CO2 were measured by the variable volume method. In parallel with the experimental analysis, the necessary mathematical models were developed to describe the operation of the catalytic membrane reactor and to compare its performance with the conventional reactor. .

  7. MDS Nordion - a Canadian success story

    International Nuclear Information System (INIS)

    Boyd, F.

    1998-01-01

    MDS Nordion,the world leading supplier of radioisotopes, had its beginnings as a small division of the crown company Eldorado Mining and Refining Ltd. in the late 1940's. With the end of World War II, Eldorado found itself with a supply of radium left over from the production of uranium during the war and set up a small group in Ottawa to sell it. Most of the stock of radium was disposed of when the development of the NRX reactor made reactor produced radioisotopes possible. The company turned to selling these radioisotopes and soon focused on Cobalt 60 for its use in cancer therapy. In 1952, the group transferred to AECL (Atomic Energy of Canada Ltd.), as part of the Commercial Products Division. In 1988, as part of the federal government's move to privatization, the organization was split into two companies, Theratronics, which manufactures the cancer therapy machines and other equipment for the use of radioisotopes, and Nordion International, which continued the primary business of processing, packaging, selling radioisotopes. In 1991 Nordion International was sold to MDS Inc., a large international group of companies in the life and sciences fields with headquarters in Toronto. The name was changed to MDS Nordion in December 1996

  8. Hydrogen production using Rhodopseudomonas palustris WP 3-5 with hydrogen fermentation reactor effluent

    International Nuclear Information System (INIS)

    Chi-Mei Lee; Kuo-Tsang Hung

    2006-01-01

    The possibility of utilizing the dark hydrogen fermentation stage effluents for photo hydrogen production using purple non-sulfur bacteria should be elucidated. In the previous experiments, Rhodopseudomonas palustris WP3-5 was proven to efficiently produce hydrogen from the effluent of hydrogen fermentation reactors. The highest hydrogen production rate was obtained at a HRT value of 48 h when feeding a 5 fold effluent dilution from anaerobic hydrogen fermentation. Besides, hydrogen production occurred only when the NH 4 + concentration was below 17 mg-NH 4 + /l. Therefore, for successful fermentation effluent utilization, the most important things were to decrease the optimal HRT, increase the optimal substrate concentration and increase the tolerable ammonia concentration. In this study, a lab-scale serial photo-bioreactor was constructed. The reactor overall hydrogen production efficiency with synthetic wastewater exhibiting an organic acid profile identical to that of anaerobic hydrogen fermentation reactor effluent and with effluent from two anaerobic hydrogen fermentation reactors was evaluated. (authors)

  9. Research reactor core conversion guidebook. V. 3: Analytical verification (Appendices G and H)

    International Nuclear Information System (INIS)

    1992-04-01

    Volume 3 consists of Appendix G which contains detailed results of a safety-related benchmark problem for an idealized reactor and Appendix H which contains detailed comparisons of calculated and measured data for actual cores with moderately enriched uranium and low enriched uranium fuels. The results of the benchmark calculations in Appendix G are summarized in Chapter 7 of Volume 1 and the results of the comparisons between calculations and measurements are summarized in Chapter 8 of Volume 1. Both the approaches described in these appendices are very useful in ensuring that the calculational methods employed in the preparation of a Safety Report are accurate. As a first step, it is recommended that reactor operators/physicists use their own methods and codes to first calculate the benchmark problem, and then compare the results of calculations with measurements in their own reactor or in one of the reactors for which measured data is available in Appendix H. (author). Refs, figs and tabs

  10. Reactor science and technology: operation and control of reactors

    International Nuclear Information System (INIS)

    Qiu Junlong

    1994-01-01

    This article is a collection of short reports on reactor operation and research in China in 1991. The operation of and research activities linked with the Heavy Water Research Reactor, Swimming Pool Reactor and Miniature Neutron Source Reactor are briefly surveyed. A number of papers then follow on the developing strategies in Chinese fast breeder reactor technology including the conceptual design of an experimental fast reactor (FFR), theoretical studies of FFR thermo-hydraulics and a design for an immersed sodium flowmeter. Reactor physics studies cover a range of topics including several related to work on zero power reactors. The section on reactor safety analysis is concerned largely with the assessment of established, and the presentation of new, computer codes for use in PWR safety calculations. Experimental and theoretical studies of fuels and reactor materials for FBRs, PWRs, BWRs and fusion reactors are described. A final miscellaneous section covers Mo-Tc isotope production in the swimming pool reactor, convective heat transfer in tubes and diffusion of tritium through plastic/aluminium composite films and Li 2 SiO 3 . (UK)

  11. Effect of 3-D moderator flow configurations on the reactivity of CANDU nuclear reactors

    International Nuclear Information System (INIS)

    Zadeh, Foad Mehdi; Etienne, Stephane; Chambon, Richard; Marleau, Guy; Teyssedou, Alberto

    2017-01-01

    Highlights: • 3-D CFD simulations of CANDU-6 moderator flows are presented. • A thermal-hydraulic code using thermal physical fluid properties is used. • The numerical approach and convergence is validated against available data. • Flow configurations are correlated using Richardson’s number. • The interaction between moderator temperatures with reactivity is determined. - Abstract: The reactivity of nuclear reactors can be affected by thermal conditions prevailing within the moderator. In CANDU reactors, the moderator and the coolant are mechanically separated but not necessarily thermally isolated. Hence, any variation of moderator flow properties may change the reactivity. Until now, nuclear reactor calculations have been performed by assuming uniform moderator flow temperature distribution. However, CFD simulations have predicted large time dependent flow fluctuations taking place inside the calandria, which can bring about local temperature variations that can exceed 50 °C. This paper presents robust CANDU 3-D CFD moderator simulations coupled to neutronic calculations. The proposed methodology makes it possible to study not only different moderator flow configurations but also their effects on the reactor reactivity coefficient.

  12. Proceedings of 2. Yugoslav symposium on reactor physics, Part 3, Herceg Novi (Yugoslavia), 27-29 Sep 1966

    International Nuclear Information System (INIS)

    1966-01-01

    This Volume 3 of the Proceedings of 2. Yugoslav symposium on reactor physics includes three papers describing the following: model for spatial synthesis of automated control system of the GCR type reactor; model for analysis of hydrodynamic processes at the BHWR type reactors; mathematical model for safety analysis of heavy water power reactor

  13. Technical report on implementation of reactor internal 3D modeling and visual database system

    International Nuclear Information System (INIS)

    Kim, Yeun Seung; Eom, Young Sam; Lee, Suk Hee; Ryu, Seung Hyun

    1996-06-01

    In this report was described a prototype of reactor internal 3D modeling and VDB system for NSSS design quality improvement. For improving NSSS design quality several cases of the nuclear developed nation's integrated computer aided engineering system, such as Mitsubishi's NUWINGS (Japan), AECL's CANDID (Canada) and Duke Power's PASCE (USA) were studied. On the basis of these studies the strategy for NSSS design improvement system was extracted and detail work scope was implemented as follows : 3D modelling of the reactor internals were implemented by using the parametric solid modeler, a prototype system of design document computerization and database was suggested, and walk-through simulation integrated with 3D modeling and VDB was accomplished. Major effects of NSSS design quality improvement system by using 3D modeling and VDB are the plant design optimization by simulation, improving the reliability through the single design database system and engineering cost reduction by improving productivity and efficiency. For applying the VDB to full scope of NSSS system design, 3D modelings of reactor coolant system and nuclear fuel assembly and fuel rod were attached as appendix. 2 tabs., 31 figs., 7 refs. (Author) .new

  14. Technical report on implementation of reactor internal 3D modeling and visual database system

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yeun Seung; Eom, Young Sam; Lee, Suk Hee; Ryu, Seung Hyun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-06-01

    In this report was described a prototype of reactor internal 3D modeling and VDB system for NSSS design quality improvement. For improving NSSS design quality several cases of the nuclear developed nation`s integrated computer aided engineering system, such as Mitsubishi`s NUWINGS (Japan), AECL`s CANDID (Canada) and Duke Power`s PASCE (USA) were studied. On the basis of these studies the strategy for NSSS design improvement system was extracted and detail work scope was implemented as follows : 3D modelling of the reactor internals were implemented by using the parametric solid modeler, a prototype system of design document computerization and database was suggested, and walk-through simulation integrated with 3D modeling and VDB was accomplished. Major effects of NSSS design quality improvement system by using 3D modeling and VDB are the plant design optimization by simulation, improving the reliability through the single design database system and engineering cost reduction by improving productivity and efficiency. For applying the VDB to full scope of NSSS system design, 3D modelings of reactor coolant system and nuclear fuel assembly and fuel rod were attached as appendix. 2 tabs., 31 figs., 7 refs. (Author) .new.

  15. RB Research nuclear reactor RB reactor, Annual report for 2000

    International Nuclear Information System (INIS)

    Milosevic, M.

    2000-12-01

    Report on RB reactor operation during 2000 contains 3 parts. Part one contains a brief description of reactor operation and reactor components, relevant dosimetry data and radiation protection issues, personnel and financial data. Part two is devoted to maintenance of the reactor components, namely, fuel, heavy water, reactor vessel, heavy water circulation system, absorption rods and heavy water level-meters, maintenance of electronic, mechanical, electrical and auxiliary equipment. Part three contains data concerned with reactor operation and utilization with a comprehensive list of publications resulting from experiments done at the RB reactor. It contains data about reactor operation during previous 14 years, i.e. from 1986 - 2000

  16. Anaerobic Digestion of Sugarcane Vinasse Through a Methanogenic UASB Reactor Followed by a Packed Bed Reactor.

    Science.gov (United States)

    Cabrera-Díaz, A; Pereda-Reyes, I; Oliva-Merencio, D; Lebrero, R; Zaiat, M

    2017-12-01

    The anaerobic treatment of raw vinasse in a combined system consisting in two methanogenic reactors, up-flow anaerobic sludge blanket (UASB) + anaerobic packed bed reactors (APBR), was evaluated. The organic loading rate (OLR) was varied, and the best condition for the combined system was 12.5 kg COD m -3 day -1 with averages of 0.289 m 3 CH 4  kg COD r -1 for the UASB reactor and 4.4 kg COD m -3 day -1 with 0.207 m 3 CH 4  kg COD r -1 for APBR. The OLR played a major role in the emission of H 2 S conducting to relatively stable quality of biogas emitted from the APBR, with H 2 S concentrations <10 mg L -1 . The importance of the sulphate to COD ratio was demonstrated as a result of the low biogas quality recorded at the lowest ratio. It was possible to develop a proper anaerobic digestion of raw vinasse through the combined system with COD removal efficiency of 86.7% and higher CH 4 and a lower H 2 S content in biogas.

  17. Some local dilution transient in a pressurized water reactor

    International Nuclear Information System (INIS)

    Jacobson, S.

    1989-01-01

    Reactivity accidents are important in the safety analysis of a pressurized water reactor. In this anlysis ejected control rod, steam line break, start of in-active loop and boron dilution accidents are usually dealt with. However, in the analysis is not included what reactivity excursions might happen when a zone,depleted of boron passes the reactor core. This thesis investigates during what operation and emergency conditions diluted zones might exist in a pressurized water reactor and what should be the maximum volumes for then. The limiting transport means are also established in terms of reactivty addition, for the depleted zones. In order to describe the complicated mixing process in the reactor vessel during such a transportation, a typical 3-loop reactor vessel has been modulated by means of TRAC-PF1's VESSEL component. Three cases have been analysed. In the first case the reactor is in a cold condition and the ractor coolant has boron concentration of 2000 ppm. To the reactor vessel is injected an clean water colume of 14 m 3 . In the two other cases the reactor is close to hot shutdown and borated to 850 ppm. To the reactor vessel is added 41 and 13 m 3 clean water, respectively. In the thesis is shown what spatial distribution the depleted zone gets when passing through the reactor vessel in the three cases. The boron concentration in the first case did not decrease the values which would bring the reactor to critical condition. For case two was shown by means of TRAC's point kinetics model that the reactor reaches prompt criticality after 16.03 seconds after starting of the reactor coolant pump. Another prompt criticality occured two seconds later. The total energy developed during the two power escalations were about 55 GJ. A comparision with the criteria used to evaluate the ejected control rod reactivity transient showed that none of these criteria were exceeded. (64 figs.)

  18. Burn up calculations for the Iranian miniature reactor: A reliable and safe research reactor

    International Nuclear Information System (INIS)

    Faghihi, F.; Mirvakili, S.M.

    2009-01-01

    Presenting neutronic calculations pertaining to the Iranian miniature research reactor is the main goal of this article. This is a key to maintaining safe and reliable core operation. The following reactor core neutronic parameters were calculated: clean cold core excess reactivity (ρ ex ), control rod and shim worth, shut down margin (SDM), neutron flux distribution of the reactor core components, and reactivity feedback coefficients. Calculations for the fuel burnup and radionuclide inventory of the Iranian miniature neutron source reactor (MNSR), after 13 years of operational time, are carried out. Moreover, the amount of uranium burnup and produced plutonium, the concentrations and activities of the most important fission products, the actinide radionuclides accumulated, and the total radioactivity of the core are estimated. Flux distribution for both water and fuel temperature increases are calculated and changes of the central control rod position are investigated as well. Standard neutronic simulation codes WIMS-D4 and CITATION are employed for these studies. The input model was validated by the experimental data according to the final safety analysis report (FSAR) of the reactor. The total activity of the MNSR core is calculated including all radionuclides at the end of the core life and it is found to be equal to 1.3 x 10 3 Ci. Our investigation shows that the reactor is operating under safe and reliable conditions.

  19. Burn up calculations for the Iranian miniature reactor: A reliable and safe research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Faghihi, F. [Department of Nuclear Engineering, School of Engineering, Shiraz University, Shiraz 71345 (Iran, Islamic Republic of); Research Center for Radiation Protection, Shiraz University, Shiraz (Iran, Islamic Republic of)], E-mail: faghihif@shirazu.ac.ir; Mirvakili, S.M. [Department of Nuclear Engineering, School of Engineering, Shiraz University, Shiraz 71345 (Iran, Islamic Republic of)

    2009-06-15

    Presenting neutronic calculations pertaining to the Iranian miniature research reactor is the main goal of this article. This is a key to maintaining safe and reliable core operation. The following reactor core neutronic parameters were calculated: clean cold core excess reactivity ({rho}{sub ex}), control rod and shim worth, shut down margin (SDM), neutron flux distribution of the reactor core components, and reactivity feedback coefficients. Calculations for the fuel burnup and radionuclide inventory of the Iranian miniature neutron source reactor (MNSR), after 13 years of operational time, are carried out. Moreover, the amount of uranium burnup and produced plutonium, the concentrations and activities of the most important fission products, the actinide radionuclides accumulated, and the total radioactivity of the core are estimated. Flux distribution for both water and fuel temperature increases are calculated and changes of the central control rod position are investigated as well. Standard neutronic simulation codes WIMS-D4 and CITATION are employed for these studies. The input model was validated by the experimental data according to the final safety analysis report (FSAR) of the reactor. The total activity of the MNSR core is calculated including all radionuclides at the end of the core life and it is found to be equal to 1.3 x 10{sup 3}Ci. Our investigation shows that the reactor is operating under safe and reliable conditions.

  20. MULTI - A multigroup or multipoint P{sub 3} programme for calculating thermal neutron spectra in a reactor cell

    Energy Technology Data Exchange (ETDEWEB)

    Matausek, M V [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)

    1968-06-15

    Programme MULTI calculates the space energy distribution of thermal neutrons in a multizone, cylindrical, infinitely long reactor lattice by using the multigroup or multipoint P{sub 3} approximation. This report presents a short description of the algorithm and the programme and gives the instructions for its exploitation. (author)

  1. 3-D seismic response of a base-isolated fast reactor

    International Nuclear Information System (INIS)

    Kitamura, S.; Morishita, M.; Iwata, K.

    1992-01-01

    This paper describes a 3-D response analysis methodology development and its application to a base-isolated fast breeder reactor (FBR) plant. At first, studies on application of a base-isolation system to an FBR plant were performed to identify a range of appropriate characteristics of the system. A response analysis method was developed based on mathematical models for the restoring force characteristics of several types of the systems. A series of shaking table tests using a small scale model was carried out to verify the analysis method. A good agreement was seen between the test and analysis results in terms of the horizontal and vertical responses. Parametric studies were then made to assess the effects of various factors which might be influential to the seismic response of the system. Moreover, the method was applied to evaluate three-dimensional response of the base-isolated FBR. (author)

  2. Moderators for the design of a cold neutron source for the RA 3 reactor

    International Nuclear Information System (INIS)

    Cantargi, F; Sbaffoni, M; Granada, R

    2004-01-01

    The cold neutron production of hydrogenous materials was studied, taking into account their radiation resistance, for the conceptual design of a cold neutron source for the RA-3 reactor.Low spontaneous release of chemical energy was found in mesitylene.Libraries for hidrogen in mesitylene were generated using the NJOY nuclear processing system and the resulting cross sections were compared with experimental data.Good agreement between measurements and calculations was found in those cases where data are available.New calculations using the RA-3 geometry and these validated libraries will be performed [es

  3. Master-3.0: multi-purpose analyzer for static and transient effects of reactors

    International Nuclear Information System (INIS)

    Cho, Byung Oh; Joo, Han Gyu; Cho, Jin Young; Song, Jae Seung; Zee, Sung Quun

    2002-03-01

    MASTER-3.0 (Multi-purpose Analyzer for Static and Transient Effects of Reactors) is a nuclear design code based on the multi-group diffusion theory to calculate the steady-state and transient pressurized water reactor core in a 3-dimensional Cartesian or hexagonal geometry. Its neutronics model solves the space-time dependent neutron diffusion equations with NIM (Nodal Integration Method), NEM (Nodal Expansion Method), AFEN (Analytic Function Expansion Nodal Method)/NEM Hybrid Method, NNEM (Non-linear Nodal Expansion Method) or NANM (Non-linear Analytic Nodal Method) for a Cartesian geometry and with NTPEN (Non-linear Triangle-based Polynomial Expansion Nodal Method), AFEN (Analytic Function Expansion Nodal)/NEM Hybrid Method or NLFM (Non-linear Local Fine-Mesh Method) for a hexagonal one. Coarse mesh rebalancing, Krylov Subspace method, energy group restriction/prolongation method and asymptotic extrapolation method are implemented to accelerate the convergence of iteration process. MASTER-3.0 performs microscopic depletion calculations using microscopic cross sections provided by CASMO-3 or HELIOS and also has the reconstruction capability of pin information by use of MSS-IAS (Method of Successive Smoothing with Improved Analytic Solution). For the thermal-hydraulic calculation, fuel temperature table or COBRA3-C/P or MATRA model can be used selectively. In addition, MASTER-3.0 is designed to cover various PWRs including SMART as well as WH- and CE-type reactors, providing all data required in their design procedures

  4. Simulation of A Main Steam Line Break Accident Using the Coupled 'System Thermal-Hydraulics, 3D reactor Kinetics, and Hot Channel' Analysis Capability of MARS 3.0

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Jae Jun; Chung, Bub Dong

    2005-09-15

    For realistic analysis of thermal-hydraulics (T-H) transients in light water reactors, KAERI has developed the best-estimate T-H system code, MARS. The code has been improved from the consolidated version of the RELAP5/MOD3 and COBRA-TF codes. Then, the MARS code was coupled with a three-dimensional (3-D) reactor kinetics code, MASTER. This coupled calculation feature, in conjunction with the existing hot channel analysis capabilities of the MARS and MASTER codes, allows for more realistic simulations of nuclear system transients. In this work, a main steam line break (MSLB) accident is simulated using the coupled 'system T-H, 3-D reactor kinetics, and hot channel analysis' feature of the MARS code. Two coupled calculations are performed for demonstration. First, a coupled calculation of the 'system T-H and 3-D reactor kinetics' with a refined core T-H nodalization is carried out to obtain global core power and local departure from nucleate boiling (DNB) ratio (DNBR) behaviors. Next, for a more accurate DNBR prediction, another coupled calculation with subchannel meshes for the hot channels is performed. The results of the coupled calculations are very reasonable and consistent so that these can be used to remove the excessive conservatism in the conventional safety analysis.

  5. TRIO a general computer code for reactor 3-D flows analysis. Application to a LMFBR hot plenum

    International Nuclear Information System (INIS)

    Magnaud, J.P.; Rouzaud, P.

    1985-09-01

    TRIO is a code developed at CEA to investigate general incompressible 2D and 3D viscous flows. Two calculations are presented: the lid driven cubic cavity at Re=400; steady state (velocity and temperature field) of a LMFBR hot plenum, carried out in order to prepare the calculation of a cold shock consecutive to a reactor scram. 8 refs., 26 figs.

  6. Comparison and analysis of 1D/2D/3D neutronics modeling for a fusion reactor

    International Nuclear Information System (INIS)

    Li, J.; Zeng, Q.; Chen, M.; Jiang, J.; Wu, Y.

    2007-01-01

    During the course of analyzing the characteristics for fusion reactors, the refined calculations are needed to confirm that the nuclear design requirements are met. Since the long computational time is consumed, the refined three-dimensional (3D) representation has been used primarily for establishing the baseline reference values, analyzing problems which cannot be reduced by symmetry considerations to lower dimensions, or where a high level of accuracy is desired locally. The two-dimensional (2D) or one-dimensional (1D) description leads itself readily to resolve many problems, such as the studies for the material fraction optimization, or for the blanket size optimization. The purpose of this paper is to find out the differences among different geometric descriptions, which can guide the way to approximate and simplify the computational model. The fusion power reactor named FDS-II was designed as an advanced fusion power reactor to demonstrate and validate the commercialization of fusion power by Institute of Plasma Physics, Chinese Academy of Science. In this contribution, the dual-cooled lithium lead (DLL) blanket of FDS-II was used as a reference for neutronics comparisons and analyses. The geometric descriptions include 1D concentric sphere model, 1D, 2D and 3D cylinder models. The home-developed multi-functional neutronics analysis code system VisualBUS, the Monte Carlo transport code MCNP and nuclear data library HENDL have been used for these analyses. The neutron wall loading distribution, tritium breeding ratio (TBR) and nuclear heat were calculated to evaluate the nuclear performance. The 3D calculation has been used as a comparison reference because it has the least errors in the treatment of geometry. It is suggested that the value of TBR calculated by the 1D approach should be greater than 1.3 to satisfy the practical need of tritium self-sufficiency. The distribution of nuclear heat based on the 2D and 3D models were similar since they all consider

  7. Core configuration of a gas-cooled reactor as a tritium production device for fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nakaya, H., E-mail: nakaya@nucl.kyushu-u.ac.jp [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Fukuoka 8190395 (Japan); Matsuura, H.; Nakao, Y. [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Fukuoka 8190395 (Japan); Shimakawa, S.; Goto, M.; Nakagawa, S. [Japan Atomic Energy Agency, 4002 Oarai, Ibaraki (Japan); Nishikawa, M. [Malaysia-Japan International Institute of Technology, UTM, Kuala Lumpur 54100 (Malaysia)

    2014-05-01

    The performance of a high-temperature gas-cooled reactor as a tritium production device is examined, assuming the compound LiAlO{sub 2} as the tritium-producing material. A gas turbine high-temperature reactor of 300 MWe nominal capacity (GTHTR300) is assumed as the calculation target, and using the continuous-energy Monte Carlo transport code MVP-BURN, burn-up simulations are carried out. To load sufficient Li into the core, LiAlO{sub 2} is loaded into the removable reflectors that surround the ring-shaped fuel blocks in addition to the burnable poison insertion holes. It is shown that module high-temperature gas-cooled reactors with a total thermal output power of 3 GW can produce almost 8 kg of tritium in a year.

  8. 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)

  9. Advances in Reactor physics, mathematics and computation. Volume 3

    Energy Technology Data Exchange (ETDEWEB)

    1987-01-01

    These proceedings of the international topical meeting on advances in reactor physics, mathematics and computation, volume 3, are divided into sessions bearing on: - poster sessions on benchmark and codes: 35 conferences - review of status of assembly spectrum codes: 9 conferences - Numerical methods in fluid mechanics and thermal hydraulics: 16 conferences - stochastic transport and methods: 7 conferences.

  10. Annealing of the BR3 reactor pressure vessel

    International Nuclear Information System (INIS)

    Fabry, A.; Motte, F.; Stiennon, G.; Debrue, J.; Gubel, P.; Van de Velde, J.; Minsart, G.; Van Asbroeck, P.

    1985-01-01

    The pressure vessel of the Belgian BR-3 plant, a small (11 MWe) PWR presently used for fuel testing programs and operated since 1962, was annealed during March, 1984. The anneal was performed under wet conditions for 168 hours at 650 0 F with core removal and within plant design margins justification for the anneal, summary of plant characteristics, description of materials sampling, summary of reactor physics and dosimetry, development of embrittlement trend curves, hypothesized pressurized and overcooling thermal shock accidents, and conclusions are provided in detail

  11. A multi-purpose reactor

    International Nuclear Information System (INIS)

    Changwen Ma

    2000-01-01

    An integrated natural circulation self pressurized reactor can be used for sea water desalination, electrogeneration, ship propulsion and district or process heating. The reactor can be used for ship propulsion because it has following advantages: it is a integrated reactor. Whole primary loop is included in a size limited pressure vessel. For a 200 MW reactor the diameter of the pressure vessel is about 5 m. It is convenient to arranged on a ship. Hydraulic driving facility of control rods is used on the reactor. It notably decreases the height of the reactor. For ship propulsion, smaller diameter and smaller height are important. Besides these, the operation reliability of the reactor is high enough, because there is no rotational machine (for example, circulating pump) in safety systems. Reactor systems are simple. There are no emergency water injection system and boron concentration regulating system. These features for ship propulsion reactor are valuable. Design of the reactor is based on existing demonstration district heating reactor design. The mechanic design principles are the same. But boiling is introduced in the reactor core. Several variants to use the reactor as a movable seawater desalination plant are presented in the paper. When the sea water desalination plant is working to produce fresh water, the reactor can supply electricity at the same time to the local electricity network. Some analyses for comprehensive application of the reactor have been done. Main features and parameters of the small (Thermopower 200 MW) reactor are given in the paper. (author)

  12. Gas pollutant cleaning by a membrane reactor

    Energy Technology Data Exchange (ETDEWEB)

    George E. Skodras; Sotiris Kaldis; Savas G. Topis; Dimitris Koutsonikolas; George P. Sakellaropoulos [Aristotle University of Thessaloniki, Thessaloniki (Greece). Chemical Process Engineering Laboratory, Dept. of Chemical Engineering

    2006-07-01

    An alternative technology for the removal of gas pollutants at the intergrated gasification combined cycle process for power generation is the use of a catalytic membrane reactor. In the present study, ammonia decomposition in a catalytic reactor, with a simultaneous removal of hydrogen through a ceramic membrane, was investigated. A Ni/Al{sub 2}O{sub 3} catalyst was prepared by the dry and wet impregnation method and characterized by ICP, SEM, XRD and N{sub 2} adsorption before and after activation. Commercially available {alpha}-Al{sub 2}O{sub 3} membranes were also characterized and the permeabilities and permselectivities of H{sub 2}, N{sub 2} and CO{sub 2} were measured by the variable volume method. In parallel with the experimental analysis, the necessary mathematical models were developed to describe the operation of the catalytic membrane reactor and to compare its performance with the conventional reactor. 9 refs., 6 figs., 1 tab.

  13. Fast Reactor Physics. Vol. II. Proceedings of a Symposium on Fast Reactor Physics and Related Safety Problems

    International Nuclear Information System (INIS)

    1968-01-01

    Proceedings of a Symposium organized by the IAEA and held in Karlsruhe, 30 October - 3 November 1967. The meeting was attended by 183 scientists from 23 countries and three international organizations. Contents: (Vol.1) Review of national programmes (5 papers); Nuclear data for fast reactors (12 papers); Experimental methods (3 papers); Zoned systems (7 papers); Kinetics (7 papers). (Vol.11) Fast critical experiments (8 papers); Heterogeneity in fast critical experiments (5 papers); Fast power reactors (13 papers); Fast pulsed reactors (3 papers); Panel discussion. Each paper is in its original language (50 English, 11 French and 3 Russian) and is preceded by an abstract in English with a second one in the original language if this is not English. Discussions are in English. (author)

  14. Investigation of slightly forced buoyant flow in a training reactor

    International Nuclear Information System (INIS)

    Legradi, G.; Aszodi, A.; Por, G.

    2001-01-01

    A measurement based on the temperature noise analysis method was carried out in the Training Reactor of the Budapest University of Technology and Economics. The main goals were the estimation of the flow velocity immediately above the reactor core and investigation of the thermal-hydraulical conditions of the reactor, mainly in the core. Subsequently 2D and 3D computations were carried out with the aid of the code CFX- 4.3. The main objective of the 2D calculation was to clarify the thermal-hydraulical conditions of the whole reactor tank with a reasonable computing demand. It was also necessary to accomplish 3D numerical investigations of the reactor core and the space above since three dimensional effects of the flow could only be studied in this way. In addition, obtaining certain boundary conditions of the 3D computations was another significant aim of the 2D investigations. It is important that the results of the noise analysis and the operational measuring system of the reactor gave us a basis for verifying our computations.(author)

  15. Gas pollutant cleaning by a membrane reactor

    Energy Technology Data Exchange (ETDEWEB)

    Topis, S.; Koutsonikolas, D.; Kaldis, S. (and others) [Aristotle University of Thessaloniki, Thessaloniki (Greece). Dept. of Chemical Engineering

    2005-07-01

    An alternative technology for the removal of gas pollutants at the integrated gasification combined cycle process for power generation is the use of a catalytic membrane reactor. In the present study, ammonia decomposition in a catalytic reactor, with simultaneous removal of hydrogen through a ceramic membrane, was investigated. A Ni/Al{sub 2}O{sub 3} catalyst was prepared by the dry and wet impregnation method and characterized by ICP, SEM, XRD and N{sub 2} adsorption before and after activation. Commercially available {alpha}-Al{sub 2}O{sub 3} membranes were also characterized and the permeabilities and selectivities of H{sub 2}, N{sub 2} and CO{sub 2} were measured by the variable volume method. In parallel with the experimental analysis, the necessary mathematical models were developed to describe the operation of the catalytic membrane reactor and to compare its performance with the conventional reactor. 5 refs., 6 figs., 1 tab.

  16. 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.

  17. Review on the seismic safety of JRR-3 according to the revised regulatory code on seismic design for nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kobayashi, Tetsuya; Araki, Masaaki; Ohba, Toshinobu; Torii, Yoshiya [Japan Atomic Energy Agency, Tokai, Ibaraki (Japan); Takeuchi, Masaki [Nuclear Safety Commission (Japan)

    2012-03-15

    JRR-3(Japan Research Reactor No.3) with the thermal power of 20MW is a light water moderated and cooled, swimming pool type research reactor. JRR-3 has been operated without major troubles. This paper presents about review on the seismic safety of JRR-3 according to the revised regulatory code on seismic design for nuclear reactors. In addition, some topics concerning damages in JRR-3 due to the Great East Japan Earthquake are presented. (author)

  18. RELAP5/Mod3.3 and MARS3.0a Modeling of a Siphon Break Experiment

    International Nuclear Information System (INIS)

    Park, Su Ki; Kim, Heon Il; Park, Cheol; Yoon, Ju Hyeon

    2011-01-01

    Pool water plays a very important role as a final heat sink for most pool-type research reactors following postulated events. Therefore, one of design criteria for the reactors is that the water level of reactor pool must not decrease below a predefined elevation even against the most severe accident due to ruptures of coolant boundary of connecting systems to the reactor pool. In order to accomplish the design criterion, all the connecting systems are usually arranged to be above the elevation of reactor core. However, some research reactors with a downward flow in the reactor core have a primary cooling system located below the elevation of reactor core because of meeting an available net positive suction head of pumps in the system. These reactors have a provision consisting of pipes penetrating a reactor pool wall at a higher elevation than that of reactor core and siphon break devices to meet the design criterion. A series of experiments was carried out to figure out thermal hydraulic characteristics during siphon is blocked and establish design requirements for siphon breaker. The experimental study provided a lot of data and observations to the process of siphon break, but it does not provide a sufficient theoretical analysis and present practical design requirements applicable to industry. The experimental range is not also sufficient to cover operating conditions of siphon breakers for research reactors. A series of numerical simulations on the experimental data has been tried by using thermal hydraulic system analysis codes, RELAP5/Mod3.3 and MARS3.0a. This paper includes a part of the numerical simulations. First output from this study shows an importance of an adequate use of thermal hydraulic models in the codes and a big different prediction between the two codes especially in relation to the use of choked flow option. From this study, it seems that RELAP5/Mod3.3 has some problems on the control of a choked flow option-flag or the prediction of a

  19. RA reactor exploitation, task 3.08/01; Zadatak 3.08/01 - Eksploatacija reaktora RA

    Energy Technology Data Exchange (ETDEWEB)

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

    1963-12-15

    During 1963 the RA reactor was operated for 1852 hours at mean power of 5.7 MW (total power production was 10716 MWh). Reactor was used for irradiation according to the demand of 356 users, and 15 experiments. The reason for decreased operation in comparison with the previous year was repair of all the reactor equipment and decontamination of the heavy water system. This report contains detailed data about reactor power, reactivity changes and fuel burnup. Mean monthly usage of the reactor experimental channels as well as samples which were irradiated are part of this report.

  20. Operating reactors licensing actions summary. Vol. 3, No. 6

    International Nuclear Information System (INIS)

    1983-07-01

    The operating reactors licensing actions summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Management and Program Analysis. This summary report is published primarily for internal NRC use in managing the operating reactors licensing actions program. Its content will change based on NRC management informational requirements

  1. Numerical investigation of the 3-dimensional steady-state temperature- and flow distribution in the core of a pebble bed high temperature reactor

    International Nuclear Information System (INIS)

    Verfondern, K.

    1983-01-01

    This work presents a computer model determining the steady-state temperature- and flow field in 3 dimensions in the core of a pebble bed high temperature reactor. The numerical sprinkler method, basind on the Thermix-model, allows to describe the thermo-hydraulics of a non-rotational-symmetric core-geometry. The AVR-reactor in Juelich, in operation since 1967, represents a suitable investigation-object for the computer model of Thermix-3D. It is in a 3D-mesh-structure to reproduce very precisely the so called ''graphite noses'', in which the shut-down rods are conducted as well as the filling cones in the inner and outer area. The results of the final calculation of the normal operation condition for the AVR-reactor unambiguously show, that within the core reproduced in 3 dimensions there are evident deviations in the flow profile and in the temperatures of the cooling gas in contrast to a 2D-handling. (orig.) [de

  2. Conclusions of the experts group of the RA reactor at the meeting held on November 2 and 3 1964 - Annex 12a

    International Nuclear Information System (INIS)

    Pavicevic, M.

    1964-01-01

    Conclusions of the experts group of the RA reactor are related to: analyses of reactor operation at 6.5 MW power with heavy water coolant flow of 250 m 3 /h (2 pumps rotation speed 1500 rotations/min); decisions of future operation; further preparation activities related to reactor operation in forced regime and reduced cooling conditions

  3. Wastewater treatment in a hybrid activated sludge baffled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tizghadam, Mostafa [Laboratoire des Sciences de l' Eau et de l' Environnement, Universite de Limoges, ENSIL, Parc ESTER, 16 Rue Atlantis, F-87068 Limoges Cedex (France); Dagot, Christophe [Laboratoire des Sciences de l' Eau et de l' Environnement, Universite de Limoges, ENSIL, Parc ESTER, 16 Rue Atlantis, F-87068 Limoges Cedex (France)], E-mail: dagot@ensil.unilim.fr; Baudu, Michel [Laboratoire des Sciences de l' Eau et de l' Environnement, Universite de Limoges, ENSIL, Parc ESTER, 16 Rue Atlantis, F-87068 Limoges Cedex (France)

    2008-06-15

    A novel hybrid activated sludge baffled reactor (HASBR), which contained both suspended and attached-growth biomass perfect mixing cells in series, was developed by installing standing and hanging baffles and introducing plastic brushes into a conventional activated sludge (CAS) reactor. It was used for the treatment of domestic wastewater. The effects on the operational performance of developing the suspended and attached-growth biomass and reactor configuration were investigated. The change of the flow regime from complete-mix to plug-flow, and the addition of plastic brushes as a support for biofilm, resulted in considerable improvements in the COD, nitrogen removal efficiency of domestic wastewater and sludge settling properties. In steady state, approximately 98 {+-} 2% of the total COD and 98 {+-} 2% of the ammonia of the influent were removed in the HASBR, when the influent wastewater concentration was 593 {+-} 11 mg COD/L and 43 {+-} 5 mg N/L, respectively, at a HRT of 10 h. These results were 93 {+-} 3 and 6 {+-} 3% for the CAS reactor, respectively. Approximately 90 {+-} 7% of the total COD was removed in the HASBR, when the influent wastewater concentration was 654 {+-} 16 mg COD/L at a 3 h HRT, and in the organic loading rate (OLR) of 5.36 kg COD m{sup -3} day{sup -1}. The result for the CAS reactor was 60 {+-} 3%. Existing CAS plants can be upgraded by changing the reactor configuration and introducing biofilm support media into the aeration tank.

  4. Irradiation of a CANDU UOsub(2) fuel element with twenty-three machined slits cut through the zircaloy sheath

    International Nuclear Information System (INIS)

    DaSilva, R.L.

    1984-09-01

    A CANDU fuel element was purposely defected, exposing a minimum UOsub(2) fuel stack area of 272 mmsup(2), by machining 23 longitudinal slits through the Zircaloy-4 sheathing. The element was then irradiated in the X-2 loop of the NRX reactor for a period of 14.64 effective full power days at a linear heat rating of 48 kW/m to investigate the relationship between fission product release and UOsub(2) oxidation behaviour in an element with minimal fuel-to-gap fission gas trapping. The fission product releases, as measured by on-line gamma-ray spectroscopy, revealed that the noble gases and radioiodines are both released from the UOsub(2) fuel matrix directly to the coolant via simple diffusion kinetics, and that their diffusivities in hyperstoichiometric UOsub(2) are approximately equal. The oxidation of UOsub(2) to the higher states UOsub(2+x), Usub(4)Osub(9) and Usub(3)Osub(8), was accompanied by substantial fuel swelling and sheath deformation preferentially located in the lower powered end of the element. The spalling and erosion behaviour of the fuel pellets was correlated to the rate of fuel oxidation

  5. A 3D heat conduction model for block-type high temperature reactors and its implementation into the code DYN3D

    International Nuclear Information System (INIS)

    Baier, Silvio; Kliem, Soeren; Rohde, Ulrich

    2011-01-01

    The gas-cooled high temperature reactor is a concept to produce energy at high temperatures with a high level of inherent safety. It gets special attraction due to e.g. high thermal efficiency and the possibility of hydrogen production. In addition to the PBMR (Pebble Bed Modular Reactor) the (V)HTR (Very high temperature reactor) concept has been established. The basic design of a prismatic HTR consists of the following elements. The fuel is coated with four layers of isotropic materials. These so-called TRISO particles are dispersed into compacts which are placed in a graphite block matrix. The graphite matrix additionally contains holes for the coolant gas. A one-dimensional model is sufficient to describe (the radial) heat transfer in LWRs. But temperature gradients in a prismatic HTR can occur in axial as well as in radial direction, since regions with different heat source release and with different coolant temperature heat up are coupled through the graphite matrix elements. Furthermore heat transfer into reflector elements is possible. DYN3D is a code system for coupled neutron and thermal hydraulics core calculations developed at the Helmholtzzentrum Dresden-Rossendorf. Concerning neutronics DYN3D consists of a two-group and multi-group diffusion approach based on nodal expansion methods. Furthermore a 1D thermal-hydraulics model for parallel coolant flow channels is included. The DYN3D code was extensively verified and validated via numerous numerical and experimental benchmark problems. That includes the NEA CRP benchmarks for PWR and BWR, the Three-Miles-Island-1 main steam line break and the Peach Bottom Turbine Trip benchmarks, as well as measurements carried out in an original-size VVER-1000 mock-up. An overview of the verification and validation activities can be found. Presently a DYN3D-HTR version is under development. It involves a 3D heat conduction model to deal with higher-(than one)-dimensional effects of heat transfer and heat conduction in

  6. Low-level radioactive river sediment particles originating from the Chalk river nuclear site carry a mixture of radionuclides and metals

    Energy Technology Data Exchange (ETDEWEB)

    Lind, Ole Christian; Cagno, Simone; Salbu, Brit [Norwegian University of Life Sciences - NMBU, Center of Excellence in Environmental Radioactivity - CERAD, P.O. Box 5003, N-1432 Aas (Norway); Falkenberg, Gerald [Photon science, DESY, Hamburg (Germany); Janssens, Koen; Nuyts, Gert; Vanmeert, Frederik [AXIL, Department of Chemistry, University of Antwerpen (Belgium); Jaroszewicz, Jakub [Faculty of Materials Science and Engineering, Warsaw University of Technology, Warsaw (Poland); Priest, Nicholas D.; Audet, Marc [Nuclear Science Division, AECL Chalk River Laboratories (Canada)

    2014-07-01

    The Chalk River Laboratory of Atomic Energy of Canada Ltd., site is located on the Ottawa River approximately 200 km northwest of Ottawa, Canada. The site has two large research reactors: NRX, which operated from 1947 to 1991 and NRU, which continues to operate and is used to produce a significant fraction of the world's supply of medical isotopes. During the course of the operation of the NRX reactor small quantities of radioactive particles were discharged to the Ottawa River through a process sewer discharge pipe. These are now located in river bed sediments within a 0.08 km² area close to the discharge pipe. In the present study, selected particles were isolated from riverbed sediments. These were then characterized by environmental scanning electron microscopy with energy dispersive micro X-ray analysis (ESEM-EDX). This was undertaken to obtain information on particle size, structure and the distribution of elements across particle surfaces. Based on the results of ESEM-EDX, particles were selected for X-ray absorption nano-tomography analysis, which provides videos showing the 3D density distribution of the particles. Furthermore, 2D and 3D Synchrotron Radiation based X-ray techniques (micro-X-ray fluorescence; micro-XRF, micro-X-ray absorption near edge spectroscopy; micro-XANES and micro-X-ray diffraction; micro-XRD) with submicron resolution (beam size 0.5 μm) were employed to investigate the elemental and phase composition (micro-XRF/XRD) and oxidation states (micro-XANES) of matrix elements with high spatial resolution and sensitivity. Results show that the particles investigated so far varied according to: 1) <~40 μm diameter sized U fuel particles similar in structure to particles observed from Chernobyl and Krasnoyarsk-26 and 2) larger particles with diameters up to several hundred μm. The larger particles comprised a matrix of low density, sediment material with high density inclusions that contained a range of metals including Cu, Cr, As

  7. Development of a general coupling interface for the fuel performance code transuranus tested with the reactor dynamic code DYN3D

    International Nuclear Information System (INIS)

    Holt, L.; Rohde, U.; Seidl, M.; Schubert, A.; Van Uffelen, P.

    2013-01-01

    Several institutions plan to couple the fuel performance code TRANSURANUS developed by the European Institute for Transuranium Elements with their own codes. One of these codes is the reactor dynamic code DYN3D maintained by the Helmholtz-Zentrum Dresden - Rossendorf. DYN3D was developed originally for VVER type reactors and was extended later to western type reactors. Usually, the fuel rod behavior is modeled in thermal hydraulics and neutronic codes in a simplified manner. The main idea of this coupling is to describe the fuel rod behavior in the frame of core safety analysis in a more detailed way, e.g. including the influence of the high burn-up structure, geometry changes and fission gas release. It allows to take benefit from the improved computational power and software achieved over the last two decades. The coupling interface was developed in a general way from the beginning. Thence it can be easily used also by other codes for a coupling with TRANSURANUS. The user can choose between a one-way as well as a two-way online coupling option. For a one-way online coupling, DYN3D provides only the time-dependent rod power and thermal hydraulics conditions to TRANSURANUS, but the fuel performance code doesn’t transfer any variable back to DYN3D. In a two-way online coupling, TRANSURANUS in addition transfers parameters like fuel temperature and cladding temperature back to DYN3D. This list of variables can be extended easily by geometric and further variables of interest. First results of the code system DYN3D-TRANSURANUS will be presented for a control rod ejection transient in a modern western type reactor. Pre-analyses show already that a detailed fuel rod behavior modeling will influence the thermal hydraulics and thence also the neutronics due to the Doppler reactivity effect of the fuel temperature. The coupled code system has therefore a potential to improve the assessment of safety criteria. The developed code system DYN3D-TRANSURANUS can be used also

  8. Synthesis of the IRSN report related to severe accidents and to the probabilistic level-2 safety study for the Flamanville EPR reactor. Referral of the Permanent Group of Experts for nuclear reactors (GPR), examination of probabilistic level-2 safety studies (EPS 2) and severe accidents (AG) of the Flamanville reactor nr 3. Opinion related to severe accidents and to the probabilistic level-2 safety study for the Flamanville EPR reactor (FA3). Electronuclear reactors - EDF - Flamanville 3 EPR reactor. Severe accidents and probabilistic level 2 studies

    International Nuclear Information System (INIS)

    2015-01-01

    This document gathers several documents. The first one recalls the main arrangements implemented on the FA3 EPR reactor regarding accidents with core fusion, reports the analysis made by the IRSN about the sizing of these arrangements to reach a controlled status of the installation after a severe accident, regarding the probabilistic level-2 safety assessment, regarding the radiological impact of a severe accident on the population and on the environment, regarding those aimed at facing a total and long duration loss of electric power sources and cold sources, and about the situation of the reactor with respect to WENRA positions on severe accidents for new reactors. The second document is a letter sent by the ASN to the Permanent Group of Experts for nuclear reactors (GPR) to address probabilistic level-2 safety studies (EPS2) and severe accidents for the Flamanville 3 reactor. The third one reports the opinion of the GPR on these both issues and proposes a set of recommendations. The next document is a letter sent by the ASN to the Flamanville 3 project manager at EDF which recalls the related objectives, the ASN opinion on the implemented arrangements for severe accidents (de-pressurization of the primary circuit, management of hydrogen-related risks, corium recovery and cooling outside the vessel, limitation of vapour explosion risks outside the vessel, heat evacuation system, containment enclosure, management of the risk of a return to criticality), to face a total and long duration loss of electricity sources and cold sources, and other aspects addressed in the IRSN analysis. Requests and remarks formulated by the ASN are provided in an appendix to this last document

  9. A friendly Maple module for one and two group reactor model

    International Nuclear Information System (INIS)

    Baptista, Camila O.; Pavan, Guilherme A.; Braga, Kelmo L.; Silva, Marcelo V.; Pereira, P.G.S.; Werner, Rodrigo; Antunes, Valdir; Vellozo, Sergio O.

    2015-01-01

    The well known two energy groups core reactor design model is revisited. A simple and friendly Maple module was built to cover the steps calculations of a plate reactor in five situations: 1. one group bare reactor, 2. two groups bare reactor, 3. one group reflected reactor, 4. 1-1/2 groups reflected reactor and 5. two groups reflected reactor. The results show the convergent path of critical size, as it should be. (author)

  10. A friendly Maple module for one and two group reactor model

    Energy Technology Data Exchange (ETDEWEB)

    Baptista, Camila O.; Pavan, Guilherme A.; Braga, Kelmo L.; Silva, Marcelo V.; Pereira, P.G.S.; Werner, Rodrigo; Antunes, Valdir; Vellozo, Sergio O., E-mail: camila.oliv.baptista@gmail.com, E-mail: pavanguilherme@gmail.com, E-mail: kelmo.lins@gmail.com, E-mail: marcelovilelasilva@gmail.com, E-mail: rodrigowerner@hotmail.com, E-mail: neutron201566@yahoo.com, E-mail: vellozo@ime.eb.br [Instituto Militar de Engenharia (IME), Rio de Janeiro, RJ (Brazil). Departamento de Engenharia Nuclear

    2015-07-01

    The well known two energy groups core reactor design model is revisited. A simple and friendly Maple module was built to cover the steps calculations of a plate reactor in five situations: 1. one group bare reactor, 2. two groups bare reactor, 3. one group reflected reactor, 4. 1-1/2 groups reflected reactor and 5. two groups reflected reactor. The results show the convergent path of critical size, as it should be. (author)

  11. Extension of the GeN-Foam neutronic solver to SP3 analysis and application to the CROCUS experimental reactor

    International Nuclear Information System (INIS)

    Fiorina, Carlo; Hursin, Mathieu; Pautz, Andreas

    2017-01-01

    Highlights: • Development and verification of an SP 3 solver based on OpenFOAM. • Integration into the GeN-Foam multi-physics platform. • Application of the new GeN-Foam SP 3 solver to the CROCUS reactor. - Abstract: The Laboratory for Reactor Physics and Systems Behaviour at the PSI and at the EPFL has been developing since 2013 a multi-physics platform for coupled reactor analysis named GeN-Foam. The developed tool includes a solver for the eigenvalue and transient solution of multi-group neutron diffusion equations. Although frequently used in reactor analysis, the diffusion theory shows some limitations for core configurations involving strong anisotropies, which is the case for the CROCUS research reactor at the EPFL. The use of an SP 3 approximation to neutron transport can often lead to visible improvements in a code predictive capabilities, especially for one-directional anisotropies, with acceptable added computational cost vs diffusion. Following some modelling issues for the CROCUS reactor, and in order to improve the GeN-Foam modelling capabilities, the GeN-Foam diffusion solver has been extended to allow for SP 3 analyses. The present paper describes such extension and a preliminary verification using a mini-core PWR benchmark. The newly developed solver is then applied to the analysis of the CROCUS experimental reactor and results are compared to Monte Carlo calculations, as well as to the results of the diffusion solver.

  12. Thermal fluid dynamics study of nuclear advanced reactors of high temperature using RELAP5-3D

    International Nuclear Information System (INIS)

    Scari, Maria Elizabeth

    2017-01-01

    Fourth Generation nuclear reactors (GEN-IV) are being designed with special features such as intrinsic safety, reduction of isotopic inventory and use of fuel in proliferation-resistant cycles. Therefore, the investigation and evaluation of operational and safety aspects of the GEN-IV reactors have been the subject of numerous studies by the international community and also in Brazil. In 2008, in Brazil, was created the National Institute of Science and Technology of Innovative Nuclear Reactors, focusing on studies of projects and systems of new generation reactors, which included GEN-IV reactors as well as advanced PWR (Pressurized Water Reactor) concepts. The Department of Nuclear Engineering of the Federal University of Minas Gerais (DEN-UFMG) is a partner of this Institute, having started studies on the GEN-IV reactors in the year 2007. Therefore, in order to add knowledge to these studies, in this work, three projects of advanced reactors were considered to verify the simulation capability of the thermo-hydraulic RELAP5-3D code for these systems, either in stationary operation or in transient situations. The addition of new working fluids such as ammonia, carbon dioxide, helium, hydrogen, various types of liquid salts, among them Flibe, lead, lithium-bismuth, lithium-lead, was a major breakthrough in this version of the code, allowing also the simulation of GEN-IV reactors. The modeling of the respective core of an HTTR (High Temperature Engineering Test Reactor), HTR-10 (High Temperature Test Module Reactor) and LS-VHTR (Liquid-Salt-Cooled Very-High-Temperature Reactor) were developed and verified in steady state comparing the values found through the calculations with reference data from other simulations, when it is possible. The first two reactors use helium gas as coolant and the LS-VHTR uses a mixture of 66% LiF and 34% of BeF 2 , the LiF-BeF 2 , also know as Flibe. All the studied reactors use enriched uranium as fuel, in form of TRISO (Tristructural

  13. Study of the effect of slight variants to a 3-loop pressurized water nuclear reactor design in order to improve the reactor safety

    International Nuclear Information System (INIS)

    Castiglia, F.; Oliveri, E.; Taibi, S.; Vella, G.

    1992-01-01

    In order to improve the safety features of a 3-loop pressurized water nuclear reactor we propose a slight design variant consisting in the introduction of a bypass hole in the divider plate of the coolant chambers of the steam generators. The aim is to reduce both the extent and the duration of the core exposure and thus the maximum value of the peak cladding temperature, in case of a hypothetical cold leg small break loss of coolant accident. The proposal, as attested by a preliminary RELAP5/MOD3 analysis, seems to deserve some attention. (6 figures) (Author)

  14. Conceptual analysis of a preliminary model for instability study in normal operation of a natural circulation reactor type EBWR, using Relap5/Mod 3.2

    International Nuclear Information System (INIS)

    Ojeda S, J.; Morales S, J.; Chavez M, C.

    2009-10-01

    This work intends a model using the code Relap5/Mod 3.2, for the instability study in normal operation of a natural circulation reactor type ESBWR. A conceptual analysis is considered because all the information was obtained of the open literature, and some of reactor operation or dimension (not available) parameters were approached. As starting point was took the pattern developed for reactor type BWR, denominated Browns Ferry and changes were focused in elimination of bonds of forced recirculation, in modification of operation parameters, dimensions and own control parameters, according to internal code structure. Additionally the nodalization outline is described analyzing for separate the four fundamental areas employees in peculiar geometry of natural circulation reactor. Comparative analysis of results of stability behavior obtained with those reported in the open literature were made, by part of commercial reactor designer ESBWR. (Author)

  15. Analysis of calculated neutron flux response at detectors of G.A. Siwabessy multipurpose reactor (RSG-GAS Reactor)

    International Nuclear Information System (INIS)

    Taryo, Taswanda

    2002-01-01

    Multi Purpose Reactor G.A. Siwabessy (RSG-GAS) reactor core possesses 4 fission-chamber detectors to measure intermediate power level of RSG-GAS reactor. Another detector, also fission-chamber detector, is intended to measure power level of RSG-GAS reactor. To investigate influence of space to the neutron flux values for each detector measuring intermediate and power levels has been carried out. The calculation was carried out using combination of WIMS/D4 and CITATION-3D code and focused on calculation of neutron flux at different detector location of RSG-GAS typical working core various scenarios. For different scenarios, all calculation results showed that each detector, located at different location in the RSG-GAS reactor core, causes different neutron flux occurred in the reactor core due to spatial time effect

  16. Wastewater treatment in a hybrid activated sludge baffled reactor

    International Nuclear Information System (INIS)

    Tizghadam, Mostafa; Dagot, Christophe; Baudu, Michel

    2008-01-01

    A novel hybrid activated sludge baffled reactor (HASBR), which contained both suspended and attached-growth biomass perfect mixing cells in series, was developed by installing standing and hanging baffles and introducing plastic brushes into a conventional activated sludge (CAS) reactor. It was used for the treatment of domestic wastewater. The effects on the operational performance of developing the suspended and attached-growth biomass and reactor configuration were investigated. The change of the flow regime from complete-mix to plug-flow, and the addition of plastic brushes as a support for biofilm, resulted in considerable improvements in the COD, nitrogen removal efficiency of domestic wastewater and sludge settling properties. In steady state, approximately 98 ± 2% of the total COD and 98 ± 2% of the ammonia of the influent were removed in the HASBR, when the influent wastewater concentration was 593 ± 11 mg COD/L and 43 ± 5 mg N/L, respectively, at a HRT of 10 h. These results were 93 ± 3 and 6 ± 3% for the CAS reactor, respectively. Approximately 90 ± 7% of the total COD was removed in the HASBR, when the influent wastewater concentration was 654 ± 16 mg COD/L at a 3 h HRT, and in the organic loading rate (OLR) of 5.36 kg COD m -3 day -1 . The result for the CAS reactor was 60 ± 3%. Existing CAS plants can be upgraded by changing the reactor configuration and introducing biofilm support media into the aeration tank

  17. Zircaloy behaviour in high temperature irradiated water

    International Nuclear Information System (INIS)

    Urbanic, V.F.

    1982-04-01

    The corrosion and hydriding of Zircaloy during irradiation in high temperature water is strongly dependent on the oxygen concentration of the water. Corrosion tests in the NRX and NRU research reactors using small samples have demonstrated the importance of water chemistry in maintaining Zircaloy corrosion and hydriding within acceptable limits. Zircaloy fuel cladding develops non-uniform, patch-type oxides during irradiation in hich temperature water containing dissolved oxygen. Results from examinations of prototype fuel cladding irradiated in the research reactors are presented to show how local variations in coolant flow, fast neutron flux, metallurgical structure and surface condition can influence the onset of non-uniform corrosion under these conditions. Destructive examinations of CANDU-PHW reactor fuel cladding have emphasized the importance of good chemistry control, especially the dissolved oxygen concentration of the water. When reactor coolants are maintained under normal reducing conditions at high pH (5 to 10 cm 3 D 2 /kg D 2 O; 2 /kg D 2 O; pH > 10 with LiOD), Zircaloy cladding develops non-uniform, patch-type oxides. These patch-type oxides tend to coalesce with time to form a thick, uniform oxide layer after extended exposure. Under reducing coolant conditions, Zircaloy cladding absorbs less than 200 mg D/kg Zr (approximately 2.5 mg/dm 2 equivalent hydrogen) in about 500 days. With oxygen in the coolant, deuterium absorption is considerably less despite the significant increase in corrosion under such conditions

  18. Study on effects of development of reactor constant in fast reactor analysis

    International Nuclear Information System (INIS)

    Chiba, Gou

    2002-12-01

    Evaluation was carried out about an effect of development of the new generation reactor constant system that substitutes for the JFS library in fast reactor analysis. Analyzed cores were ZPPR in JUPITER critical experiment and several power reactor cores that were designed in the feasibility study. In the JUPITER analysis, large effects, over 10%, were observed in sodium void reactivity and sample Doppler reactivity. The former resulted from several factors, while the latter was due to an accurate of a resonance interaction effect between Doppler sample and core fuel. In the previous study, the effect had been evaluated in power reactor cores. The effect included an effect of corrosion of weighting spectrum because JFS-3-J3.2, which had been made with the incorrect weighting spectrum, was used in the evaluation. In the present study, JFS-3-J3.2R, which had been made with the correct weighting spectrum, was used. It was confirmed that the effect of development of reactor constant in power reactor was not as large as that in critical assembly. (author)

  19. RELAP5-3D Code for Supercritical-Pressure Light-Water-Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Riemke, Richard Allan; Davis, Cliff Bybee; Schultz, Richard Raphael

    2003-04-01

    The RELAP5-3D computer program has been improved for analysis of supercritical-pressure, light-water-cooled reactors. Several code modifications were implemented to correct code execution failures. Changes were made to the steam table generation, steam table interpolation, metastable states, interfacial heat transfer coefficients, and transport properties (viscosity and thermal conductivity). The code modifications now allow the code to run slow transients above the critical pressure as well as blowdown transients (modified Edwards pipe and modified existing pressurized water reactor model) that pass near the critical point.

  20. Evaluation of LOCA in a swimming-pool type reactor using the 3D-AIRLOCA code

    International Nuclear Information System (INIS)

    Nagler, A.; Gilat, J.; Hirshfeld, H.

    1991-01-01

    The 3D-AIRLOCA code was used to calculate core temperature evolution curves in the wake of a full LOCA in a swimming pool type reactor, resulting in complete core exposure and dryout within about 1000 sec of the initiating event. The results show that fuel integrity loss thresholds (450 C for softening and 650 C for melting) are reached and exceeded over large fractions of the core at powr levels as low as 2 MW. At 4.5 MW, the softening threshold is reached even when the accident occurs up to 12 hours after reactor shutdown for continuous operation, and up to 2 hrs after shutdown for intermittent (6 hrs/day, 4 days a week) operation. The situation is even more severe in blockage cases, when the air flow through the core is blocked by residual water at the grid plate level. It is concluded that substantial fission product releases are quite likely in this class of accidents. (orig.)

  1. Operating experiences and utilization programmes of the BAEC 3 MW TRIGA Mark-II research reactor of Bangladesh

    International Nuclear Information System (INIS)

    Haque, M.M.; Soner, M.A.M.; Saha, P.K.; Salam, M.A.; Zulquarnain, M.A.

    2008-01-01

    The 3 MW TRIGA Mark-II research reactor of Bangladesh Atomic Energy Commission (BAEC) has been operating since September 14, 1986. The reactor is used for radioisotope production ( 131 I, 99m Tc, 46 Sc), various R and D activities, manpower training and education. The reactor has been operated successfully since its commissioning with the exception of a few reportable incidents. Of these, the decay tank leakage incident of 1997 is considered to be the most significant one. As a result of this incident, reactor operation at full power remained suspended for about 4 years. However, the reactor operation was continued during this period at a power level of 250 kW to cater the needs of various R and D groups, which required lower neutron flux for their experiments. This was made possible by establishing a temporary by pass connection across the decay tank using local technology. The reactor was made operational again at full power after successful replacement of the damaged decay tank in August 2001. At that time, several modifications of the reactor cooling system along with its associated structures were also implemented and then necessary testing and commissioning of the newly installed component/equipment were carried out. The other incident was the contamination of the Dry Central Thimble (DCT) that took place in March 2002 when a pyrex vial containing 50g of TeO 2 powder got melted inside the DCT. The vial was melted due to high heat generation on its surface while the reactor was operated for 8 hours at 3 MW for trial production of Iodine-131 ( 131 I). A Wet Central Thimble (WCT) was used to replace the damaged DCT in June 2002 such that the reactor operation could be resumed. The WCT was again replaced by a new DCT in June 2003 such that radioisotope production could be continued. The facility has so far been used to train up a total of 27 personnel including several foreign nationals to the level of Senior Reactor Operator (SRO) and Reactor Operator (RO). The

  2. Advanced reactors transition fiscal year 1995 multi-year program plan WBS 7.3

    International Nuclear Information System (INIS)

    Loika, E.F.

    1994-01-01

    This document describes in detail the work to be accomplished in FY-1995 and the out years for the Advanced Reactors Transition (WBS 7.3). This document describes specific milestones and funding profiles. Based upon the Fiscal Year 1995 Multi-Year Program Plan, DOE will provide authorization to perform the work outlined in the FY 1995 MYPP. Following direction given by the US Department of Energy (DOE) on December 15, 1993, Advanced Reactors Transition (ART), previously known as Advanced Reactors, will provide the planning and perform the necessary activities for placing the Fast Flux Test Facility (FFTF) in a radiologically and industrially safe shutdown condition. The DOE goal is to accomplish the shutdown in approximately five years. The Advanced Reactors Transition Multi-Year Program Plan, and the supporting documents; i.e., the FFTF Shutdown Program Plan and the FFTF Shutdown Project Resource Loaded Schedule (RLS), are defined for the life of the Program. During the transition period to achieve the Shutdown end-state, the facilities and systems will continue to be maintained in a safe and environmentally sound condition. Additionally, facilities that were associated with the Office of Nuclear Energy (NE) Programs, and are no longer required to support the Liquid Metal Reactor Program will be deactivated and transferred to an alternate sponsor or the Decontamination and Decommissioning (D and D) Program for final disposition, as appropriate

  3. Sources of gamma radiation in a reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Roos, Matts

    1959-05-15

    In a thermal reactor the gamma ray sources of importance for shielding calculations and related aspects are 1) fission, 2) decay of fission products, 3) capture processes in fuel, poison and other materials, 4) inelastic scattering in the fuel and 5) decay of capture products. The energy release and the gamma ray spectra of these sources have been compiled or estimated from the latest information available, and the results are presented in a general way to permit application to any thermal reactor, fueled with a mixture of {sup 235}U and {sup 238}U. As an example the total spectrum and the spectrum of radiation escaping from a fuel rod in the Swedish R3-reactor are presented.

  4. A Study on Dismantling of Westinghouse Type Nuclear Reactor

    International Nuclear Information System (INIS)

    Jeong, Woo-Tae; Lee, Sang-Guk

    2014-01-01

    KHNP started a research project this year to develop a methodology to dismantle nuclear reactors and internals. In this paper, we reviewed 3D design model of the reactor and suggested feasible cutting scheme.. Using 3-D CAD model of Westinghouse type nuclear reactor and its internals, we reviewed possible options for disposal. Among various options of dismantling the nuclear reactor, plasma cutting was selected to be the best feasible and economical method. The upper internals could be segmented by using a band saw. It is relatively fast, and easily maintained. For cutting the lower internals, plasma torch was chosen to be the best efficient tool. Disassembling the baffle and the former plate by removing the baffle former bolts was also recommended for minimizing storage volume. When using plasma torch for cutting the reactor vessel and its internal, installation of a ventilation system for preventing pollution of atmosphere was recommended. For minimizing radiation exposure during the cutting operation, remotely controlled robotic tool was recommended to be used

  5. Nuclear reactor (1960)

    International Nuclear Information System (INIS)

    Maillard, M.L.

    1960-01-01

    The first French plutonium-making reactors G1, G2 and G3 built at Marcoule research center are linked to a power plant. The G1 electrical output does not offset the energy needed for operating this reactor. On the contrary, reactors G2 and G3 will each generate a net power of 25 to 30 MW, which will go into the EDF grid. This power is relatively small, but the information obtained from operation is great and will be helpful for starting up the power reactor EDF1, EDF2 and EDF3. The paper describes how, previous to any starting-up operation, the tests performed, especially those concerned with the power plant and the pressure vessel, have helped to bring the commissioning date closer. (author) [fr

  6. FMDP Reactor Alternative Summary Report: Volume 3 - partially complete LWR alternative

    International Nuclear Information System (INIS)

    Greene, S.R.; Fisher, S.E.; Bevard, B.B.

    1996-09-01

    The Department of Energy Office of Fissile Materials Disposition (DOE/MD) initiated a detailed analysis activity to evaluate each of ten plutonium disposition alternatives that survived an initial screening process. This document, Volume 3 of a four volume report summarizes the results of these analyses for the partially complete LWR (PCLWR) reactor based plutonium disposition alternative

  7. FMDP Reactor Alternative Summary Report: Volume 3 - partially complete LWR alternative

    Energy Technology Data Exchange (ETDEWEB)

    Greene, S.R.; Fisher, S.E.; Bevard, B.B. [and others

    1996-09-01

    The Department of Energy Office of Fissile Materials Disposition (DOE/MD) initiated a detailed analysis activity to evaluate each of ten plutonium disposition alternatives that survived an initial screening process. This document, Volume 3 of a four volume report summarizes the results of these analyses for the partially complete LWR (PCLWR) reactor based plutonium disposition alternative.

  8. The behaviour of zirconium alloys in Santowax OM organic coolant at high temperatures

    International Nuclear Information System (INIS)

    Sawatzky, A.

    1964-10-01

    Zirconium alloys have been exposed to Santowax OM at temperatures of 320 to 400 o C for times as long as 5000 hours. Short-term experiments (less than 2 weeks) were done in stainless-steel bombs and small out-of-pile loops. The X-7 organic loop in the NRX reactor was used to study long-term oxidation and hydriding both in-flux and out-of-flux. The results obtained lead to several tentative conclusions: Aluminum cladding serves as an effective hydrogen barrier; Considerable protection against hydriding is given by zirconium oxide, provided impurities in the organic are carefully controlled; Hydriding is greatly enhanced by the presence of chlorine in the coolant; and, Hydriding is somewhat enhanced by neutron irradiation. Of considerable significance is the fact that a Zircaloy-4 in-reactor test section of the X-7 loop was exposed to Santowax OM at 320 to 400 o C for more than 5000 hours without excessive hydriding. (author)

  9. Anaerobic biogranulation in a hybrid reactor treating phenolic waste

    International Nuclear Information System (INIS)

    Ramakrishnan, Anushyaa; Gupta, S.K.

    2006-01-01

    Granulation was examined in four similar anaerobic hybrid reactors 15.5 L volume (with an effective volume of 13.5 L) during the treatment of synthetic coal wastewater at the mesophilic temperature of 27 ± 5 deg. C. The hybrid reactors are a combination of UASB unit at the lower part and an anaerobic filter at the upper end. Synthetic wastewater with an average chemical oxygen demand (COD) of 2240 mg/L, phenolics concentration of 752 mg/L and a mixture of volatile fatty acids was fed to three hybrid reactors. The fourth reactor, control system, was fed with a wastewater containing sodium acetate and mineral nutrients. Coal waste water contained phenol (490 mg/L); m-, o-, p-cresols (123.0, 58.6, 42 mg/L); 2,4-, 2,5-, 3,4- and 3,5-dimethyl phenols (6.3, 6.3, 4.4 and 21.3 mg/L) as major phenolic compounds. A mixture of anaerobic digester sludge and partially granulated sludge (3:1) were used as seed materials for the start up of the reactors. Granules were observed after 45 days of operation of the systems. The granules ranged from 0.4 to 1.2 mm in diameter with good settling characteristics with an SVI of 12 mL/g SS. After granulation, the hybrid reactor performed steadily with phenolics and COD removal efficiencies of 93% and 88%, respectively at volumetric loading rate of 2.24 g COD/L d and hydraulic retention time of 24 h. The removal efficiencies for phenol and m/p-cresols reached 92% and 93% (corresponding to 450.8 and 153 mg/L), while o-cresol was degraded to 88% (corresponding to 51.04 mg/L). Dimethyl phenols could be removed completely at all the organic loadings and did not contribute much to the residual organics. Biodegradation of o-cresol was obtained in the hybrid-UASB reactors

  10. Development of a general coupling interface for the fuel performance code TRANSURANUS – Tested with the reactor dynamics code DYN3D

    International Nuclear Information System (INIS)

    Holt, L.; Rohde, U.; Seidl, M.; Schubert, A.; Van Uffelen, P.; Macián-Juan, R.

    2015-01-01

    Highlights: • A general coupling interface was developed for couplings of the TRANSURANUS code. • With this new tool simplified fuel behavior models in codes can be replaced. • Applicable e.g. for several reactor types and from normal operation up to DBA. • The general coupling interface was applied to the reactor dynamics code DYN3D. • The new coupled code system DYN3D–TRANSURANUS was successfully tested for RIA. - Abstract: A general interface is presented for coupling the TRANSURANUS fuel performance code with thermal hydraulics system, sub-channel thermal hydraulics, computational fluid dynamics (CFD) or reactor dynamics codes. As first application the reactor dynamics code DYN3D was coupled at assembly level in order to describe the fuel behavior in more detail. In the coupling, DYN3D provides process time, time-dependent rod power and thermal hydraulics conditions to TRANSURANUS, which in case of the two-way coupling approach transfers parameters like fuel temperature and cladding temperature back to DYN3D. Results of the coupled code system are presented for the reactivity transient scenario, initiated by control rod ejection. More precisely, the two-way coupling approach systematically calculates higher maximum values for the node fuel enthalpy. These differences can be explained thanks to the greater detail in fuel behavior modeling. The numerical performance for DYN3D–TRANSURANUS was proved to be fast and stable. The coupled code system can therefore improve the assessment of safety criteria, at a reasonable computational cost

  11. Method of operating a reactor

    International Nuclear Information System (INIS)

    Oosumi, Katsumi; Yamamoto, Michiyoshi.

    1980-01-01

    Purpose: To prevent stress corrosion cracking in the structural material of a reactor pressure vessel. Method: Prior to the starting of a reactor, the reactor pressure vessel is evacuated to carry out degassing of reactor water, and, at the same time, reactor water is heated. After reactor water is heated to a predetermined temperature, control rods are extracted to start nuclear heating. While the temperature of the reactor water is in a temperature range where elution of a metal which is a structural material of the reactor pressure vessel becomes vigorous and the sensitivity to the stress corrosion cracks increases, the reactor is operated at the maximum permissible temperature raising speed or maximum permissible cooling speed. (Aizawa, K.)

  12. Characteristics of UV-MicroO3 Reactor and Its Application to Microcystins Degradation during Surface Water Treatment

    Directory of Open Access Journals (Sweden)

    Guangcan Zhu

    2015-01-01

    Full Text Available The UV-ozone (UV-O3 process is not widely applied in wastewater and potable water treatment partly for the relatively high cost since complicated UV radiation and ozone generating systems are utilized. The UV-microozone (UV-microO3, a new advanced process that can solve the abovementioned problems, was introduced in this study. The effects of air flux, air pressure, and air humidity on generation and concentration of O3 in UV-microO3 reactor were investigated. The utilization of this UV-microO3 reactor in microcystins (MCs degradation was also carried out. Experimental results indicated that the optimum air flux in the reactor equipped with 37 mm diameter quartz tube was determined to be 18∼25 L/h for efficient O3 generation. The air pressure and humidity in UV-microO3 reactor should be low enough in order to get optimum O3 output. Moreover, microcystin-RR, YR, and LR (MC-RR, MC-YR, and MC-LR could be degraded effectively by UV-microO3 process. The degradation of different MCs was characterized by first-order reaction kinetics. The pseudofirst-order kinetic constants for MC-RR, MC-YR, and MC-LR degradation were 0.0093, 0.0215, and 0.0286 min−1, respectively. Glucose had no influence on MC degradation through UV-microO3. The UV-microO3 process is hence recommended as a suitable advanced treatment method for dissolved MCs degradation.

  13. Reactor Physics Programme

    Energy Technology Data Exchange (ETDEWEB)

    De Raedt, C

    2000-07-01

    The Reactor Physics and Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis on reactor fuel. This expertise is applied within the Reactor Physics and MYRRHA Research Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments. Progress and achievements in 1999 in the following areas are reported on: (1) investigations on the use of military plutonium in commercial power reactors; (2) neutron and gamma calculations performed for BR-2 and for other reactors; (3) the updating of neutron and gamma cross-section libraries; (4) the implementation of reactor codes; (6) the management of the UNIX workstations; and (6) fuel cycle studies.

  14. Reactor Physics Programme

    International Nuclear Information System (INIS)

    De Raedt, C.

    2000-01-01

    The Reactor Physics and Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis on reactor fuel. This expertise is applied within the Reactor Physics and MYRRHA Research Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments. Progress and achievements in 1999 in the following areas are reported on: (1) investigations on the use of military plutonium in commercial power reactors; (2) neutron and gamma calculations performed for BR-2 and for other reactors; (3) the updating of neutron and gamma cross-section libraries; (4) the implementation of reactor codes; (6) the management of the UNIX workstations; and (6) fuel cycle studies

  15. Modelling and thermal hydraulic analysis of the Angra-2 nuclear reactor using RELAP5-3D code

    International Nuclear Information System (INIS)

    González Mantecón, Javier

    2015-01-01

    The evaluation of Nuclear Power Plants (NPPs) performance during steady-state and accident conditions has been one of the main research subjects in the nuclear field. In order to simulate the behavior of water-cooled reactors, several complex thermal-hydraulic codes systems have been developed. Particularly, the RELAP5 code, developed by the Idaho National Laboratory, is a best-estimate thermal-hydraulic analysis tool and one of the most used in nuclear industry. The RELAP5-3D 3.0.0 code was used to develop a detailed model of Angra 2 nuclear reactor using reference data from the Final Safety Analysis Report. Angra 2 is the second Brazilian NPP, which began commercial operation in 2001. The plant is equipped with a Pressurized Water Reactor (PWR) type with 3771.0 MWt. Simulations of the reactor behavior during normal operation conditions and postulated accident conditions were performed. Results achieved in the reactor steady-state simulation were compared with nominal parameters of the NPP. These results proved to be in good agreement, with relative errors less than 1%. In the transient simulation, the obtained results were coherent and satisfactory. This study demonstrates that the RELAP5-3D model is capable to reproduce the thermal-hydraulic behavior of the Angra-2 PWR during diverse operation conditions and it can contribute for the process of the plant safety analysis. (author)

  16. Examination of a failed reactor coolant pump rotating assembly from Crystal River Unit 3

    International Nuclear Information System (INIS)

    Hayner, G.O.; Lubnow, T.; Clary, M.

    1990-01-01

    On January 18, 1989, the A reactor coolant pump rotating assembly at the Crystal River Unit 3 Nuclear Power Plant failed during operation. A rotating assembly from this pump had previously failed in 1986. The reactor coolant pump was fabricated by Byron Jackson Pump Division of Borg-Warner Ind. Products, Inc. from UNS S66286 superalloy (Alloy A286). A root cause failure analysis examination was performed on the pump shaft and other components. The failure analysis included shaft vibrational mode and stress analyses, pump clearance and alignment analyses, and detailed destructive examination of the shaft and hydrostatic bearing assemblies. Based on the detailed physical examination of the shaft it was concluded that cracks initiated in the pump shaft at two sites approximately 180 0 apart in a band of shallow, thermally induced fatigue cracks. The cracks initiated at the bottom edge of the motor end shrink fit pad under the shrink fit sleeve supporting the hydrostatic bearing journal. The band of thermally induced fatigue cracks was apparently caused by mixing of cold seal injection water and hot reactor coolant in gaps between the pump shaft and sleeve. The motor end shrink fit was apparently not effective in preventing introduction of the seal injection water to this area. Initial crack propagation occurred by fatigue due to lateral vibration; however, the majority of crack propagation occurred by abnormal torsional fatigue loading induced by contact and sticking between the rotating and stationary portions of the hydrostatic bearing. Final fracture of the shaft occurred by torsional overload. Metallurgical characteristics and mechanical properties of the shaft were within design specification and probably did not significantly influence the cracking process

  17. Analysis of SBO accident for a swimming pool reactor

    International Nuclear Information System (INIS)

    Wang Guimin; Li Weiwei; Li Ning; Guo Wenhui

    2015-01-01

    The RELAP5/MOD3.3 code was adopted to compute the SBO accident condition of a swimming pool reactor. The coolant flow reversal process was calculated, and the influence of parameters of the flow between the core leakage and components on the flow reversal in the SBO accident condition was analyzed. The calculated results show that in the situation the reactor loses all forced flow, the residual heat of the reactor can be removed by the natural circulation flow, and the fuel subassembly will not be damaged. (authors)

  18. Collective occupational dose for nuclear reactors of the 2., 3. and 4. generation

    International Nuclear Information System (INIS)

    Guidez, J.; Saturnin, A.

    2016-01-01

    In France during reactor operation the individual occupational doses are collected and recorded according to the law. When you sum up all the individual doses you get the yearly collective dose expressed in Man.Sv/year. This piece of information can be used to make comparisons between various types of reactors and between reactors of the same type. The results show a steady decrease of the collective dose for all types of reactors over the time except for CANDU reactors for which a slight increase of the dose has appeared since the years 1996-1998. The decrease is due to the continuous improvement of reactor operating and to changes in the reactor design. There is also a constant gap over time between the collective dose for a BWR reactor (1.12 Man.Sv/y) and a PWR reactor 0.60 Man.Sv/y), this gap is certainly due to N 16 nuclide that is created in the primary circuit and transported to turbines in the case of a BWR reactor. For sodium-cooled fast reactors (RNR-Na) the collective dose is below 0.40 Man.Sv/y except for the BN-600 reactor. (A.C.)

  19. Probabilistic safety criteria on high burnup HWR fuels

    International Nuclear Information System (INIS)

    Marino, A.C.

    2002-01-01

    BACO is a code for the simulation of the thermo-mechanical and fission gas behaviour of a cylindrical fuel rod under operation conditions. Their input parameters and, therefore, output ones may include statistical dispersion. In this paper, experimental CANDU fuel rods irradiated at the NRX reactor together with experimental MOX fuel rods and the IAEA-CRP FUMEX cases are used in order to determine the sensitivity of BACO code predictions. The techniques for sensitivity analysis defined in BACO are: the 'extreme case analysis', the 'parametric analysis' and the 'probabilistic (or statistics) analysis'. We analyse the CARA and CAREM fuel rods relation between predicted performance and statistical dispersion in order of enhanced their original designs taking account probabilistic safety criteria and using the BACO's sensitivity analysis. (author)

  20. Neutron flux calculations for the Rossendorf research reactor in (hex)- and (hex,z)-geometry using SNAP-3D

    International Nuclear Information System (INIS)

    Koch, R.; Findeisen, A.

    1986-04-01

    The multigroup neutron diffusion theory code SNAP-3D has been used to perform time independent neutron flux and power calculations of the 10 MW Rossendorf research reactor of the type WWR-SM. The report describes these calculations, as well as the actual reactor configuration, some details of the code SNAP-3D, and two- and three-dimensional reactor models. For evaluating the calculations some flux values and control rod worths have been compared with those of measurements. (author)

  1. Neurexin Dysfunction in Adult Neurons Results in Autistic-like Behavior in Mice

    Directory of Open Access Journals (Sweden)

    Luis G. Rabaneda

    2014-07-01

    Full Text Available Autism spectrum disorders (ASDs comprise a group of clinical phenotypes characterized by repetitive behavior and social and communication deficits. Autism is generally viewed as a neurodevelopmental disorder where insults during embryonic or early postnatal periods result in aberrant wiring and function of neuronal circuits. Neurexins are synaptic proteins associated with autism. Here, we generated transgenic βNrx1ΔC mice in which neurexin function is selectively impaired during late postnatal stages. Whole-cell recordings in cortical neurons show an impairment of glutamatergic synaptic transmission in the βNrx1ΔC mice. Importantly, mutant mice exhibit autism-related symptoms, such as increased self-grooming, deficits in social interactions, and altered interaction for nonsocial olfactory cues. The autistic-like phenotype of βNrx1ΔC mice can be reversed after removing the mutant protein in aged animals. The defects resulting from disruption of neurexin function after the completion of embryonic and early postnatal development suggest that functional impairment of mature circuits can trigger autism-related phenotypes.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  3. Complete Flow Blockage of a Fuel Channel for Research Reactor

    International Nuclear Information System (INIS)

    Lee, Byeonghee; Park, Suki

    2015-01-01

    The CHF correlation suitable for narrow rectangular channels are implemented in RELAP5/MOD3.3 code for the analyses, and the behavior of fuel temperatures and MCHFR(minimum critical heat flux ratio) are compared between the original and modified codes. The complete flow blockage of fuel channel for research reactor is analyzed using original and modified RELAP5/MOD3.3 and the results are compared each other. The Sudo-Kaminaga CHF correlation is implemented into RELAP5/MOD3.3 for analyzing the behavior of fuel adjacent to the blocked channel. A flow blockage of fuel channels can be postulated by a foreign object blocking cooling channels of fuels. Since a research reactor with plate type fuel has isolated fuel channels, a complete flow blockage of one fuel channel can cause a failure of adjacent fuel plates by the loss of cooling capability. Although research reactor systems are designed to prevent foreign materials from entering into the core, partial flow blockage accidents and following fuel failures are reported in some old research reactors. In this report, an analysis of complete flow blockage accident is presented for a 15MW pool-type research reactor with plate type fuels. The fuel surface experience different heat transfer regime in the results from original and modified RELAP5/MOD3.3. By the discrepancy in heat transfer mode of two cases, a fuel melting is expected by the modified RELAP5/MOD3.3, whereas the fuel integrity is ensured by the original code

  4. Neutronics and mass transport in a chemical reactor associated with controlled thermonuclear fusion reactor

    International Nuclear Information System (INIS)

    Dang, V.D.; Steinberg, M.; Lazareth, O.W.; Powell, J.R.

    1976-05-01

    The formation of ozone from oxygen and the dissociation carbon dioxide to carbon monoxide and oxygen is studied in a gamma-neutron chemical process blanket associated with a controlled thermonuclear reactor. Materials used for reactor tube wall will affect the efficiency of the energy absorption by the reactants and consequently the yield of reaction products. Three kinds of materials, aluminum, stainless steel and fiber (Al 2 O 3 )-aluminium are investigated for the tube wall material in the study

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  6. A study of reactor vessel integrity assessment

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Hoon [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of); Kim, Jong Kyung; Shin, Chang Ho; Seo, Bo Kyun [Hanyang Univ., Seoul (Korea, Republic of)

    1999-02-15

    The fast neutron fluence at the Reactor Pressure Vessel(RPV) of KNGR designed for 60 years lifetime was calculated by full-scope Monte Carlo simulation for reactor vessel integrity assessment. KNGR core geometry was modeled on a three-dimensional representation of the one-sixteenth of the reactor in-vessel component. Each fuel assemblies were modeled explicitly, and each fuel pins were axially divided into 5 segments. The maximum flux of 4.3 x 10{sup 10} neutrons/cm{sup 2}. sec at the RPV was obtained by tallying neutrons crossing the beltline of inner surface of the RPV.

  7. Reactor safety study applied to the Forsmark 3 Power Plant

    International Nuclear Information System (INIS)

    Ericsson, G.; Tiren, L.I.

    1978-01-01

    A reactor safety study of the Forsmark 3 BWR power plant has been carried out for the purpose of calculating core melt probabilities using WASH-1400 methods. A sensitivity analysis shows that the calculated core melt probability is changed by approximately a factor of 10 depending on assumptions made with respect to the probability of human error. The importance of the availability of off-site power and the influence of common cause failure is also discussed. (author)

  8. The conversion of NRU from HEU to LEU fuel

    International Nuclear Information System (INIS)

    Sears, D.F.; Atfield, M.D.; Kennedy, I.C.

    1990-01-01

    The program at Chalk River Nuclear Laboratories (CRNL) to develop and test low-enriched uranium fuel (LEU, 3 Si, USiAl, USi Al and U 3 Si 2 (U-3.96 wt% Si; U-3.5 wt% Si-1.5 wt% AL; U-3.2 wt%; Si-3 wt% Al; U-7.3 wt% Si, respectively). Fuel elements were fabricated with uranium loadings suitable for NRU, 3.15 gU/cm 3 , and for NRX, 4.5 gU/cm 3 , and were irradiated under normal fuel-operating conditions. Eight experimental irradiations involving 100 mini-elements and 84 full-length elements (7X12-element rods) were completed to qualify the LEU fuel and the fabrication technology. Post irradiation examinations confirmed that the performance of the LEU fuel, and that of a medium enrichment uranium (MEU, 45% U-235) alloy fuel tested as a back-up, was comparable to the HEU fuel. The uranium silicide dispersion fuel swelling was approximately linear up to burnups exceeding NRU's design terminal burnup (80 at%). NRU was partially converted to LEU fuel when the first 31 prototype fuel rods manufactured with industrial scale production equipment were installed in the reactor. The rods were loaded in NRU at a fuelling rate of about two rods per week over the period 1988 September to December. This partial LEU core (one third of a full NRU core) has allowed the reactor engineers and physicists to evaluate the bulk effects of the LEU conversion on NRU operations. As expected, the irradiation is proceeding without incident

  9. PRE design of a molten salt thorium reactor loop

    International Nuclear Information System (INIS)

    Caire, Jean-Pierre; Roure, Anthony

    2007-01-01

    This study is a contribution to the 2004 PCR-RSF program of the Centre National de la Recherche Scientifique (CNRS) devoted to research on high temperature thorium molten salt reactors. A major issue of high temperature molten salt reactors is the very large heat duty to be transferred from primary to secondary loop of the reactor with minimal thermal losses. A possible inner loop made of a series of conventional graphite filter plate exchangers, pipes and pumps was investigated. The loop was assumed to use two counter current flows of the same LiF, BeF 2 , ZrF 4 , UF 4 molten salt flowing through the reactor. The 3D model used the coupling of k-ε turbulent Navier-Stokes equations and thermal applications of the Heat Transfer module of COMSOL Multiphysics. For a reactor delivering 2700 MWth, the model required a set of 114 identical exchangers. Each one was optimized to limit the heat losses to 2882 W. The pipes made of a succession of graphite, ceramics, Hastelloy-N alloy and insulating Microtherm layers led to a thermal loss limited to 550 W per linear meter. In such conditions, the global thermal losses represent only 0.013% of the reactor thermal power for elements covered with an insulator only 3 cm thick. (author)

  10. Corrosion and hydriding behaviour of some Zr 2.5 wt% Nb alloys in water, steam and various gases at high temperature

    Energy Technology Data Exchange (ETDEWEB)

    Dalgaard, S. B.

    1962-05-15

    Fuel sheaths and pressure tubes in Canadian power reactors are at present made from Zircaloy-2. Mechanical properties of a suitably heat treated Zr 2.5 wt% Nb alloy are superior to those of Zircaloy-2, but any new alloy must have resistance to corrosion and hydriding by the coolant and by the gas that insulates the pressure tube from the cold moderator. Exposed to water at temperatures up to 325{sup o}C, the Zr 2.5 wt% Nb alloy has corrosion resistance acceptable for power reactors. Resistance to air and carbon dioxide is less favourable. Addition of tin, or iron and chromium, to the base alloy have little effect on the corrosion resistance, but the addition of copper reduces corrosion in water and steam to some extent and in air and carbon dioxide to a greater extent. Studies of the effect of heat treatment suggest that the amount of niobium in a solid-solution controls the rate of oxidation and hydriding and that concentration, size and distribution of second phase is of little importance. Initial results obtained in NRX indicate that a thermal flux of 3-7 x 10{sup 13} n/cm{sup 2}/sec has little or no effect on oxidation and hydriding in high temperature water. (author)

  11. Mathematical Modelling of Catalytic Fixed-Bed Reactor for Carbon Dioxide Reforming of Methane over Rh/Al2O3 Catalyst

    Directory of Open Access Journals (Sweden)

    New Pei Yee

    2008-04-01

    Full Text Available A one-dimensional mathematical model was developed to simulate the performance of catalytic fixed bedreactor for carbon dioxide reforming of methane over Rh/Al2O3 catalyst at atmospheric pressure. The reactionsinvolved in the system are carbon dioxide reforming of methane (CORM and reverse water gas shiftreaction (RWGS. The profiles of CH4 and CO2 conversions, CO and H2 yields, molar flow rate and molefraction of all species as well as reactor temperature along the axial bed of catalyst were simulated. In addition,the effects of different reactor temperature on the reactor performance were also studied. The modelscan also be applied to analyze the performances of lab-scale micro reactor as well as pilot-plant scale reactorwith certain modifications and model verification with experimental data. © 2008 BCREC UNDIP. All rights reserved.[Received: 20 August 2008; Accepted: 25 September 2008][How to Cite: N.A.S. Amin, I. Istadi, N.P. Yee. (2008. Mathematical Modelling of Catalytic Fixed-Bed Reactor for Carbon Dioxide Reforming of Methane over Rh/Al2O3 Catalyst. Bulletin of Chemical Reaction Engineering and Catalysis, 3 (1-3: 21-29. doi:10.9767/bcrec.3.1-3.19.21-29

  12. Reactor noise analysis of experimental fast reactor 'JOYO'

    International Nuclear Information System (INIS)

    Ohtani, Hideji; Yamamoto, Hisashi

    1980-01-01

    As a part of dynamics tests in experimental fast reactor ''JOYO'', reactor noise tests were carried out. The reactor noise analysis techniques are effective for study of plant characteristics by determining fluctuations of process signals (neutron signal, reactor inlet temperature signals, etc.), which are able to be measured without disturbances for reactor operations. The aims of reactor noise tests were to confirm that no unstable phenomenon exists in ''JOYO'' and to gain initial data of the plant for reference of the future data. Data for the reactor noise tests treated in this paper were obtained at 50 MW power level. Fluctuations of process signals were amplified and recorded on analogue tapes. The analysis was performed using noise code (NOISA) of digital computer, with which statistical values of ASPD (auto power spectral density), CPSD (cross power spectral density), and CF (coherence function) were calculated. The primary points of the results are as follows. 1. RMS value of neutron signal at 50 MW power level is about 0.03 MW. This neutron fluctuation is not disturbing reactor operations. 2. The fluctuations of A loop reactor inlet temperatures (T sub(AI)) are larger than the fluctuations of B loop reactor inlet temperature (T sub(BI)). For this reason, the major driving force of neutron fluctuations seems to be the fluctuations of T sub(AI). 3. Core and blanket subassemblies can be divided into two halves (A and B region), with respect to the spacial motion of temperature in the reactor core. A or B region means the region in which sodium temperature fluctuations in subassembly are significantly affected by T sub(AI) or T sub(BI), respectively. This phenomenon seems to be due to the lack of mixing of A and B loop sodium in lower plenum of reactor vessel. (author)

  13. Development of a simultaneous partial nitrification and anaerobic ammonia oxidation process in a single reactor.

    Science.gov (United States)

    Cho, Sunja; Fujii, Naoki; Lee, Taeho; Okabe, Satoshi

    2011-01-01

    Up-flow oxygen-controlled biofilm reactors equipped with a non-woven fabric support were used as a single reactor system for autotrophic nitrogen removal based on a combined partial nitrification and anaerobic ammonium oxidation (anammox) reaction. The up-flow biofilm reactors were initiated as either a partial nitrifying reactor or an anammox reactor, respectively, and simultaneous partial nitrification and anammox was established by careful control of the aeration rate. The combined partial nitrification and anammox reaction was successfully developed in both biofilm reactors without additional biomass inoculation. The reactor initiated as the anammox reactor gave a slightly higher and more stable mean nitrogen removal rate of 0.35 (±0.19) kg-N m(-3) d(-1) than the reactor initiated as the partial nitrifying reactor (0.23 (±0.16) kg-N m(-3) d(-1)). FISH analysis revealed that the biofilm in the reactor started as the anammox reactor were composed of anammox bacteria located in inner anoxic layers that were surrounded by surface aerobic AOB layers, whereas AOB and anammox bacteria were mixed without a distinguishable niche in the biofilm in the reactor started as the partial nitrifying reactor. However, it was difficult to efficiently maintain the stable partial nitrification owing to inefficient aeration in the reactor, which is a key to development of the combined partial nitrification and anammox reaction in a single biofilm reactor. Copyright © 2010 Elsevier Ltd. All rights reserved.

  14. Preliminary design of a Binary Breeder Reactor

    International Nuclear Information System (INIS)

    Garcia C, E. Y.; Francois, J. L.; Lopez S, R. C.

    2014-10-01

    A binary breeder reactor (BBR) is a reactor that by means of the transmutation and fission process can operates through the depleted uranium burning with a small quantity of fissile material. The advantages of a BBR with relation to other nuclear reactor types are numerous, taking into account their capacity to operate for a long time without requiring fuel reload or re-arrangement. In this work four different simulations are shown carried out with the MCNPX code with libraries Jeff-3.1 to 1200 K. The objective of this study is to compare two different models of BBR: a spherical reactor and a cylindrical one, using two fuel cycles for each one of them (U-Pu and Th-U) and different reflectors for the two different geometries. For all the models a super-criticality state was obtained at least 10.9 years without carrying out some fuel re-arrangement or reload. The plutonium-239 production was achieved in the models where natural uranium was used in the breeding area, while the production of uranium-233 was observed in the cases where thorium was used in the fertile area. Finally, a behavior of stationary wave reactor was observed inside the models of spherical reactor when contemplating the power uniform increment in the breeding area, while inside the cylindrical models was observed the behavior of a traveling wave reactor when registering the displacement of the burnt wave along the cylindrical model. (Author)

  15. Mechanical behavior of a fast reactor core: Application of the 3D codes to SUPER PHENIX 1

    International Nuclear Information System (INIS)

    Bernard, A.; Masoni, P.; Dorsselaere, J.P. van

    1983-01-01

    The series of the 3-dimensional mechanical codes of a fast reactor core was used for the first time within the framework of a design study of an industrial reactor: SUPER-PHENIX 1. These codes are the following ones: - ARGOH which calculates the behavior of an isolated subassembly. - HARMONIE which calculates the core mechanical equilibrium - TRACAR which yields a graphic visualization of HARMONIE results, and calculates the handling forces and support reactions - HARMOREA which calculates the reactivity variations between given equilibrium states (for instance: pads effect and diagrid effect); now at the end of its development. The calculations were performed on 1/3 of the SPX1 core. Their purpose is double: - on the one hand, to check design criteria, and provide the loadings for the subassembly mechanical design studies; on the other hand, to evaluate the reactivity effects, related to the horizontal core deformations, and useful for operation and safety studies. The results of these calculations showed that the design criteria were verified for the contractual lifetime of the subassemblies. (orig.)

  16. 2D and 3D CFD modelling of a reactive turbulent flow in a double shell supercritical water oxidation reactor

    International Nuclear Information System (INIS)

    Moussiere, S.; Roubaud, A.; Fournel, B.; Joussot-Dubien, C.; Boutin, O.; Guichardon, P.

    2012-01-01

    In order to design and define appropriate dimensions for a supercritical oxidation reactor, a comparative 2D and 3D simulation of the fluid dynamics and heat transfer during an oxidation process has been performed. The solver used is a commercial code, Fluent 6.2 (R). The turbulent flow field in the reactor, created by the stirrer, is taken into account with a k-omega model and a swirl imposed to the fluid. In the 3D case the rotation of the stirrer can be modelled using the sliding mesh model and the moving reference frame model. This work allows comparing 2D and 3D velocity and heat transfer calculations. The predicted values (mainly species concentrations and temperature profiles) are of the same order in both cases. The reactivity of the system is taken into account with a classical Eddy Dissipation Concept combustion model. Comparisons with experimental temperature measurements validate the ability of the CFD modelling to simulate the supercritical water oxidation reactive medium. Results indicate that the flow can be considered as plug flow-like and that heat transfer is strongly enhanced by the stirring. (authors)

  17. 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

  18. The research reactor as a tool in the master in nuclear reactors in Argentina

    International Nuclear Information System (INIS)

    Notari, Carla

    2003-01-01

    complete the Master with a seminar: Nuclear Power Plants, and a Thesis. In the frame of the academic plan, multiple activities are organized related to research reactors and also to nuclear power plants. Since the very beginning the performance of selected experiments in a nuclear reactor was recognized as an extraordinary tool to give the students an insight in the principal phenomena associated with the chain reaction and the related engineering problems. This experiments have an intrinsic elevated cost, associated with the relevance of the installation and with the specialized personnel involved. CNEA provides the career with this educational instrument through the Ra-1 and RA-3 reactors located at Constituyentes and Ezeiza Atomic Centers respectively. Various activities are under way but the most established, in the Reactor Physics Course, is the estimation of kinetic parameters in RA-1 reactor. The practice includes three different experiments: Approach to critical and calibration of control rods by the compensation method: Starting in a subcritical state with source the calibration of control rod B1 vs B2 is done by introduction of the first and withdrawal of the second. The methods used are based on the Point Kinetic Model; Measurement of control rods effectivity by the rod-drop method: Separate Rod Drop of rods B1 B2 B3 of the overall ensemble B1 B2 B3 B4 and total scram starting with three withdrawn and one partially inserted, is the procedure followed to estimate the reactivity worth of B1 B2 B3 and scram. The Point Kinetic Model and the Modal Kinetic Model are used; Reactor noise technique for the estimation of reactor parameters: α and Λ. The kinetic parameters are estimated assuring that the Point Kinetic Model is valid (detection chambers near to the core), that the fluctuation of the fission density is the dominant source of the correlated part of neutron noise (measurement at low power, <10kw), the dominance of the fundamental armonic (simultaneous use of

  19. Effect of increasing nitrobenzene loading rates on the performance of anaerobic migrating blanket reactor and sequential anaerobic migrating blanket reactor/completely stirred tank reactor system

    International Nuclear Information System (INIS)

    Kuscu, Ozlem Selcuk; Sponza, Delia Teresa

    2009-01-01

    A laboratory scale anaerobic migrating blanket reactor (AMBR) reactor was operated at nitrobenzene (NB) loading rates increasing from 3.33 to 66.67 g NB/m 3 day and at a constant hydraulic retention time (HRT) of 6 days to observe the effects of increasing NB concentrations on chemical oxygen demand (COD), NB removal efficiencies, bicarbonate alkalinity, volatile fatty acid (VFA) accumulation and methane gas percentage. Moreover, the effect of an aerobic completely stirred tank reactor (CSTR) reactor, following the anaerobic reactor, on treatment efficiencies was also investigated. Approximately 91-94% COD removal efficiencies were observed up to a NB loading rate of 30.00 g/m 3 day in the AMBR reactor. The COD removal efficiencies decreased from 91% to 85% at a NB loading rate of 66.67 g/m 3 day. NB removal efficiencies were approximately 100% at all NB loading rates. The maximum total gas, methane gas productions and methane percentage were found to be 4.1, 2.6 l/day and 59%, respectively, at a NB loading rate of 30.00 g/m 3 day. The optimum pH values were found to be between 7.2 and 8.4 for maximum methanogenesis. The total volatile fatty acid (TVFA) concentrations in the effluent were 110 and 70 mg/l in the first and second compartments at NB loading rates as high as 66.67 and 6.67 g/m 3 day, respectively, while they were measured as zero in the effluent of the AMBR reactor. In this study, from 180 mg/l NB 66 mg/l aniline was produced in the anaerobic reactor while aniline was completely removed and transformed to 2 mg/l of cathechol in the aerobic CSTR reactor. Overall COD removal efficiencies were found to be 95% and 99% for NB loading rates of 3.33 and 66.67 g/m 3 day in the sequential anaerobic AMBR/aerobic CSTR reactor system, respectively. The toxicity tests performed with Photobacterium phosphoreum (LCK 480, LUMIStox) and Daphnia magna showed that the toxicity decreased with anaerobic/aerobic sequential reactor system from the influent, anaerobic and to

  20. Verification of using SABINE-3.1 code for calculations of radioactive inventory in reactor shield

    International Nuclear Information System (INIS)

    Moukhamadeev, R.; Suvorov, A.

    2000-01-01

    This report presents the results of calculations of radioactive inventory and doses of activation radiation for the International Benchmark Calculations of Radioactive Inventory for Fission Reactor Decommissioning, IAEA, and measurements of activation doses in shield of WWER-440 (Armenian NPP), using one-dimension modified code SABINE-3.1. For decommissioning of NPP it is very important to evaluate in correct manner radioactive inventory in reactor construction and shield materials. One-dimension code SABINE-3.1 (removing-diffusion method for neutron calculation) was modified to perform calculation of radioactive inventory in reactor shield materials and dose from activation photons behind them. These calculations are carried out on the base of nuclear constant system ABBN-78 and new library of activation data for a number of long-lived isotopes, prepared by authors on the base of [9], which present at shield materials as microimpurities and manage radiation situation under the decay more than 1 year. (Authors)

  1. Generation IV reactors: economics

    International Nuclear Information System (INIS)

    Dupraz, B.; Bertel, E.

    2003-01-01

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

  2. Analysis of the VVER-440 reactor steam generator secondary side with the RELAP5/MOD3 code

    International Nuclear Information System (INIS)

    Tuunanen, J.

    1993-01-01

    Nuclear Engineering Laboratory of the Technical Research Centre of Finland has widely used RELAP5/MOD2 and -MOD3 codes to simulate horizontal steam generators. Several models have been developed and successfully used in the VVER-safety analysis. Nevertheless, the models developed have included only rather few nodes in the steam generator secondary side. The secondary side has normally been divided into about 10 to 15 nodes. Since the secondary side at the steam generators of VVER-440 type reactors consists of a rather large water pool, these models were only roughly capable to predict secondary side flows. The paper describes an attempt to use RELAP5/MOD3 code to predict secondary side flows in a steam generator of a VVER-440 reactor. A 2D/3D model has been developed using RELAP5/MOD3 codes cross-flow junctions. The model includes 90 volumes on the steam generator secondary side. The model has been used to calculate steady state flow conditions in the secondary side of a VVER-440 reactor steam generator. (orig.) (1 ref., 9 figs., 2 tabs.)

  3. Method of operating a nuclear reactor

    International Nuclear Information System (INIS)

    Gyorey, G.L.; Parkos, G.R.; Roupe, G.A.; Thomson, O.A.; Crowther, R.L.

    1979-01-01

    The invention concerns the configuration of control rods in the lattice of the reactor core, as well as an instruction on the sequence of with drawal for the control rods, arranged in groups, in order to achieve for the control rod reactivity of the control rods remaining in the reactor core to adopt the lowest possible value. The rods are combined in several 3 x 3 matrices which in their turn are grouped into two networks. The groups are moved successively according to a specified schedule. There can be achieved maximum control rod reactivities between 0.025 and 0.035 (referred to the totally withdrawn state). (RW) 891 RW/RW 892 MKO [de

  4. State space modeling of reactor core in a pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ashaari, A.; Ahmad, T.; M, Wan Munirah W. [Department of Mathematical Science, Faculty of Science, Universiti Teknologi Malaysia, 81310 Skudai, Johor (Malaysia); Shamsuddin, Mustaffa [Institute of Ibnu Sina, Universiti Teknologi Malaysia, 81310 Skudai, Johor (Malaysia); Abdullah, M. Adib [Swinburne University of Technology, Faculty of Engineering, Computing and Science, Jalan Simpang Tiga, 93350 Kuching, Sarawak (Malaysia)

    2014-07-10

    The power control system of a nuclear reactor is the key system that ensures a safe operation for a nuclear power plant. However, a mathematical model of a nuclear power plant is in the form of nonlinear process and time dependent that give very hard to be described. One of the important components of a Pressurized Water Reactor is the Reactor core. The aim of this study is to analyze the performance of power produced from a reactor core using temperature of the moderator as an input. Mathematical representation of the state space model of the reactor core control system is presented and analyzed in this paper. The data and parameters are taken from a real time VVER-type Pressurized Water Reactor and will be verified using Matlab and Simulink. Based on the simulation conducted, the results show that the temperature of the moderator plays an important role in determining the power of reactor core.

  5. Simulation software of 3-D two-neutron energy groups for ship reactor with hexagonal fuel subassembly

    International Nuclear Information System (INIS)

    Zhang Fan; Cai Zhangsheng; Yu Lei; Gui Xuewen

    2005-01-01

    Core simulation software for 3-D two-neutron energy groups is developed. This software is used to simulate the ship reactor with hexagonal fuel subassembly after 10, 150 and 200 burnup days, considering the hydraulic and thermal feedback. It accurately simulates the characteristics of the fast and thermal neutrons and the detailed power distribution in a reactor under normal and abnormal operation condition. (authors)

  6. Plasma nitriding monitoring reactor: A model reactor for studying plasma nitriding processes using an active screen

    Science.gov (United States)

    Hamann, S.; Börner, K.; Burlacov, I.; Spies, H.-J.; Strämke, M.; Strämke, S.; Röpcke, J.

    2015-12-01

    A laboratory scale plasma nitriding monitoring reactor (PLANIMOR) has been designed to study the basics of active screen plasma nitriding (ASPN) processes. PLANIMOR consists of a tube reactor vessel, made of borosilicate glass, enabling optical emission spectroscopy (OES) and infrared absorption spectroscopy. The linear setup of the electrode system of the reactor has the advantages to apply the diagnostic approaches on each part of the plasma process, separately. Furthermore, possible changes of the electrical field and of the heat generation, as they could appear in down-scaled cylindrical ASPN reactors, are avoided. PLANIMOR has been used for the nitriding of steel samples, achieving similar results as in an industrial scale ASPN reactor. A compact spectrometer using an external cavity quantum cascade laser combined with an optical multi-pass cell has been applied for the detection of molecular reaction products. This allowed the determination of the concentrations of four stable molecular species (CH4, C2H2, HCN, and NH3). With the help of OES, the rotational temperature of the screen plasma could be determined.

  7. Plasma nitriding monitoring reactor: A model reactor for studying plasma nitriding processes using an active screen

    International Nuclear Information System (INIS)

    Hamann, S.; Röpcke, J.; Börner, K.; Burlacov, I.; Spies, H.-J.; Strämke, M.; Strämke, S.

    2015-01-01

    A laboratory scale plasma nitriding monitoring reactor (PLANIMOR) has been designed to study the basics of active screen plasma nitriding (ASPN) processes. PLANIMOR consists of a tube reactor vessel, made of borosilicate glass, enabling optical emission spectroscopy (OES) and infrared absorption spectroscopy. The linear setup of the electrode system of the reactor has the advantages to apply the diagnostic approaches on each part of the plasma process, separately. Furthermore, possible changes of the electrical field and of the heat generation, as they could appear in down-scaled cylindrical ASPN reactors, are avoided. PLANIMOR has been used for the nitriding of steel samples, achieving similar results as in an industrial scale ASPN reactor. A compact spectrometer using an external cavity quantum cascade laser combined with an optical multi-pass cell has been applied for the detection of molecular reaction products. This allowed the determination of the concentrations of four stable molecular species (CH 4 , C 2 H 2 , HCN, and NH 3 ). With the help of OES, the rotational temperature of the screen plasma could be determined

  8. Plasma nitriding monitoring reactor: A model reactor for studying plasma nitriding processes using an active screen

    Energy Technology Data Exchange (ETDEWEB)

    Hamann, S., E-mail: hamann@inp-greifswald.de; Röpcke, J. [INP-Greifswald, Felix-Hausdorff-Str. 2, 17489 Greifswald (Germany); Börner, K.; Burlacov, I.; Spies, H.-J. [TU Bergakademie Freiberg, Institute of Materials Engineering, Gustav-Zeuner-Str. 5, 09599 Freiberg (Germany); Strämke, M.; Strämke, S. [ELTRO GmbH, Arnold-Sommerfeld-Ring 3, 52499 Baesweiler (Germany)

    2015-12-15

    A laboratory scale plasma nitriding monitoring reactor (PLANIMOR) has been designed to study the basics of active screen plasma nitriding (ASPN) processes. PLANIMOR consists of a tube reactor vessel, made of borosilicate glass, enabling optical emission spectroscopy (OES) and infrared absorption spectroscopy. The linear setup of the electrode system of the reactor has the advantages to apply the diagnostic approaches on each part of the plasma process, separately. Furthermore, possible changes of the electrical field and of the heat generation, as they could appear in down-scaled cylindrical ASPN reactors, are avoided. PLANIMOR has been used for the nitriding of steel samples, achieving similar results as in an industrial scale ASPN reactor. A compact spectrometer using an external cavity quantum cascade laser combined with an optical multi-pass cell has been applied for the detection of molecular reaction products. This allowed the determination of the concentrations of four stable molecular species (CH{sub 4}, C{sub 2}H{sub 2}, HCN, and NH{sub 3}). With the help of OES, the rotational temperature of the screen plasma could be determined.

  9. Activity report on the utilization of research reactors (JRR-3 and JRR-4). Japanese fiscal year, 2008

    International Nuclear Information System (INIS)

    2014-02-01

    JRR-3 is used for the purposes below; Experimental studies such as neutron scattering, prompt gamma-ray analyses, neutron radiography, Irradiation for activation analyses, radioisotope (RI) productions, fission tracks, Irradiation test of reactor materials, etc. JRR-4 is used for the purposes below; Medical irradiation (Boron Neutron Capture Therapy : BNCT), Prompt gamma-ray analyses, Sensitivity measurement of radiation detectors, Experiment and practice in the nuclear reactor training, Irradiation for activation analyses, RI productions, fission tracks etc. In the fiscal year 2008, the research reactor JRR-3 was operated for 7 cycles (cycle operation : 26days/cycle) for utilization sharing of facility. The research reactor JRR-4 was not operated in 2008. Because a crack was found on the weld of the aluminum cladding of a graphite reflector element. JRR-4 has remained shutdown until the reflector elements were replaced. The volume contains 250activity reports, which are categorized into the fields of neutron scattering (11 subcategories), neutron radiography, neutron activation analyses, and others submitted by the users in JAEA and other Organizations. (author)

  10. Preliminary design of a Binary Breeder Reactor; Diseno preliminar de un reactor esferico de quema/cria

    Energy Technology Data Exchange (ETDEWEB)

    Garcia C, E. Y.; Francois, J. L.; Lopez S, R. C., E-mail: eliasgarcerv@hotmail.com [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac No. 8532, 62550 Jiutepec, Morelos (Mexico)

    2014-10-15

    A binary breeder reactor (BBR) is a reactor that by means of the transmutation and fission process can operates through the depleted uranium burning with a small quantity of fissile material. The advantages of a BBR with relation to other nuclear reactor types are numerous, taking into account their capacity to operate for a long time without requiring fuel reload or re-arrangement. In this work four different simulations are shown carried out with the MCNPX code with libraries Jeff-3.1 to 1200 K. The objective of this study is to compare two different models of BBR: a spherical reactor and a cylindrical one, using two fuel cycles for each one of them (U-Pu and Th-U) and different reflectors for the two different geometries. For all the models a super-criticality state was obtained at least 10.9 years without carrying out some fuel re-arrangement or reload. The plutonium-239 production was achieved in the models where natural uranium was used in the breeding area, while the production of uranium-233 was observed in the cases where thorium was used in the fertile area. Finally, a behavior of stationary wave reactor was observed inside the models of spherical reactor when contemplating the power uniform increment in the breeding area, while inside the cylindrical models was observed the behavior of a traveling wave reactor when registering the displacement of the burnt wave along the cylindrical model. (Author)

  11. Core thermohydraulic design with LEU fuels for upgraded research reactor, JRR-3

    Energy Technology Data Exchange (ETDEWEB)

    Sudo, Y; Ando, H; Ikawa, H; Ohnishi, N [Department of Research Reactor Operation, Japan Atomic Energy Research Institute (JAERI), 319-11 Tokai-Mura, Ibaraki-Ken (Japan)

    1985-07-01

    This paper presents the outline of core thermohydraulic design and analysis of the research reactor, JRR-3, which is to be upgraded to a 20 MWt pool-type, light water-cooled reactor with 20% LEU plate-type fuels. The major feature of core thermohydraulics of the upgraded JRR-3 is that core flow is a downflow at the condition of normal operation, with which fuel plates are exposed to a severer condition than with an upflow in case of operational transients and accidents. The core thermo-hydraulic design was, therefore, done for the condition of normal operation so that fuel plates may have enough safety margin both against the onset of nucleate boiling not to allow the nucleate boiling anywhere in the core and against the initiation of DNB, and the safety margin for these were evaluated. The core velocity thus designed is at the optimum condition where fuel plates have the maximum margin against the onset of nucleate boiling. The core thermohydraulic characteristics were also clarified for the natural circulation cooling mode. (author)

  12. Status of French reactors

    International Nuclear Information System (INIS)

    Ballagny, A.

    1997-01-01

    The status of French reactors is reviewed. The ORPHEE and RHF reactors can not be operated with a LEU fuel which would be limited to 4.8 g U/cm 3 . The OSIRIS reactor has already been converted to LEU. It will use U 3 Si 2 as soon as its present stock of UO 2 fuel is used up, at the end of 1994. The decision to close down the SILOE reactor in the near future is not propitious for the start of a conversion process. The REX 2000 reactor, which is expected to be commissioned in 2005, will use LEU (except if the fast neutrons core option is selected). Concerning the end of the HEU fuel cycle, the best option is reprocessing followed by conversion of the reprocessed uranium to LEU

  13. Numerical modeling of a nuclear production reactor cooling lake

    International Nuclear Information System (INIS)

    Hamm, L.L.; Pepper, D.W.

    1987-01-01

    A finite element model has been developed which predicts flow and temperature distributions within a nuclear reactor cooling lake at the Savannah River Plant near Aiken, South Carolina. Numerical results agree with values obtained from a 3-D EPA numerical lake model and actual measurements obtained from the lake. Because the effluent water from the reactor heat exchangers discharges directly into the lake, downstream temperatures at mid-lake could exceed the South Carolina DHEC guidelines for thermal exchanges during the summer months. Therefore, reactor power was reduced to maintain temperature compliance at mid-lake. Thermal mitigation measures were studied that included placing a 6.1 m deep fabric curtain across mid-lake and moving the reactor outfall upstream. These measurements were calculated to permit about an 8% improvement in reactor power during summer operation

  14. Biological biogas upgrading capacity of a hydrogenotrophic community in a trickle-bed reactor

    International Nuclear Information System (INIS)

    Rachbauer, Lydia; Voitl, Gregor; Bochmann, Günther; Fuchs, Werner

    2016-01-01

    Highlights: • Data on long term operation of a system supplied with real biogas are presented. • Ex-situ biological methanation is feasible for biogas upgrading. • Gas quality obtained complies with strictest direct grid injection criteria. • Biomethane can act as flexible storage for renewable surplus electricity. - Abstract: The current study reports on biological biogas upgrading by means of hydrogen addition to obtain biomethane. A mesophilic (37 °C) 0.058 m"3 trickle-bed reactor with an immobilized hydrogenotrophic enrichment culture was operated for a period of 8 months using a substrate mix of molecular hydrogen (H_2) and biogas (36–42% CO_2). Complete CO_2 conversion (> 96%) was achieved up to a H_2 loading rate of 6.5 m_n"3 H_2/m"3_r_e_a_c_t_o_r _v_o_l_. × d, corresponding to 2.3 h gas retention time. The optimum H_2/CO_2 ratio was determined to be between 3.67 and 4.15. CH_4 concentrations above 96% were achieved with less than 0.1% residual H_2. This gas quality complies even with tightest standards for grid injection without the need for additional CO_2 removal. If less rigid standards must be fulfilled H_2 loading rates can be almost doubled (10.95 versus 6.5 m_n"3 H_2/m"3_r_e_a_c_t_o_r _v_o_l_. × d) making the process even more attractive. At this H_2 loading the achieved methane productivity was 2.52 m_n"3 CH_4/m"3_r_e_a_c_t_o_r _v_o_l_. × d. In terms of biogas this corresponds to an upgrading capacity of 6.9 m_n"3 biogas/m"3_r_e_a_c_t_o_r _v_o_l_. × d. The conducted experiments demonstrate that biological methanation in an external reactor is well feasible for biogas upgrading under the prerequisite that an adequate H_2 source is available.

  15. Proposal of a benchmark for core burnup calculations for a VVER-1000 reactor core

    International Nuclear Information System (INIS)

    Loetsch, T.; Khalimonchuk, V.; Kuchin, A.

    2009-01-01

    In the framework of a project supported by the German BMU the code DYN3D should be further validated and verified. During the work a lack of a benchmark on core burnup calculations for VVER-1000 reactors was noticed. Such a benchmark is useful for validating and verifying the whole package of codes and data libraries for reactor physics calculations including fuel assembly modelling, fuel assembly data preparation, few group data parametrisation and reactor core modelling. The benchmark proposed specifies the core loading patterns of burnup cycles for a VVER-1000 reactor core as well as a set of operational data such as load follow, boron concentration in the coolant, cycle length, measured reactivity coefficients and power density distributions. The reactor core characteristics chosen for comparison and the first results obtained during the work with the reactor physics code DYN3D are presented. This work presents the continuation of efforts of the projects mentioned to estimate the accuracy of calculated characteristics of VVER-1000 reactor cores. In addition, the codes used for reactor physics calculations of safety related reactor core characteristics should be validated and verified for the cases in which they are to be used. This is significant for safety related evaluations and assessments carried out in the framework of licensing and supervision procedures in the field of reactor physics. (authors)

  16. Evaluation of Pressure Changes in HANARO Reactor Hall after a Reactor Shutdown

    International Nuclear Information System (INIS)

    Han, Geeyang; Han, Jaesam; Ahn, Gukhoon; Jung, Hoansung

    2013-01-01

    The major objective of this work is intended to evaluate the characteristics of the thermal behavior regarding how the decay heat will be affected by the reactor hall pressure change and the increase of pool water temperature induced in the primary coolant after a reactor shutdown. The particular reactor pool water temperature at the surface where it is evaporated owing to the decay heat resulting in the local heat transfer rate is related to the pressure change response in the reactor hall associated with the primary cooling system because of the reduction of the heat exchanger to remove the heat. The increase in the pool water temperature is proportional to the heat transfer rate in the reactor pool. Consequently, any limit on the reactor pool water temperature imposes a corresponding limit on the reactor hall pressure. At HANARO, the decay heat after a reactor shutdown is mainly removed by the natural circulation cooling in the reactor pool. This paper is written for the safety feature of the pressure change related leakage rate from the reactor hall. The calculation results show that the increase of pressure in the reactor hall will not cause any serious problems to the safety limits although the reactor hall pressure is slightly increased. Therefore, it was concluded that the pool water temperature increase is not so rapid as to cause the pressure to vary significantly in the reactor hall. Furthermore, the mathematical model developed in this work can be a useful analytical tool for scoping and parametric studies in the area of thermal transient analysis, with its proper representation of the interaction between the temperature and pressure in the reactor hall

  17. Plutonium-burn high temperature gas-cooled reactor for 3E+3S

    International Nuclear Information System (INIS)

    Okamoto, Koji

    2015-01-01

    The Nuclear Energy Development in Japan is facing a very difficult conditions after Fukushima-Daiichi NPP Accident. Nuclear Energy has strong advantages on 3E, i.e., Energy security, Economical efficiency and Environment. However, people does not believe the Safety 'S' of Nuclear Energy, now. The disadvantage of 'S' overrides the advantages of '3E'. In Nuclear Energy, 'S' is expanded into 3S, i.e., Safety, Security and Safeguards. Especially, the management of Plutonium inventory in Spent Fuel generated by the NPP operation is very important in the viewpoints of non-proliferation. The high-temperature gas cooled reactor (HTGR) is the solution of these disadvantages of '3S' in Nuclear Energy. The fuel of HTGR is composed by 1 mm spherical fuel particle, i.e., TRISO made by fuel, graphite and silicon-carbide. The silicon-carbide can confine the fission products in any conditions of fuel life cycle, i.e., during operation, accidents and disposal for 1 million years. The confinement of the radioactive materials can be confirmed by the TRISO. The HTGR core has strong negative feedback for temperature. So, the fission automatically stopped at the accidental conditions, such as loss of flow and LOCA. Also, the residual heat can be cooled by the radiation heat transfer to reactor vessel wall. The HTGR system usually has passive vessel wall cooling system. When the passive cooling system had been failed, the heat can be transferred to the land by heat conductions, and fuel does not reach the SiC broken temperature. The fission chain reaction has been stopped automatically by negative feedback, i.e., physics. The residual heat had been cooled automatically by radiation. The radioactive materials had been confined automatically by silicon-carbide. The HTGR is superior for 'S' safety. Plutonium can be burned by the HTGR. In the viewpoints of non-proliferation, the fuel should be made by YSZ-PuO 2 , stabilized buffer

  18. Verification of the CASMO-3/SIMULATE-3 pin power accuracy by comparison with operating boiling water reactor measurements

    International Nuclear Information System (INIS)

    Uegata, T.; Saji, E.; Tanaka, H.

    1993-01-01

    Intranodal pin power distributions calculated by the CASMO-3/SIMULATE-3 code have been compared with pin gamma scan measurements. These data were obtained from the depleted core of an operating boiling water reactor (BWR), which is more complicated than a pressurized water reactor to calculate because of the existence of coolant void distributions and cruciform control blades. Furthermore, measured bundles include mixed-oxide (MOX) bundles in which steep thermal flux gradients occur. Both UO 2 and MOX bundles have been calculated in the same manner based on the standard CASMO-3/SIMULATE-3 methods. The total pin power root-mean-square (rms) error is 2.7%, which includes measurement error, from an 896-point comparison. There is no obvious dependency on axial elevations (void fractions) and no significant difference between fuel types (UO 2 or MOX), although the errors in a peripheral bundle, which is less important from the standpoint of core design, are somewhat larger than those in the internal bundles. If the peripheral bundle is excluded, the total rms error is reduced to 2.2%. From these results, it is concluded that excellent agreement has been obtained between the calculations and measurements and that the calculational capability of CASMO-3/SIMULATE-3 for the intranodal pin power distribution is quite satisfactory and useful for BWR core design

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  20. Convective cooling in a pool-type research reactor

    Science.gov (United States)

    Sipaun, Susan; Usman, Shoaib

    2016-01-01

    A reactor produces heat arising from fission reactions in the nuclear core. In the Missouri University of Science and Technology research reactor (MSTR), this heat is removed by natural convection where the coolant/moderator is demineralised water. Heat energy is transferred from the core into the coolant, and the heated water eventually evaporates from the open pool surface. A secondary cooling system was installed to actively remove excess heat arising from prolonged reactor operations. The nuclear core consists of uranium silicide aluminium dispersion fuel (U3Si2Al) in the form of rectangular plates. Gaps between the plates allow coolant to pass through and carry away heat. A study was carried out to map out heat flow as well as to predict the system's performance via STAR-CCM+ simulation. The core was approximated as porous media with porosity of 0.7027. The reactor is rated 200kW and total heat density is approximately 1.07+E7 Wm-3. An MSTR model consisting of 20% of MSTR's nuclear core in a third of the reactor pool was developed. At 35% pump capacity, the simulation results for the MSTR model showed that water is drawn out of the pool at a rate 1.28 kg s-1 from the 4" pipe, and predicted pool surface temperature not exceeding 30°C.

  1. TRIGA reactor main systems

    International Nuclear Information System (INIS)

    Boeck, H.; Villa, M.

    2007-01-01

    This module describes the main systems of low power (<2 MW) and higher power (≥2 MW) TRIGA reactors. The most significant difference between the two is that forced reactor cooling and an emergency core cooling system are generally required for the higher power TRIGA reactors. However, those TRIGA reactors that are designed to be operated above 3 MW also use a TRIGA fuel that is specifically designed for those higher power outputs (3 to 14 MW). Typical values are given for the respective systems although each TRIGA facility will have unique characteristics that may only be determined by the experienced facility operators. Due to the inherent wide scope of these research reactor facilities construction and missions, this training module covers those systems found at most operating TRIGA reactor facilities but may also discuss non-standard equipment that was found to be operationally useful although not necessarily required. (author)

  2. UCLA program in reactor studies: The ARIES tokamak reactor study

    International Nuclear Information System (INIS)

    1991-01-01

    The ARIES research program is a multi-institutional effort to develop several visions of tokamak reactors with enhanced economic, safety, and environmental features. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Four ARIES visions are currently planned for the ARIES program. The ARIES-1 design is a DT-burning reactor based on ''modest'' extrapolations from the present tokamak physics database and relies on either existing technology or technology for which trends are already in place, often in programs outside fusion. ARIES-2 and ARIES-4 are DT-burning reactors which will employ potential advances in physics. The ARIES-2 and ARIES-4 designs employ the same plasma core but have two distinct fusion power core designs; ARIES-2 utilize the lithium as the coolant and breeder and vanadium alloys as the structural material while ARIES-4 utilizes helium is the coolant, solid tritium breeders, and SiC composite as the structural material. Lastly, the ARIES-3 is a conceptual D- 3 He reactor. During the period Dec. 1, 1990 to Nov. 31, 1991, most of the ARIES activity has been directed toward completing the technical work for the ARIES-3 design and documenting the results and findings. We have also completed the documentation for the ARIES-1 design and presented the results in various meetings and conferences. During the last quarter, we have initiated the scoping phase for ARIES-2 and ARIES-4 designs

  3. Construction of Research Reactors for Gen 3 and Gen 4 Reactors Development

    International Nuclear Information System (INIS)

    Behar, Christophe

    2014-01-01

    Christophe Behar, Director of the Nuclear Energy Division at CEA, detailed the different kind of research reactors and the issues in term of investment, use, side application such as the medical isotopes production

  4. Evaluation on activation activity of reactor in JRR-2 applied 3 dimensional model to neutron flux calculation

    International Nuclear Information System (INIS)

    Kishimoto, Katsumi; Arigane, Kenji

    2005-03-01

    Revaluation to activation activity of reactor evaluated at the notification of dismantling submitted in 1997 was carried out in JRR-2 where decommissioning was advanced now. In the revaluation, estimation accuracy on neutron streaming at various horizontal experimental tubes was improved by applying 3 dimensional model to neutron transport calculation that had been carried out by 2 dimensional model, and calculating with TORT. As the result, excessive overestimations on horizontal experimental tubes and biological shield that had greatly contributed to total activation activity in evaluation at the notification of dismantling was revised, sum of their activation activities in the revaluation decreased to 1/18 (case after 1 year from the permanent shutdown of reactor) of evaluation at the notification of dismantling, and the structural materials that had large activation activity were changed. By the above, it was shown that introducing 3 dimensional model was effective in evaluation on activation activity of the research reactor that had a lot of various experimental tubes. Total activation activity of reactor by the revaluation depended on control rods, thermal shield plates and horizontal experimental tubes, and the value after 1 year from the permanent shutdown of reactor was 1.9x10 14 Bq. (author)

  5. Advances in reactor physics education: Visualization of reactor parameters

    International Nuclear Information System (INIS)

    Snoj, L.; Kromar, M.; Zerovnik, G.

    2012-01-01

    Modern computer codes allow detailed neutron transport calculations. In combination with advanced 3D visualization software capable of treating large amounts of data in real time they form a powerful tool that can be used as a convenient modern educational tool for reactor operators, nuclear engineers, students and specialists involved in reactor operation and design. Visualization is applicable not only in education and training, but also as a tool for fuel management, core analysis and irradiation planning. The paper treats the visualization of neutron transport in different moderators, neutron flux and power distributions in two nuclear reactors (TRIGA type research reactor and a typical PWR). The distributions are calculated with MCNP and CORD-2 computer codes and presented using Amira software. (authors)

  6. Rotating reactors : a review

    NARCIS (Netherlands)

    Visscher, F.; Schaaf, van der J.; Nijhuis, T.A.; Schouten, J.C.

    2013-01-01

    This review-perspective paper describes the current state-of-the-art in the field of rotating reactors. The paper has a focus on rotating reactor technology with applications at lab scale, pilot scale and industrial scale. Rotating reactors are classified and discussed according to their geometry:

  7. Detailed modeling of KALININ-3 NPP VVER-1000 reactor pressure vessel by the coupled system code ATHLET/BIPR-VVER

    International Nuclear Information System (INIS)

    Nikonov, S.P.; Velkov, K.; Pautz, A.

    2011-01-01

    The paper gives an overview of the recent developments of a new reactor pressure vessel (RPV) model of VVER-1000 for the coupled system code ATHLET/BIPR-VVER. Based on the previous experience a methodology is worked out for modeling the RPV in a pseudo-3D way with the help of a multiple parallel thermal-hydraulic channel scheme that follows the hexagonal fuel assembly structure from the bottom to the top of the reactor. The results of the first application of the new modeling are discussed on the base of the OECD/NEA coupled code benchmark for Kalinin-3 NPP transient. Coolant mass flow distributions in reactor volume of VVER 1000 reactor are presented and discussed. It is shown that along the core height a mass flow re-distribution of the coolant takes place starting approximately at an axial layer located 1 meter below the core outlet. (author)

  8. Biodegradation of chlorinated hydrocarbons in a vapor phase reactor

    International Nuclear Information System (INIS)

    Ensley, B.D.

    1992-01-01

    A bench scale gas lift loop reactor was constructed to evaluate the feasibility of trichloroethylene (TCE) degradative microorganisms being used to treat TCE contaminated air. Two different microorganisms were used as biocatalysts in this reactor. After proper operating conditions were established for use of this reactor/biocatalyst combination, both microorganisms could degrade 95% of inlet TCE at air flow rates of up to 3% of the total reactor volume per minute. TCE concentrations of between 300 μg/L (60ppmv) and 3000 μg/L (600 ppmv) were degraded with 95% or better efficiency. Preliminary economic evaluations suggest that bioremediation may be the low cost alternative for treating certain TCE contaminated air streams and field trials of a scaled-up reactor system based on this technology are currently underway

  9. DRAGON 3.05D, Reactor Cell Calculation System with Burnup

    International Nuclear Information System (INIS)

    2007-01-01

    1 - Description of program or function: The computer code DRAGON contains a collection of models that can simulate the neutron behavior of a unit cell or a fuel assembly in a nuclear reactor. It includes all of the functions that characterize a lattice cell code, namely: the interpolation of microscopic cross sections supplied by means of standard libraries; resonance self-shielding calculations in multidimensional geometries; multigroup and multidimensional neutron flux calculations that can take into account neutron leakage; transport-transport or transport-diffusion equivalence calculations as well as editing of condensed and homogenized nuclear properties for reactor calculations; and finally isotopic depletion calculations. 2 - Methods: The code DRAGON contains a multigroup flux solver conceived that can use a various algorithms to solve the neutron transport equation for the spatial and angular distribution of the flux. Each of these algorithms is presented in the form of a one-group solution procedure where the contributions from other energy groups are considered as sources. The current release of DRAGON contains five such algorithms. The JPM option that solves the integral transport equation using the J+- method, (interface current method applied to homogeneous blocks); the SYBIL option that solves the integral transport equation using the collision probability method for simple one dimensional (1-D) or two dimensional (2-D) geometries and the interface current method for 2-D Cartesian or hexagonal assemblies; the EXCELL/NXT option to solve the integral transport equation using the collision probability method for more general 2-D geometries and for three dimensional (3-D) assemblies; the MOCC option to solve the transport equation using the method of cyclic characteristics in 2-D Cartesian, and finally the MCU option to solve the transport equation using the method of characteristics (non cyclic) for 3-D Cartesian geometries. The execution of DRAGON is

  10. Fuel slugs considered for use in the high flux reactor EL3; Elements combustibles envisages pour la pile a haut flux EL 3

    Energy Technology Data Exchange (ETDEWEB)

    Stohr, J A; Caillat, R; Gauthron, M; Montagne, R [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1958-07-01

    EL3 was designed essentially for the study, under irradiation conditions, of materials used in the construction of atomic reactors. The study schedule allocates considerable time and effort to new types of fuel slugs. The present report described the various types of slug being tested or scheduled for tests. After laboratory study, each slug is tested in an experimental cell in the pile. The best are retained and used to charge the reactor (the present charge is purely provisional to permit first criticality and power rise tests)ren. [French] La pile EL3 est essentiellement destinee a l'etude sous irradiation des materiaux utilises dans la construction des reacteurs atomiques. Dans ce programme, une tres large part est reservee a l'etude de nouveaux elements combustibles. Le present rapport decrit les differentes solutions de cartouches dont l'essai est envisage ou en cours. Apres etude en laboratoire, chacune de ces solutions est testee dans une cellule experimentale en pile. Les meilleures seront retenues pour constituer le chargement normal de la pile (le chargement actuel etant essentiellement une solution provisoire qui a permis la divergence de la pile et les premiers essais de montee en puissance). (auteur)

  11. Cirus reactor: a milestone in Indian Atomic Energy Programme

    International Nuclear Information System (INIS)

    Ranjan, Rakesh; Karhadkar, C.G.; Bhattacharya, S.

    2017-01-01

    Cirus, a 40 MW_t_h, high flux, thermal neutron research reactor achieved first criticality on 10"t"h July 1960. It had vertical core, natural metallic Uranium rods in Aluminium clad as fuel, demineralised light water as coolant, heavy water as moderator, Helium as cover gas and graphite as reflector. A low-pressure containment was provided for the reactor and some of the important associated reactor systems. Reactor start-up and power regulation was effected by controlled adjustment of moderator level in the reactor vessel. Boron carbide rods were used as primary shut down devices. Dumping of heavy water from core worked as secondary shut down device. Seawater was used as secondary coolant for removal of the fission heat of the reactor. Initial operation of Cirus was marred by several difficulties, primarily arising out of water chemistry in primary cooling water system. It took almost 3 years to systematically resolve these problems and achieve stable operation of reactor. Cirus could be operated at its rated power by the year 1963

  12. Radiation protection at the RA Reactor in 1993, RA research reactor, Part

    International Nuclear Information System (INIS)

    Ninkovic, M.; Pavlovic, R.; Mandic, M.; Sipka, V.; Grsic, Z.

    1993-01-01

    Radiation protection tasks which enable safe operation of the RA reactor, and are defined according the the legal regulations and IAEA safety recommendations are sorted into four categories in this report: (1) Control of the working environment, dosimetry and radiation protection at the RA reactor; (2) decontamination, collecting and treatment of fluid effluents and solid wastes; (3) Radioactivity control in the vicinity of the reactor and (4)meteorology measurements; (3). Each of the category is described as a separate annex of this report [sr

  13. Effect of fuel assembly when changing from AFA 2G to AFA 3G on seismic loads of reactor internal

    International Nuclear Information System (INIS)

    Liu Wenjin; Zeng Zhongxiu; Ye Xianhui; Wu Wanjun

    2013-01-01

    Nonlinear seismic model for reactor with fuel assemblies of AFA 2G and AFA 3G is established. Using ANSYS software, seismic nonlinear time -history analysis is completed and the effects on seismic loads of reactor system are obtained. The result shows that when the fuel assembly changing from AFA 2G to AFA 3G, it is necessary to reevaluate the fuel assembly itself, but not the reactor internal. (authors)

  14. Comparative studies of JENDL-3.3, JENDL-3.2, JEFF-3, JEF-2.2 and ENDF/B-6.8 data libraries on the Monte Carlo continuous energy modeling of the gas turbine-modular helium reactor operating with thorium fuels

    International Nuclear Information System (INIS)

    Talamo, Alberto; Gudowski, Waclaw

    2005-01-01

    One of the major benefits of the Gas Turbine-Modular Helium Reactor is the capability to operate with several different types of fuel; either Light Water Reactors waste, military plutonium or thorium represent valid candidates as possible types of fuel. In the present studies, we performed a comparison of various nuclear data libraries by the Monte Carlo Continuous Energy Burnup Code MCB applied to the Gas Turbine-Modular Helium Reactor operating on a thorium fuel. A thorium fuel offers valuable attractive advantages: low fuel cost, high reduction of actinides production and the possibility to enable the reactor to act as a breeder of fuel by the neutron capture of fertile 232 Th. We evaluated the possibility to mix thorium with small quantities, about 3% in atomic composition, of 239 Pu, 233 U and 235 U. The mass of thorium must be much larger than that one of plutonium or uranium because of the low capture cross section of thorium compared to the fission one of the fissile nuclides; at the same time, the quantity of the fissile isotopes must grant the criticality condition. These two simultaneous constraints force to load a huge mass of fuel in the reactor; consequently, we propose to allocate the fuel in TRISO particles with a large radius of the kernel. For each of the three different fuels we calculated the evolution of the fuel composition by the MCB code equipped with five different nuclear data libraries: JENDL-3.3, JENDL-3.2, JEFF-3, JEF-2.2 and ENDF/B. (author)

  15. Validation and application of 3D-methods for the design and safety analysis of high temperature reactors

    International Nuclear Information System (INIS)

    Bader, J.; Lapins, J.; Buck, M; Bernnat, W.; Laurien, E.

    2011-01-01

    Some of the concepts for future nuclear reactors are high-temperature gas-cooled reactors. Previous simulation codes for their cores were often based on one- or two-dimensional models, but today's increasing computer capabilities make an advance to 3D-codes possible now. Our thermal-hydraulic code ATTICA3D (Advanced Thermal-hydraulic Tool for In-vessel and Core Analysis in 3 Dimensions) is based on the porous media approach, including 3-D models of heat conduction and gas flow, using a coarse-grid integration method for the time-dependent conservation equations of mass, momentum and energy. Results of numerical calculations for various validation cases are presented: First, the test facility SANA is chosen, which has been used to study heat transfer phenomena inside a coolant-gas filled pebble-bed core, which was heated by embedded electrical heating elements. Calculations were carried out for different tests taken from the experimental database. Measured and calculated temperatures at different positions are compared and found in good agreement. Second, our code was used to simulate a depressurized loss of forced cooling experiment with simulated decay heat in the AVR Experimental Reactor. Due to its design with the shut-down rods located inside columnar noses, which extend into the pebble bed of the core, geometry and power distribution are genuinely three-dimensional. The power distribution was calculated by the 3D-Neutronic Diffusion Code CITATION in conjunction with the spectral code MICROX-2. The neutronics and thermal-hydraulics calculations were carried out for a 3D, 45°-degree section of the reactor. It is demonstrated, that the experimental results could be qualitatively reproduced. (author)

  16. Status of French reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ballagny, A. [Commissariat a l`Energie Atomique, Saclay (France)

    1997-08-01

    The status of French reactors is reviewed. The ORPHEE and RHF reactors can not be operated with a LEU fuel which would be limited to 4.8 g U/cm{sup 3}. The OSIRIS reactor has already been converted to LEU. It will use U{sub 3}Si{sub 2} as soon as its present stock of UO{sub 2} fuel is used up, at the end of 1994. The decision to close down the SILOE reactor in the near future is not propitious for the start of a conversion process. The REX 2000 reactor, which is expected to be commissioned in 2005, will use LEU (except if the fast neutrons core option is selected). Concerning the end of the HEU fuel cycle, the best option is reprocessing followed by conversion of the reprocessed uranium to LEU.

  17. Biological oxidation of dissolved methane in effluents from anaerobic reactors using a down-flow hanging sponge reactor.

    Science.gov (United States)

    Hatamoto, Masashi; Yamamoto, Hiroki; Kindaichi, Tomonori; Ozaki, Noriatsu; Ohashi, Akiyoshi

    2010-03-01

    Anaerobic wastewater treatment plants discharge dissolved methane, which is usually not recovered. To prevent emission of methane, which is a greenhouse gas, we utilized an encapsulated down-flow hanging sponge reactor as a post-treatment to biologically oxidize dissolved methane. Within 3 weeks after reactor start-up, methane removal efficiency of up to 95% was achieved with a methane removal rate of 0.8 kg COD m(-3) day(-1) at an HRT of 2 h. After increasing the methane-loading rate, the maximum methane removal rate reached 2.2 kg COD m(-3) day(-1) at an HRT of 0.5 h. On the other hand, only about 10% of influent ammonium was oxidized to nitrate during the first period, but as airflow was increased to 2.5 L day(-1), nitrification efficiency increased to approximately 70%. However, the ammonia oxidation rate then decreased with an increase in the methane-loading rate. These results indicate that methane oxidation occurred preferentially over ammonium oxidation in the reactor. Cloning of the 16S rRNA and pmoA genes as well as phylogenetic and T-RFLP analyses revealed that type I methanotrophs were the dominant methane oxidizers, whereas type II methanotrophs were detected only in minor portion of the reactor. Copyright 2009 Elsevier Ltd. All rights reserved.

  18. Compact stellarators as reactors

    International Nuclear Information System (INIS)

    Lyon, J.F.; Valanju, P.; Zarnstorff, M.C.; Hirshman, S.; Spong, D.A.; Strickler, D.; Williamson, D.E.; Ware, A.

    2001-01-01

    Two types of compact stellarators are examined as reactors: two- and three-field-period (M=2 and 3) quasi-axisymmetric devices with volume-average =4-5% and M=2 and 3 quasi-poloidal devices with =10-15%. These low-aspect-ratio stellarator-tokamak hybrids differ from conventional stellarators in their use of the plasma-generated bootstrap current to supplement the poloidal field from external coils. Using the ARIES-AT model with B max =12T on the coils gives Compact Stellarator reactors with R=7.3-8.2m, a factor of 2-3 smaller R than other stellarator reactors for the same assumptions, and neutron wall loadings up to 3.7MWm -2 . (author)

  19. Neutron transport. Physics and calculation of nuclear reactors with applications to pressurized water reactors and fast neutron reactors. 2 ed.

    International Nuclear Information System (INIS)

    Bussac, J.; Reuss, P.

    1985-01-01

    This book presents the main physical bases of neutron theory and nuclear reactor calculation. 1) Interactions of neutrons with matter and basic principles of neutron transport; 2) Neutron transport in homogeneous medium and the neutron field: kinetic behaviour, slowing-down, resonance absorption, diffusion equation, processing methods; 3) Theory of a reactor constituted with homogeneous zones: critical condition, kinetics, separation of variables, calculation and neutron balance of the fundamental mode, one-group and multigroup theories; 4) Study of heterogeneous cell lattices: fast fission factor, resonance absorption, thermal output factor, diffusion coefficient, computer codes; 5) Operation and control of reactors: perturbation theory, reactivity, fuel properties evolution, poisoning by fission products, calculation of a reactor and fuel management; 6) Study of some types of reactors: PWR and fast breeder reactors, the main reactor types of the present French program [fr

  20. Reactivity Coefficient Calculation for AP1000 Reactor Using the NODAL3 Code

    Science.gov (United States)

    Pinem, Surian; Malem Sembiring, Tagor; Tukiran; Deswandri; Sunaryo, Geni Rina

    2018-02-01

    The reactivity coefficient is a very important parameter for inherent safety and stability of nuclear reactors operation. To provide the safety analysis of the reactor, the calculation of changes in reactivity caused by temperature is necessary because it is related to the reactor operation. In this paper, the temperature reactivity coefficients of fuel and moderator of the AP1000 core are calculated, as well as the moderator density and boron concentration. All of these coefficients are calculated at the hot full power condition (HFP). All neutron diffusion constant as a function of temperature, water density and boron concentration were generated by the SRAC2006 code. The core calculations for determination of the reactivity coefficient parameter are done by using NODAL3 code. The calculation results show that the fuel temperature, moderator temperature and boron reactivity coefficients are in the range between -2.613 pcm/°C to -4.657pcm/°C, -1.00518 pcm/°C to 1.00649 pcm/°C and -9.11361 pcm/ppm to -8.0751 pcm/ppm, respectively. For the water density reactivity coefficients, the positive reactivity occurs at the water temperature less than 190 °C. The calculation results show that the reactivity coefficients are accurate because the results have a very good agreement with the design value.

  1. Farewell to a reactor

    International Nuclear Information System (INIS)

    Skanborg, P.

    1976-01-01

    Denmark's second reactor, DR 2, whose first criticality took place the night of 18/19 December 1958 was shut down for the last time on 31 October 1975. It was a light-water moderrated and cooled reactor of swimming-pool type with a thermal power of 5 MW, using 90% enriched uranium. The operation is described. The reactor and auxiliary equipment are now being put 'in store' - all fuel elements sent for reprocessing, the reactor tank and cooling circuits emptied, and a lead shielding placed over the tank opening. The rest of the equipment will remain in place. (B.P.)

  2. Performance of a UASB reactor treating coffee wet wastewater

    International Nuclear Information System (INIS)

    Guardia Puebla, Yans; Rodríguez Pérez, Suyén; Janet Jiménez Hernández; Sánchez Girón, Víctor

    2014-01-01

    The present work shows the results obtained in the anaerobic digestion process of coffee wet wastewater processing. An UASB anaerobic reactor was operated in single-stage in mesophilic temperature controlled conditions (37±1ºC). The effect of both organic loading rate (OLR) and hydraulic retention time (HRT) in the anaerobic digestion of coffee wet wastewater was investigated. The OLR values considered in the single-stage UASB reactor varied in a range of 3,6-4,1 kgCOD m-3 d-1 and the HRT stayed in a range of 21,5-15,5 hours. The evaluation results show that the best performance of UASB reactor in single-stage was obtained at OLR of 3,6 kg COD m-3 d-1 with an average value of total and soluble COD removal of 77,2% and 83,4%, respectively, and average methane concentration in biogas of 61%. The present study suggests that the anaerobic digestion is suitable to treating coffee wet wastewater. (author)

  3. Measurements of low reactivities using a reactor oscillator

    International Nuclear Information System (INIS)

    Obradovic, D.; Petrovic, M.

    1965-12-01

    Most of the methods of measuring reactivity are limited to the region from several hundreds to several thousands of pcm. The present work develops a method of measuring low reactivities from several pcm to about 600 pcm using the ROB-1 reactor oscillator on the RB reactor of the Boris Kidric Institute of Nuclear Sciences at Vinca. The accuracy of measurement is better than 1%. Several methods are used to measure low reactivities. The most often used is the method based on measuring the stable reactor period. The bottom limit of this method is about 30 porn /1,2/. For control rod calibration the method of rod oscillation is used /3,4/. This method is confronted with considerable influence of space effects /5/. Reference /6/ reports on a method for measuring the reactivity coefficient at a critical level in liquid-moderated reactors. The method is based on measuring reactor response to the oscillation of the moderator about the critical level. The present work reports on a method of determining the reactivity by measuring the phase shift between the perturbation of the effective multiplication factor and reactor response. With the use of the ROB-1 reactor oscillator, the method allows measurement of the reactivity from several pcm to about 600 pcm with an accuracy of 1% (author)

  4. Measurements of low reactivities using a reactor oscillator

    Energy Technology Data Exchange (ETDEWEB)

    Obradovic, D; Petrovic, M [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1965-12-15

    Most of the methods of measuring reactivity are limited to the region from several hundreds to several thousands of pcm. The present work develops a method of measuring low reactivities from several pcm to about 600 pcm using the ROB-1 reactor oscillator on the RB reactor of the Boris Kidric Institute of Nuclear Sciences at Vinca. The accuracy of measurement is better than 1%. Several methods are used to measure low reactivities. The most often used is the method based on measuring the stable reactor period. The bottom limit of this method is about 30 porn /1,2/. For control rod calibration the method of rod oscillation is used /3,4/. This method is confronted with considerable influence of space effects /5/. Reference /6/ reports on a method for measuring the reactivity coefficient at a critical level in liquid-moderated reactors. The method is based on measuring reactor response to the oscillation of the moderator about the critical level. The present work reports on a method of determining the reactivity by measuring the phase shift between the perturbation of the effective multiplication factor and reactor response. With the use of the ROB-1 reactor oscillator, the method allows measurement of the reactivity from several pcm to about 600 pcm with an accuracy of 1% (author)

  5. Verification of RBMK-1500 reactor main circulation circuit model with Cathare V1.3L

    International Nuclear Information System (INIS)

    Jasiulevicius, A.

    2001-01-01

    Among other computer codes, French code CATHARE is also applied for RBMK reactor calculations. In this paper results of such application for Ignalina NPP reactor (RBMK-1500 type) main circulation circuit are presented. Three transients calculations were performed: all main circulation pumps (MCP) trip, trip of one main circulation pump and trip of one main circulation pump without a closure of check valve on the pump line. Calculation results were compared to data from the Ignalina NPP, where all these transients were recorded in the years 1986, 1996 and 1998. The presented studies prove the capability of the CATHARE code to treat thermal-hydraulic transients with a reactor scram in the RBMK, in case of single or multiple pump trips. However, the presented model needs further improvements in order to simulate loss of coolant accidents. For this reason, emergency core cooling system should be included in the model. Additional model improvement is also needed in order to gain more independent pressure behavior in both loops. Also, flow rates through the reactor channels should be modeled by dividing channels into several groups, referring to channel power (in RBMK power produced in a channel, located in different parts of the core is not the same). The point-neutron kinetic model of the CATHARE code is not suitable to predict transients when the reactor is operating at a nominal power level. Such transients would require the use of 3D-neutron kinetics model to describe properly the strong space-time effect on the power distribution in the reactor core

  6. COMSORS: A light water reactor chemical core catcher

    International Nuclear Information System (INIS)

    Forsberg, C.W.; Parker, G.W.; Rudolph, J.C.; Osborne-Lee, I.W.

    1997-01-01

    The Core-Melt Source Reduction System (COMSORS) is a new approach to terminate lightwater reactor (LWR) core-melt accidents and ensure containment integrity. A special dissolution glass made of lead oxide (PbO) and boron oxide (B 2 O 3 ) is placed under the reactor vessel. If molten core debris is released onto the glass, the following sequence happens: (1) the glass absorbs decay heat as its temperature increases and the glass softens; (2) the core debris dissolves into the molten glass; (3) molten glass convective currents create a homogeneous high-level waste (HLW) glass; (4) the molten glass spreads into a wider pool, distributing the heat for removal by radiation to the reactor cavity above or transfer to water on top of the molten glass; and (5) the glass solidifies as increased surface cooling area and decreasing radioactive decay heat generation allows heat removal to exceed heat generation

  7. Survey of group data libraries for use of the DYN3D program for WWER type reactors

    International Nuclear Information System (INIS)

    Mittag, S.

    1994-06-01

    So-called few-group neutron data have to be used as input data in core models (such as DYN3D) calculating the reactor behaviour. A survey is given of qualified data libraries for the reactor cores of Russian VVER. The information about primary data used in group data generation and the accuracy reached by the cell codes is compiled in tables. To assess the quality of the data, comparisons have been made between measured and calculated reactor parameters. The information available does not show significant differences concerning the quality of the data libraries. (orig.) [de

  8. Activity report on the utilization of research reactors (JRR-3 and JRR-4). Japanese fiscal year, 2009

    International Nuclear Information System (INIS)

    2014-02-01

    JRR-3 is used for the purposes below; Experimental studies such as neutron scattering, prompt gamma-ray analyses, neutron radiography, Irradiation for activation analyses, radioisotope (RI) productions, fission tracks, Irradiation test of reactor materials, etc. JRR-4 is used for the purposes below; Medical irradiation (Boron Neutron Capture Therapy : BNCT), Prompt gamma-ray analyses, Sensitivity measurement of radiation detectors, Experiment in the nuclear reactor training, Practice of Reactor operation, Irradiation for activation analyses, RI productions, fission tracks, etc. In the fiscal year 2009, The research reactor JRR-3 was operated 7 cycles (cycle operation : 26days/cycle) for utilization sharing of the facility. And JRR-4 was operated 6 cycles (daily operation : 24 days). The volume contains 138 activity reports, which are categorized into the fields of neutron scattering (11 subcategories), neutron radiography, prompt gamma-ray analyses, neutron activation analyses, RI productions, and others submitted by the users in JAEA and from other organizations. (author)

  9. Activity report on the utilization of research reactors (JRR-3 and JRR-4). Japanese fiscal year, 2005

    International Nuclear Information System (INIS)

    2007-03-01

    In the fiscal year 2005, The research reactor JRR-3 was operated 7 cycles (cycle operation : 26days/cycle) for utilization sharing of the facility. And JRR-4 was operated 37 cycles (daily operation : 137 days). JRR-3 is used for the purposes below; Experimental studies such as neutron scattering, prompt gamma-ray analyses, neutron radiography. Irradiation for activation analyses, radioisotope (RI) productions, fission tracks. Irradiation test of reactor materials etc. JRR-4 is used for the purposes below; Medical irradiation (Boron Neutron Capture Therapy : BNCT). Prompt gamma-ray analyses. Sensitivity measurement of radiation detectors. Experiment in the nuclear reactor training. Practice of Reactor operation. Irradiation for activation analyses, RI productions, fission tracks etc. The volume contains 100 activity reports, which are categorized into the fields of neutron scattering (9 subcategories), neutron radiography, neutron activation analyses, RI productions, prompt gamma-ray analyses, and others submitted by the users in JAEA and from other organizations. (author)

  10. Activity report on the utilization of research reactors (JRR-3 and JRR-4). Japanese fiscal year, 2006

    International Nuclear Information System (INIS)

    2009-01-01

    In the fiscal year 2006, the research reactor JRR-3 was operated 7 cycles (cycle operation: 26 days/cycle) for utilization sharing of the facility. And JRR-4 was operated 37 cycles (daily operation: 151 days). JRR-3 is used for the purposes below; Experimental studies such as neutron scattering, prompt gamma-ray analyses, neutron radiography, Irradiation for activation analyses, radioisotope (RI) productions, fission tracks, Irradiation test of reactor materials, etc. JRR-4 is used for the purposes below; Medical irradiation (Boron Neutron Capture Therapy : BNCT), Prompt gamma-ray analyses, Sensitivity measurement of radiation detectors, Experiment in the nuclear reactor training, Practice of Reactor operation, Irradiation for activation analyses, RI productions, fission tracks, etc. The volume contains 294 activity reports, which are categorized into the fields of neutron scattering (11 subcategories), neutron radiography, neutron activation analyses, RI productions, prompt gamma-ray analyses, and others submitted by the users in JAEA and from other organizations. (author)

  11. Activity report on the utilization of research reactors (JRR-3 and JRR-4). Japanese fiscal year, 2009

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2014-02-15

    JRR-3 is used for the purposes below; Experimental studies such as neutron scattering, prompt gamma-ray analyses, neutron radiography, Irradiation for activation analyses, radioisotope (RI) productions, fission tracks, Irradiation test of reactor materials, etc. JRR-4 is used for the purposes below; Medical irradiation (Boron Neutron Capture Therapy : BNCT), Prompt gamma-ray analyses, Sensitivity measurement of radiation detectors, Experiment in the nuclear reactor training, Practice of Reactor operation, Irradiation for activation analyses, RI productions, fission tracks, etc. In the fiscal year 2009, The research reactor JRR-3 was operated 7 cycles (cycle operation : 26days/cycle) for utilization sharing of the facility. And JRR-4 was operated 6 cycles (daily operation : 24 days). The volume contains 138 activity reports, which are categorized into the fields of neutron scattering (11 subcategories), neutron radiography, prompt gamma-ray analyses, neutron activation analyses, RI productions, and others submitted by the users in JAEA and from other organizations. (author)

  12. 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)

  13. Qualification of the nuclear reactor core model DYN3D coupled to the thermohydraulic system code ATHLET, applied as an advanced tool for accident analysis of VVER-type reactors. Final report

    International Nuclear Information System (INIS)

    Grundmann, U.; Kliem, S.; Krepper, E.; Mittag, S; Rohde, U.; Schaefer, F.; Seidel, A.

    1998-03-01

    The nuclear reactor core model DYN3D with 3D neutron kinetics has been coupled to the thermohydraulic system code ATHLET. In the report, activities on qualification of the coupled code complex ATHLET-DYN3D as a validated tool for the accident analysis of russian VVER type reactors are described. That includes: - Contributions to the validation of the single codes ATHLET and DYN3D by the analysis of experiments on natural circulation behaviour in thermohydraulic test facilities and solution of benchmark tasks on reactivity initiated transients, - the acquisition and evaluation of measurement data on transients in nuclear power plants, the validation of ATHLET-DYN3D by calculating an accident with delayed scram and a pump trip in VVER plants, - the complementary improvement of the code DYN3D by extension of the neutron physical data base, implementation of an improved coolant mixing model, consideration of decay heat release and xenon transients, - the analysis of steam leak scenarios for VVER-440 type reactors with failure of different safety systems, investigation of different model options. The analyses showed, that with realistic coolant mixing modelling in the downcomer and the lower plenum, recriticality of the scramed reactor due to overcooling can be reached. The application of the code complex ATHLET-DYN3D in Czech Republic, Bulgaria and the Ukraine has been started. Future work comprises the verification of ATHLET-DYN3D with a DYN3D version for the square fuel element geometry of western PWR. (orig.) [de

  14. 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

  15. Neutron source for a reactor

    International Nuclear Information System (INIS)

    Kobayashi, Hiromasa.

    1975-01-01

    Object: To easily increase a start-up power of a reactor without irradiation in other reactors. Structure: A neutron source comprises Cf 252 , a natural antimony rod, a layer of beryllium, and a vessel of neutron source. On upper and lower portion of Cf 252 are arranged natural antimony rods, which are surrounded by the Be layer, the entirety being charged into the vessel. The Cf 252 may emit neutron, has a half life more than a period of operating cycle of the reactor and is less deteriorated even irradiated by radioactive rays while being left within the reactor. The natural antimony rod is radioactivated by neutron from Cf 252 and neutron as reactor power increases to emit γ rays. The Be absorbs γ rays to emit the neutron. The antimony rod is irradiated within the reactor. Further, since the Cf 252 is small in neutron absorption cross section, it is hard to be deteriorated even while being inserted within the reactor. (Kamimura, M.)

  16. Processing requirements for property optimization of Eu2O3-W cermets for fast reactor neutron absorber applications

    International Nuclear Information System (INIS)

    Pasto, A.E.; Tennery, V.J.

    1977-01-01

    Europium sesquioxide is a candidate fast reactor neutron absorber material. It possesses several desirable characteristics for this application, but has a low thermal conductivity. This gives rise to pellet cracking during reactor operation. To increase the thermal conductivity without great sacrifice in nuclear worth, addition of tungsten to Eu 2 O 3 has been evaluated. Synthesis and fabrication techniques described allow preparation of high density compacts of Eu 2 O 3 -15 vol. percent tungsten, possessing favorable thermal conductivity and thermal expansion characteristics

  17. Convective cooling in a pool-type research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sipaun, Susan, E-mail: susan@nm.gov.my [Malaysian Nuclear Agency, Industrial Technology Division, Blok 29T, Bangi 43200, Selangor (Malaysia); Usman, Shoaib, E-mail: usmans@mst.edu [Missouri University of Science and Technology, Nuclear Engineering, 222 Fulton Hall 301 W.14th St., Rolla 64509 MO (United States)

    2016-01-22

    A reactor produces heat arising from fission reactions in the nuclear core. In the Missouri University of Science and Technology research reactor (MSTR), this heat is removed by natural convection where the coolant/moderator is demineralised water. Heat energy is transferred from the core into the coolant, and the heated water eventually evaporates from the open pool surface. A secondary cooling system was installed to actively remove excess heat arising from prolonged reactor operations. The nuclear core consists of uranium silicide aluminium dispersion fuel (U{sub 3}Si{sub 2}Al) in the form of rectangular plates. Gaps between the plates allow coolant to pass through and carry away heat. A study was carried out to map out heat flow as well as to predict the system’s performance via STAR-CCM+ simulation. The core was approximated as porous media with porosity of 0.7027. The reactor is rated 200kW and total heat density is approximately 1.07+E7 Wm{sup −3}. An MSTR model consisting of 20% of MSTR’s nuclear core in a third of the reactor pool was developed. At 35% pump capacity, the simulation results for the MSTR model showed that water is drawn out of the pool at a rate 1.28 kg s{sup −1} from the 4” pipe, and predicted pool surface temperature not exceeding 30°C.

  18. CANDU reactors with reactor grade plutonium/thorium carbide fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sahin, Suemer [Atilim Univ., Ankara (Turkey). Faculty of Engineering; Khan, Mohammed Javed; Ahmed, Rizwan [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan); Gazi Univ., Ankara (Turkey). Faculty of Technology

    2011-08-15

    Reactor grade (RG) plutonium, accumulated as nuclear waste of commercial reactors can be re-utilized in CANDU reactors. TRISO type fuel can withstand very high fuel burn ups. On the other hand, carbide fuel would have higher neutronic and thermal performance than oxide fuel. In the present work, RG-PuC/ThC TRISO fuels particles are imbedded body-centered cubic (BCC) in a graphite matrix with a volume fraction of 60%. The fuel compacts conform to the dimensions of sintered CANDU fuel compacts are inserted in 37 zircolay rods to build the fuel zone of a bundle. Investigations have been conducted on a conventional CANDU reactor based on GENTILLYII design with 380 fuel bundles in the core. Three mixed fuel composition have been selected for numerical calculation; (1) 10% RG-PuC + 90% ThC; (2) 30% RG-PuC + 70% ThC; (3) 50% RG-PuC + 50% ThC. Initial reactor criticality values for the modes (1), (2) and (3) are calculated as k{sub {infinity}}{sub ,0} = 1.4848, 1.5756 and 1.627, respectively. Corresponding operation lifetimes are {proportional_to} 2.7, 8.4, and 15 years and with burn ups of {proportional_to} 72 000, 222 000 and 366 000 MW.d/tonne, respectively. Higher initial plutonium charge leads to higher burn ups and longer operation periods. In the course of reactor operation, most of the plutonium will be incinerated. At the end of life, remnants of plutonium isotopes would survive; and few amounts of uranium, americium and curium isotopes would be produced. (orig.)

  19. Challenges in licensing a sodium-cooled advanced recycling reactor

    International Nuclear Information System (INIS)

    Levin, Alan E.

    2008-01-01

    As part of the Global Nuclear Energy Partnership (GNEP), the U.S. Department of Energy (DOE) has focused on the use of sodium-cooled fast reactors (SFRs) for the destruction of minor actinides derived from used reactor fuel. This approach engenders an array of challenges with respect to the licensing of the reactor: the U.S. Nuclear Regulatory Commission (NRC) has never completed the review of an application for an operating license for a sodium-cooled reactor. Moreover, the current U.S. regulatory structure has been developed to deal almost exclusively with light-water reactor (LWR) designs. Consequently, the NRC must either (1) develop a new regulatory process for SFRs, or (2) reinterpret the existing regulations to apply them, as appropriate, to SFR designs. During the 1980s and 1990s, the NRC conducted preliminary safety assessments of the Sodium Advanced Fast Reactor (SAFR) and the Power Reactor Innovative Small Module (PRISM) designs, and in that context, began to consider how to apply LWR-based regulations to SFR designs. This paper builds on that work to consider the challenges, from the reactor designer's point of view, associated with licensing an SFR today, considering (1) the evolution of SFR designs, (2) the particular requirements of reactor designs to meet GNEP objectives, and (3) the evolution of NRC regulations since the conclusion of the SAFR and PRISM reviews. (author)

  20. Atomization of U3Si2 for research reactor fuel

    International Nuclear Information System (INIS)

    Kim, C.K.; Kim, K.H.; Lee, C.T.; Kuk, I.H.

    1995-01-01

    Rotating disk atomization technique is applied to KMRR (Korea Multi-purpose Research Reactor) fuel fabrication. A rotating disk atomizer is designed and manufactured locally and U-4.0 wt. % Si alloy powders are produced. The atomized powders are heat-treated to transform into U 3 Si and the mixture of U 3 Si and Al are extruded to fuel meat. Most of the atomized powders are spherical in shape. The microstructure of the powder is fine due to the rapid solidification. The time required for peritectoid reaction is reduced due to the fine microstructures and the resultant U 3 Si grain size is finer than ever obtained from ingot process. The mechanical properties of the fuel meat are improved: yield strength about 30 %, tensile strength 10% and elongation 250 % increased. (author)

  1. Development of multi-physics code systems based on the reactor dynamics code DYN3D

    Energy Technology Data Exchange (ETDEWEB)

    Kliem, Soeren; Gommlich, Andre; Grahn, Alexander; Rohde, Ulrich [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany); Schuetze, Jochen [ANSYS Germany GmbH, Darmstadt (Germany); Frank, Thomas [ANSYS Germany GmbH, Otterfing (Germany); Gomez Torres, Armando M.; Sanchez Espinoza, Victor Hugo [Karlsruher Institut fuer Technologie (KIT), Eggenstein-Leopoldshafen (Germany)

    2011-07-15

    The reactor dynamics code DYN3D has been coupled with the CFD code ANSYS CFX and the 3D thermal hydraulic core model FLICA4. In the coupling with ANSYS CFX, DYN3D calculates the neutron kinetics and the fuel behavior including the heat transfer to the coolant. The physical data interface between the codes is the volumetric heat release rate into the coolant. In the coupling with FLICA4 only the neutron kinetics module of DYN3D is used. Fluid dynamics and related transport phenomena in the reactor's coolant and fuel behavior is calculated by FLICA4. The correctness of the coupling of DYN3D with both thermal hydraulic codes was verified by the calculation of different test problems. These test problems were set-up in such a way that comparison with the DYN3D stand-alone code was possible. This included steady-state and transient calculations of a mini-core consisting of nine real-size PWR fuel assemblies with ANSYS CFX/DYN3D as well as mini-core and a full core steady-state calculation using FLICA4/DYN3D. (orig.)

  2. Development of multi-physics code systems based on the reactor dynamics code DYN3D

    International Nuclear Information System (INIS)

    Kliem, Soeren; Gommlich, Andre; Grahn, Alexander; Rohde, Ulrich; Schuetze, Jochen; Frank, Thomas; Gomez Torres, Armando M.; Sanchez Espinoza, Victor Hugo

    2011-01-01

    The reactor dynamics code DYN3D has been coupled with the CFD code ANSYS CFX and the 3D thermal hydraulic core model FLICA4. In the coupling with ANSYS CFX, DYN3D calculates the neutron kinetics and the fuel behavior including the heat transfer to the coolant. The physical data interface between the codes is the volumetric heat release rate into the coolant. In the coupling with FLICA4 only the neutron kinetics module of DYN3D is used. Fluid dynamics and related transport phenomena in the reactor's coolant and fuel behavior is calculated by FLICA4. The correctness of the coupling of DYN3D with both thermal hydraulic codes was verified by the calculation of different test problems. These test problems were set-up in such a way that comparison with the DYN3D stand-alone code was possible. This included steady-state and transient calculations of a mini-core consisting of nine real-size PWR fuel assemblies with ANSYS CFX/DYN3D as well as mini-core and a full core steady-state calculation using FLICA4/DYN3D. (orig.)

  3. Effect of kinetic parameters on simultaneous ramp reactivity insertion plus beam tube flooding accident in a typical low enriched U{sub 3}Si{sub 2}-Al fuel-based material testing reactor-type research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nasir, Rubina; Mirza, Nasir M. [Dept. of, Physics, Air University, Islamabad (Pakistan); Mirza, Sikander M. [Dept. of, Physics and Applied Mathematics, Pakistan Institute of Engineering and Applied Sciences, Post Office Nilore, Islamabad (Pakistan)

    2017-06-15

    This work looks at the effect of changes in kinetic parameters on simultaneous reactivity insertions and beam tube flooding in a typical material testing reactor-type research reactor with low enriched high density (U{sub 3}Si{sub 2}-Al) fuel. Using a modified PARET code, various ramp reactivity insertions (from $0.1/0.5 s to $1.3/0.5 s) plus beam tube flooding ($0.5/0.25 s) accidents under uncontrolled conditions were analyzed to find their effects on peak power, net reactivity, and temperature. Then, the effects of changes in kinetic parameters including the Doppler coefficient, prompt neutron lifetime, and delayed neutron fractions on simultaneous reactivity insertion and beam tube flooding accidents were analyzed. Results show that the power peak values are significantly sensitive to the Doppler coefficient of the system in coupled accidents. The material testing reactor-type system under such a coupled accident is not very sensitive to changes in the prompt neutron life time; the core under such a coupled transient is not very sensitive to changes in the effective delayed neutron fraction.

  4. Development of a core follow calculational system for research reactors

    International Nuclear Information System (INIS)

    Muller, E.Z.; Ball, G.; Joubert, W.R.; Schutte, H.C.; Stoker, C.C.; Reitsma, F.

    1994-01-01

    Over the last few years a comprehensive Pressurized Water Reactor and Materials Testing Reactor core analysis code system based on modern reactor physics methods has been under development by the Atomic Energy Corporation of South Africa. This system, known as OSCAR-3, will incorporate a customized graphical user interface and data management system to ensure user-friendliness and good quality control. The system has now reached the stage of development where it can be used for practical MTR core analyses. This paper describes the current capabilities of the components of the OSCAR-3 package, their integration within the package, and outlines future developments. 10 refs., 1 tab., 1 fig

  5. 3 MW TRIGA Research Reactor facility of BAEC and its Utilization

    International Nuclear Information System (INIS)

    Molla, N.I.; Bhuiyan, S.I.; Wadud Mondal, M.A.; Ahmed, F.U.; Islam, M.N.; Hossain, S.M.; Ahmed, K.; Zulquarnain, A.; Abedin, Z.

    1999-01-01

    The paper briefly describes the Utilisation of 3 MW TRIGA Research Reactor of BAEC for neutron beam research, neutron activation analysis are isotope production. It includes the installation of the triple axis neutron spectrometer at the radial piercing beam port and a neutron radiography set-up at the tangential beam port and their uses for material analysis and condensed matter research and material testing. Nuclear and magnetic structures of some ferrites have been studied in powder diffraction method in the double axis mode. SANS technique with double crystal diffraction known as Bonse and Hart's method has been adopted in an experiment with alumina sample. The neutron radiography set-up and its use in the detection of corrosion in alumina have been reported. Determination of arsenic concentration in drinking water from tube well via Instrumental Neutron Activation Analysis and production of radioiodine-131 by dry distillation method are presented. Our experience on the removal of N-16 decay tank because of the leakage of coolant and bringing the research reactor back to operational by-passing the decay tank have been focussed. A possible reconfiguration of the existing TRIGA core, without exceeding the safety margins, providing additional irradiation channel and upgrading the neutron flux for increased radioisotope production has been attempted. Cross section library ENDF/B-VI and JENDL3.2, code NJOY94.10, WIMSD package, 3-D code CITATION, PARET and Monte Carlo code MCNP4B2 have been employed to achieve the objective. (author)

  6. Power measurement of the RA-3 reactor using the neutron noise technique and {sup 16}N; Medicion de la potencia del reactor RA-3, mediante la tecnica de ruido neutronico y nitrogeno 16

    Energy Technology Data Exchange (ETDEWEB)

    Gomez, Angel [Comision Nacional de Energia Atomica, General San Martin (Argentina). Dept. Reactores y Centrales Nucleares

    2003-07-01

    This work describes a measurement method based on the neutron noise technique which is used for determining the relation between the power and the currents of two ionization chambers. These chambers are sensitive to the gamma radiation from the {sup 16}N decay produced in the RA-3 reactor core. The power during operation is obtained from the calibration factors by measuring those currents. As this calibration factors depend on the cooler flow that circulates in the reactor core and in the {sup 16}N measuring system, an estimator, that is a function of the ratio of this currents, is proposed in order to detect flow changes. (author)

  7. Pilot plant production at Riso of LEU silicide fuel for the Danish reactor DR3

    International Nuclear Information System (INIS)

    Toft, P.; Borring, J.; Adolph, E.

    1988-01-01

    A pilot plant for fabricating LEU silicide fuel elements has been established at Riso National Laboratory. Three test elements for the Danish reactor DR3 have been fabricated, based on 19.88% enriched U 3 Si 2 powder that has been purchased elsewhere. The pilot plant has been set up and 3 test elements fabricated without any major difficulties

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

    International Nuclear Information System (INIS)

    Zecevic, V.

    1963-12-01

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

  9. MOX recycling in GEN 3 + EPR Reactor homogeneous and stable full MOX core

    Energy Technology Data Exchange (ETDEWEB)

    Arslan, M.; Villele, E. de; Gauthier, J.C.; Marincic, A. [AREVA - Tour AREVA, 1 Place Jean Millier, 92084 Paris La Defense (France)

    2013-07-01

    In the case of the EPR (European Pressurized Reactor) reactor, 100% MOX core management is possible with simple design adaptations which are not significantly costly. 100% MOX core management offers several highly attractive advantages. First, it is possible to have the same plutonium content in all the rods of a fuel assembly instead of having rods with 3 different plutonium contents, as in MOX assemblies in current PWRs. Secondly, the full MOX core is more homogeneous. Thirdly, the stability of the core is significantly increased due to a large reduction in the Xe effect. Fourthly, there is a potential for the performance of the MOX fuel to match that of new high performance UO{sub 2} fuel (enrichment up to 4.95 %) in terms of increased burn up and cycle length. Fifthly, since there is only one plutonium content, the manufacturing costs are reduced. Sixthly, there is an increase in the operating margins of the reactor, and in the safety margins in accident conditions. The use of 100% MOX core will improve both utilisation of natural uranium resources and reductions in high level radioactive waste inventory.

  10. MOX recycling in GEN 3 + EPR Reactor homogeneous and stable full MOX core

    International Nuclear Information System (INIS)

    Arslan, M.; Villele, E. de; Gauthier, J.C.; Marincic, A.

    2013-01-01

    In the case of the EPR (European Pressurized Reactor) reactor, 100% MOX core management is possible with simple design adaptations which are not significantly costly. 100% MOX core management offers several highly attractive advantages. First, it is possible to have the same plutonium content in all the rods of a fuel assembly instead of having rods with 3 different plutonium contents, as in MOX assemblies in current PWRs. Secondly, the full MOX core is more homogeneous. Thirdly, the stability of the core is significantly increased due to a large reduction in the Xe effect. Fourthly, there is a potential for the performance of the MOX fuel to match that of new high performance UO 2 fuel (enrichment up to 4.95 %) in terms of increased burn up and cycle length. Fifthly, since there is only one plutonium content, the manufacturing costs are reduced. Sixthly, there is an increase in the operating margins of the reactor, and in the safety margins in accident conditions. The use of 100% MOX core will improve both utilisation of natural uranium resources and reductions in high level radioactive waste inventory

  11. RA Reactor applications, Annex A

    International Nuclear Information System (INIS)

    Cupac, S.; Vukadin, Z.

    2000-01-01

    Full text: In 2000 Ra reactor was not operated. New instrumentation is not complete, without it, it is not possible to think about reactor start-up. Since 1985, when reactor operation was forbidden, there are 480 fuel elements left in 48 fuel channels in the reactor core. Heavy water was removed from the reactor core because of the repair of the heavy water pumps in 1986. The old instrumentation was removed. Eleven years after being left to its own destiny, it would be difficult to imagine that anybody would think of reactor restart without examining the state of reactor vessel and other vital reactor components. Maintaining the reactor under existing conditions without final decision about restart or permanent shutdown is destructive for this nuclear facility. The existing state that pertains for more than 10 years would have only one result, destruction of the RA reactor [sr

  12. RA Reactor applications, Annex A

    International Nuclear Information System (INIS)

    Cupac, S.; Vukadin, Z.

    1998-01-01

    Full text: In 1998 Ra reactor was not operated. New instrumentation is not complete, without it, it is not possible to think about reactor start-up. Since 1985, when reactor operation was forbidden, there are 480 fuel elements left in 48 fuel channels in the reactor core. Heavy water was removed from the reactor core because of the repair of the heavy water pumps in 1986. The old instrumentation was removed. Eleven years after being left to its own destiny, it would be difficult to imagine that anybody would think of reactor restart without examining the state of reactor vessel and other vital reactor components. Maintaining the reactor under existing conditions without final decision about restart or permanent shutdown is destructive for this nuclear facility. The existing state that pertains for more than 10 years would have only one result, destruction of the RA reactor [sr

  13. RA Reactor applications, Annex A

    International Nuclear Information System (INIS)

    Cupac, S.; Vukadin, Z.

    1996-01-01

    Full text: In 2000 Ra reactor was not operated. New instrumentation is not complete, without it, it is not possible to think about reactor start-up. Since 1985, when reactor operation was forbidden, there are 480 fuel elements left in 48 fuel channels in the reactor core. Heavy water was removed from the reactor core because of the repair of the heavy water pumps in 1986. The old instrumentation was removed. Eleven years after being left to its own destiny, it would be difficult to imagine that anybody would think of reactor restart without examining the state of reactor vessel and other vital reactor components. Maintaining the reactor under existing conditions without final decision about restart or permanent shutdown is destructive for this nuclear facility. The existing state that pertains for more than 10 years would have only one result, destruction of the RA reactor [sr

  14. An internally illuminated monolith reactor: Pros and cons relative to a slurry reactor

    NARCIS (Netherlands)

    Carneiro, Joana T.; Carneiro, J.T.; Berger, Rob; Moulijn, Jacob A.; Mul, Guido

    2009-01-01

    In the present study, kinetic models for the photo-oxidation of cyclohexane in two different photoreactor systems are discussed: a top illumination reactor (TIR) representative of a slurry reactor, and the so-called internally illuminated monolith reactor (IIMR) representing a reactor containing

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

    International Nuclear Information System (INIS)

    Powell, J.R.

    1977-01-01

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

  16. Cost analysis and financial risk profile for severe reactor accidents at Waterford-3

    International Nuclear Information System (INIS)

    Cutbush, J.D.; Abbott, E.C.; Carpenter, W.L. Jr.

    1992-01-01

    To support Louisiana Power and Light Company (LP and L) in determining an appropriate level of nuclear property insurance for Waterford Steam Electric Station, Unit 3 (Waterford-3), ABZ, Incorporated, performed a series of cost analyses and developed a financial risk profile. This five-month study, conducted in 1991, identified the potential Waterford-3 severe reactor accidents and described each from a cleanup perspective, estimated the cost and schedule to cleanup from each accident, developed a probability distribution of associated financial exposure, and developed a profile of financial risk as a function of insurance coverage

  17. 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.)

  18. Reactor building for a nuclear reactor

    International Nuclear Information System (INIS)

    Haidlen, F.

    1976-01-01

    The invention concerns the improvement of the design of a liner, supported by a latticed steel girder structure and destined for guaranteeing a gastight closure for the plant compartments in the reactor building of a pressurized water reactor. It is intended to provide the steel girder structure on their top side with grates, being suited for walking upon, and to hang on their lower side diaphragms in modular construction as a liner. At the edges they may be sealed with bellows in order to avoid thermal stresses. The steel girder structure may at the same time serve as supports for parts of the steam pipe. (RW) [de

  19. Neurexin-1β Binding to Neuroligin-1 Triggers the Preferential Recruitment of PSD-95 versus Gephyrin through Tyrosine Phosphorylation of Neuroligin-1

    Directory of Open Access Journals (Sweden)

    Grégory Giannone

    2013-06-01

    Full Text Available Adhesion between neurexin-1β (Nrx1β and neuroligin-1 (Nlg1 induces early recruitment of the postsynaptic density protein 95 (PSD-95 scaffold; however, the associated signaling mechanisms are unknown. To dissociate the effects of ligand binding and receptor multimerization, we compared conditions in which Nlg1 in neurons was bound to Nrx1β or nonactivating HA antibodies. Time-lapse imaging, fluorescence recovery after photobleaching, and single-particle tracking demonstrated that in addition to aggregating Nlg1, Nrx1β binding stimulates the interaction between Nlg1 and PSD-95. Phosphotyrosine immunoblots and pull-down of gephyrin by Nlg1 peptides in vitro showed that Nlg1 can be phosphorylated at a unique tyrosine (Y782, preventing gephyrin binding. Expression of Nlg1 point mutants in neurons indicated that Y782 phosphorylation controls the preferential binding of Nlg1 to PSD-95 versus gephyrin, and accordingly the formation of inhibitory and excitatory synapses. We propose that ligand-induced changes in the Nlg1 phosphotyrosine level control the balance between excitatory and inhibitory scaffold assembly during synapse formation and stabilization.

  20. 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

  1. Examination policy concerning the additional installation of No. 3 and No. 4 reactors in Takahama Nuclear Power Station and No. 3 and No. 4 reactors in Fukushima No. 2 Nuclear Power Station

    International Nuclear Information System (INIS)

    1980-01-01

    The Nuclear Safety Commission decided the annual examination policy on the modification of reactor installation in Takahama Nuclear Power Station to construct No. 3 and No. 4 reactors inquired under date of November 26, 1979, by the Minister of International Trade and Industry, so that the examination results of the accident in Three Mile Island nuclear power station are reflected to the examination for the purpose of improving reactor safety. The examination results of the accident in Three Mile Island power station are being investigated by the Committee on Examination of Reactor Safety, based on the policy shown in ''On the second report of the special committee examining the accident in a nuclear power station in the U.S.'' determined by the Nuclear Safety Commission under date of September 13, 1979. Though the Committee will further clarify the past guideline about the items concerning the criteria, design and operation management, the Committee decided the tentative policy to reflect it to safety examination. Further, a table is attached, in which 52 items to be reflected to the security measures are classified from the viewpoint of necessity to reflect them to the final examination. This table includes 13 items of criteria and examination, 7 items related to design, 10 items related to operation management, 10 antidisaster items, and 12 items related to safety research. (Wakatsuki, Y.)

  2. 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)

  3. 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)

  4. Calculation of fundamental parameters for the dynamical study of TRIGA-3-Salazar reactor (Mixed reactor core)

    International Nuclear Information System (INIS)

    Viais J, J.

    1994-01-01

    Kinetic parameters for dynamic study of two different configurations, 8 and 9, both with standard fuel, 20% enrichment and Flip (Fuel Life Improvement Program with 70% enrichment) fuel, for TRIGA Mark-III reactor from Mexico Nuclear Center, are obtained. A calculation method using both WIMS-D4 and DTF-IV and DAC1 was established, to decide which of those two configurations has the best safety and operational conditions. Validation of this methodology is done by calculate those parameters for a reactor core with new standard fuel. Configuration 9 is recommended to be use. (Author)

  5. Physics analysis of the Apollo D-3He tokamak reactor

    International Nuclear Information System (INIS)

    Santarius, J.F.; Emmert, G.A.

    1990-01-01

    Recent developments in the analysis and conceptual design of Apollo, a D- 3 He Tokamak Reactor are presented. Encouraging experimental results on TEXT motivated a key change in the Apollo concept utilization of an ergodic magnetic limiter for impurity control instead of a divertor. Parameters for the updated Apollo design and an analysis of the ergoidc magnetic limiter are given. The Apollo reference case uses direct conversion of synchrotron radiation to electricity by rectifying antennas (rectennas) for its power conversion system. Previous analyses of this concept are expanded, including further details of the rectennas and of the loss of synchrotron power to the waveguides and walls. Although Apollo will burn D- 3 He fuel, a significant amount of unburned tritium will be generated by D4D reactions. The possibility of operating a short, dedicated, T+ 3 He burn phase to eliminate this tritium will be examined

  6. Revision of fast reactor group constant set JFS-3-J2

    International Nuclear Information System (INIS)

    Takano, Hideki; Kaneko, Kunio.

    1989-10-01

    To improve the fast reactor group constant set JFS-3-J2 to be applicable for high burnup reactor calculations, group constants for 155 fission product nuclides and the lumped group cross sections for four mother fission isotopes of U-235, U-238, Pu-239 and Pu-241 have been generated. Furthermore, the group constants for higher actinides such as Am and Cm have been produced on the basis of the JENDL-2 nuclear data, so as to be able to use for TRU-transmutation calculations. Benchmark test of this revised set has been performed by analysing the 21 fast critical experimental assemblies. Benchmark calculation system based on one-dimensional Sn-method has been developed to investigate the accuracy of one-dimensional diffusion calculations. Significant difference between the results obtained with the diffusion and transport calculations was observed for small cores and the assemblies with iron or nickel reflector. (author)

  7. Preliminary design of a tandem mirror reactor

    International Nuclear Information System (INIS)

    Strohmayer, J.N.

    1984-04-01

    The purpose of this thesis is to examine the TARA mirror experiment as a possible tandem mirror reactor configuration. This is a preliminary study to size the coil structure based on using the smallest end cell axial length that physics and engineering allow, zeroing the central cell parallel currents and having interchange stability. The input powers are estimated for the final reactor design so a Q value may be estimated. The Q value is defined as the fusion power divided by the total injected power absorbed by the plasma. A computer study was performed on the effect of the transition size, the transition vertical spacing and transition current. These parameters affect the central cell parallel currents, the recircularization of the flux tube and the ratio of central cell beta to anchor beta needed for marginal stability. Two designs were identified. The first uses 100 keV and 13 keV neutral beams to pump the ions that trap in the thermal barrier. The Q value of this reactor is 11.3. The second reactor uses a pump beam at 40 keV. This energy is chosen because there is a resonance for the charge exchange cross section between D 0 and He 2+ at this energy, thus the alpha ash will be pumped along with the deuterium and tritium. The Q value of this reactor is 11.6

  8. Study of a new automatic reactor power control for the TRIGA Mark II reactor at University of Pavia

    Energy Technology Data Exchange (ETDEWEB)

    Borio Di Tigliole, A.; Magrotti, G. [Laboratorio Energia Nucleare Applicata (L.E.N.A.), University of Pavia, Via Aselli 41, 27100 (Italy); Cammi, A.; Memoli, V. [Politecnico di Milano, Department of Energy, Nuclear Engineering Division (CeSNEF), Via Ponzio 34/3, 20133 Milano (Italy); Gadan, M. A. [Instrumentation and Control Department, National Atomic Energy Comission of Argentina, University of Pavia (Italy)

    2009-07-01

    The installation of a new Instrumentation and Control (IC) system for the TRIGA Mark-II reactor at University of Pavia has recently been completed in order to assure a safe and continuous reactor operation for the future. The intervention involved nearly the whole IC system and required a channel-by-channel component substitution. One of the most sensitive part of the intervention concerned the Automatic Reactor Power Controller (ARPC) which permits to keep the reactor at an operator-selected power level acting on the control rod devoted to the fine regulation of system reactivity. This controller installed can be set up using different control logics: currently the system is working in relay mode. The main goal of the work presented in this paper is to set up a Proportional-Integral-Derivative (PID) configuration of the new controller installed on the TRIGA reactor of Pavia so as to optimize the response to system perturbations. The analysis have shown that a continuous PID offers generally better results than the relay mode which causes power oscillations with an amplitude of 3% of the nominal power

  9. Draft environmental impact statement for the siting, construction, and operation of New Production Reactor capacity. Volume 3, Sections 7-12, Appendices A-C

    Energy Technology Data Exchange (ETDEWEB)

    1991-04-01

    This Environmental Impact Statement (EIS) assesses the potential environmental impacts, both on a broad programmatic level and on a project-specific level, concerning a proposed action to provide new tritium production capacity to meet the nation`s nuclear defense requirements well into the 21st century. A capacity equivalent to that of about a 3,000-megawatt (thermal) heavy-water reactor was assumed as a reference basis for analysis in this EIS; this is the approximate capacity of the existing production reactors at DOE`s Savannah River Site near Aiken, South Carolina. The EIS programmatic alternatives address Departmental decisions to be made on whether to build new production facilities, whether to build one or more complexes, what size production capacity to provide, and when to provide this capacity. Project-specific impacts for siting, constructing, and operating new production reactor capacity are assessed for three alternative sites: the Hanford Site near Richland, Washington; the Idaho National Engineering Laboratory near Idaho Falls, Idaho; and the Savannah River Site. For each site, the impacts of three reactor technologies (and supporting facilities) are assessed: a heavy-water reactor, a light-water reactor, and a modular high-temperature gas-cooled reactor. Impacts of the no-action alternative also are assessed. The EIS evaluates impacts related to air quality; noise levels; surface water, groundwater, and wetlands; land use; recreation; visual environment; biotic resources; historical, archaeological, and cultural resources; socioeconomics; transportation; waste management; and human health and safety. The EIS describes in detail the potential radioactive releases from new production reactors and support facilities and assesses the potential doses to workers and the general public. This volume contains references; a list of preparers and recipients; acronyms, abbreviations, and units of measure; a glossary; an index and three appendices.

  10. Repairing liner of the reactor; Reparacion del liner del reactor

    Energy Technology Data Exchange (ETDEWEB)

    Aguilar H, F [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2001-07-15

    Due to the corrosion problems of the aluminum coating of the reactor pool, a periodic inspections program by ultrasound to evaluate the advance grade and the corrosion speed was settled down. This inspections have shown the necessity to repair some areas, in those that the slimming is significant, of not making it can arrive to the water escape of the reactor pool. The objective of the repair is to place patches of plates of 1/4 inch aluminum thickness in the areas of the reactor 'liner', in those that it has been detected by ultrasound a smaller thickness or similar to 3 mm. To carry out this the fuels are move (of the core and those that are decaying) to a temporary storage, the structure of the core is confined in a tank that this placed inside the pool of the reactor, a shield is placed in the thermal column and it is completely extracted the water for to leave uncover the 'liner' of the reactor. (Author)

  11. Effect of Co3O4 and CeO2 Infiltration on the Activity of a LSM15/GDC10 Highly Porous Electrochemical Reactor

    DEFF Research Database (Denmark)

    Ippolito, Davide; Kammer Hansen, Kent

    2014-01-01

    The reduction of air pollution has become an international concern over the last ten years because of increases in emissions from mobile and stationary sources. Among these sources, volatile organic compounds (VOC) represent a serious environmental problem, together with NOx, SOx and particulate...... VOC component of Diesel engine exhausts, over a wide range of temperatures. The entire reactor was thought as a highly porous catalytic filter for a possible application in a Diesel exhausts purification system. The porous reactor was used as a backbone for the infiltration of Co3O4 and Co3O4/CeO2...

  12. The effect of heavy water reactors and liquid fuel reactors on the long-term development of nuclear energy

    International Nuclear Information System (INIS)

    Brand, P.; Wiechers, W.K.

    1974-01-01

    The effects of the rates at which various combinations of power reactor types are installed on the long-range (to the year 2040) uranium and plutonium inventory requirements are examined. Consideration is given to light water reactors, fast breeder reactors, high temperature gas-cooled reactors, heavy water reactors, and thermal breeder reactors, in various combinations, and assuming alternatively a 3% and a 5% growth in energy demand

  13. Annual report of Department of Research Reactor and Tandem Accelerator, JFY2007. Operation, utilization and technical development of JRR-3, JRR-4, NSRR and tandem accelerator

    International Nuclear Information System (INIS)

    Miyazaki, Osamu; Awa, Yasuaki; Isaka, Koji; Kutsukake, Kenichi; Komeda, Masao; Shibata, Ko; Hiyama, Kazuhisa; Suzuki, Mayu; Sone, Takuya; Ohuchi, Tomoaki; Terakado, Yuichi; Sataka, Masao

    2009-06-01

    The Department of Research Reactors and Tandem Accelerator is in charge of the operation, utilization and technical development of JRR-3(Japan Research Reactor-3), JRR-4(Japan Research Reactor-4), NSRR(Nuclear Safety Research Reactor) and Tandem Accelerator. This annual report describes a summary of activities of services and technical developments carried out in the period between April 1, 2007 and March 31, 2008. The activities were categorized into five service/development fields: (1) Operation and maintenance of research reactors and tandem accelerator. (2) Utilization of research reactors and tandem accelerator. (3) Upgrading of utilization techniques of research reactors and tandem accelerator. (4) Safety administration for research reactors and tandem accelerator. (5) International cooperation. Also contained are lists of publications, meetings, granted permissions on lows and regulations concerning atomic energy, commendation, plans and outcomes in service and technical developments and so on. (author)

  14. Annual report of Department of Research Reactor and Tandem Accelerator, JFY2010. Operation, utilization and technical development of JRR-3, JRR-4, NSRR and Tandem Accelerator

    International Nuclear Information System (INIS)

    Ishii, Tetsuro; Nakamura, Kiyoshi; Kawamata, Satoshi; Yamada, Yusuke; Kawashima, Kazuhiro; Asozu, Takuhiro; Nakamura, Takemi; Arai, Masaji; Yoshinari, Shuji; Sataka, Masao

    2012-03-01

    The Department of Research Reactors and Tandem Accelerator is in charge of the operation, utilization and technical development of JRR-3(Japan Research Reactor No.3), JRR-4(Japan Research Reactor No.4), NSRR(Nuclear Safety Research Reactor) and Tandem Accelerator. This annual report describes a summary of activities of services and technical developments carried out in the period between April 1, 2010 and March 31, 2011. The activities were categorized into five service/development fields: (1) Operation and maintenance of research reactors and tandem accelerator, (2) Utilization of research reactors and tandem accelerator, (3) Upgrading of utilization techniques of research reactors and tandem accelerator, (4) Safety administration for research reactors and tandem accelerator, (5) International cooperation. Also contained are lists of publications, meetings, granted permissions on lows and regulations concerning atomic energy, commendation, outcomes in service and technical developments and so on. (author)

  15. Verification of Remote Inspection Techniques for Reactor Internal Structures of Liquid Metal Reactor

    International Nuclear Information System (INIS)

    Joo, Young Sang; Lee, Jae Han

    2007-02-01

    The reactor internal structures and components of a liquid metal reactor (LMR) are submerged in hot sodium of reactor vessel. The division 3 of ASME code section XI specifies the visual inspection as major in-service inspection (ISI) methods of reactor internal structures and components. Reactor internals of LMR can not be visually examined due to opaque liquid sodium. The under-sodium viewing techniques using an ultrasonic wave should be applied for the visual inspection of reactor internals. Recently, an ultrasonic waveguide sensor with a strip plate has been developed for an application to the under-sodium inspection. In this study, visualization technique, ranging technique and monitoring technique have been suggested for the remote inspection of reactor internals by using the waveguide sensor. The feasibility of these remote inspection techniques using ultrasonic waveguide sensor has been evaluated by an experimental verification

  16. Verification of Remote Inspection Techniques for Reactor Internal Structures of Liquid Metal Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Joo, Young Sang; Lee, Jae Han

    2007-02-15

    The reactor internal structures and components of a liquid metal reactor (LMR) are submerged in hot sodium of reactor vessel. The division 3 of ASME code section XI specifies the visual inspection as major in-service inspection (ISI) methods of reactor internal structures and components. Reactor internals of LMR can not be visually examined due to opaque liquid sodium. The under-sodium viewing techniques using an ultrasonic wave should be applied for the visual inspection of reactor internals. Recently, an ultrasonic waveguide sensor with a strip plate has been developed for an application to the under-sodium inspection. In this study, visualization technique, ranging technique and monitoring technique have been suggested for the remote inspection of reactor internals by using the waveguide sensor. The feasibility of these remote inspection techniques using ultrasonic waveguide sensor has been evaluated by an experimental verification.

  17. Principle tests of reactor pumped laser of 3He-Ar-Xe system

    International Nuclear Information System (INIS)

    Chen Hande; Jin Xingxing; Yang Chengde

    1994-01-01

    A nuclear reactor-pumped laser was demonstrated firstly by using the 3 He(P,n) 3 H reaction to excite a 3 He-Ar-Xe laser. Lasing was observed on the 1.73 μm (5d (3/2) 1 -6p(5/2) 1 transition) in Xe I. The CAEP Pulsed Reactor (CFBR-II) was used as a source of fast neutrons moderated by a 50 mm thick polyethylene sleeve around the laser cell. A laser cell was constructed from K 9 glass tubing 600 mm long and Φ34 mm with each end cut at Brewster's angle (quartz). The laser cell was baked out to 10 -3 Pa prior to filling with research-grade mixture ( 3 He/Ar/Xe = 34.7: 34.7: 0.267 kPa). A dielectric-coated 2 m radius-curvature back mirror (99.7% reflectivity at 1.73 μm) and a flat output coupler (Φ30 mm) were used to form the optical cavity. The cavity optics were placed outside the cell as shown in Fig.3. A schematic of the experimental configuration is shown. Typical laser signals obtained with the Ge detector (-80 degree C) are also shown. Laser output lags the fast neutron pulse by the neutron thermalization time (∼278 μs). The observed light output was shown to be caused by stimulated emission, because the signal disappeared when the reflecting (flat) mirror was covered. This precludes the possibility that the phenomenon was the result of γ-radiation interaction with the detector or the air in the light path. The measured laser energy was 80 times the maximum possible fluorescence energy. The observed wavelength was measured to be 1.73 μm by using glass filters. Laser output duration was approximately 735 μs

  18. Reactor utilization, Annex A

    International Nuclear Information System (INIS)

    Martinc, R.; Stanic, A.

    1984-01-01

    Reactor was operated until August 1984 due to prohibition issued by the Ministry since the reactor does not have the emergency cooling system nor special filters in the ventilation system yet. This means that the operation plan was fulfilled by 69%. This annex includes detailed tables containing data about utilization of reactor experimental channels, irradiated samples, as well as interruptions of operation. Detailed data about reactor power during this period are shown as well

  19. Reactor coolant pump for a nuclear reactor

    International Nuclear Information System (INIS)

    Burkhardt, W.; Richter, G.

    1976-01-01

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

  20. 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.

  1. Irradiation of an uranium silicide prototype in RA-3 reactor

    International Nuclear Information System (INIS)

    Calabrese, R.; Estrik, G.; Notari, C.

    1996-01-01

    The factibility of irradiation of an uranium silicide (U 3 Si 2 ) prototype in the RA-3 reactor was studied. The standard RA-3 fuel element uses U 3 O 8 as fissible material. The enrichment of both standard and prototype is the same: 20% U 235 and also the frame geometry and number of plates is identical. The differences are in the plate dimensions and the fissile content which is higher in the prototype. The cooling conditions of the core allow the insertion of the prototype in any core position, even near the water trap, if the overall power is kept below 5Mw. Nevertheless, the recommendation was to begin irradiation near the periphery and later on move the prototype towards more central positions in order to increase the burnup rate. The prototype was effectively introduced in a peripheral position and the thermal fluxes were measured between plates with the foil activation technique. These were also evaluated with the fuel management codes and a reasonable agreement was found. (author). 5 refs., 3 figs., 3 tabs

  2. Conclusions of the experts group of the RA reactor at the meeting held on November 2 and 3 1964 - Annex 12a; Prilog 12a - Zakljucci sa sastanka Strucnog kolegijuma reaktora RA odrzanog na dan 2. i 3. novembra 1964. godine

    Energy Technology Data Exchange (ETDEWEB)

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

    1964-12-15

    Conclusions of the experts group of the RA reactor are related to: analyses of reactor operation at 6.5 MW power with heavy water coolant flow of 250 m{sup 3}/h (2 pumps rotation speed 1500 rotations/min); decisions of future operation; further preparation activities related to reactor operation in forced regime and reduced cooling conditions.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  4. Computational analysis of neutronic parameters for TRIGA Mark-II research reactor using evaluated nuclear data libraries ENDF/B-VII.0 and JENDL-3.3

    International Nuclear Information System (INIS)

    Altaf, M.H.; Badrun, N.H.; Chowdhury, M.T.

    2015-01-01

    Highlights: • SRAC-PIJ code and SRAC-CITATION have been utilized to model the core. • Most of the simulated results show no significant differences with references. • Thermal peak flux varies a bit due to up condition of TRIGA. • ENDF/B-VII.0 and JENDL-3.3 libraries perform well for neutronics analysis of TRIGA. - Abstract: Important kinetic parameters such as effective multiplication factor, k eff , excess reactivity, neutron flux and power distribution, and power peaking factors of TRIGA Mark II research reactor in Bangladesh have been calculated using the comprehensive neutronics calculation code system SRAC 2006 with the evaluated nuclear data libraries ENDF/B-VII.0 and JENDL-3.3. In the code system, PIJ code was employed to obtain cross section of the core cells, followed by the integral calculation of neutronic parameters of the reactor conducted by CITATION code. All the analyses were performed using the 7-group macroscopic cross section library. Results were compared to the experimental data, the safety analysis report (SAR) of the reactor provided by General Atomic as well as to the simulated values by numerically benchmarked MCNP4C, WIMS-CITATION and SRAC-CITATION codes. The maximum power densities at the hot spot were found to be 169.7 W/cc and 170.1 W/cc for data libraries ENDF/B-VII.0 and JENDL-3.3, respectively. Similarly, the total peaking factors based on ENDF/B-VII.0 and JENDL-3.3 were calculated as 5.68 and 5.70, respectively, which were compared to the original SAR value of 5.63, as well as to MCNP4C, WIMS-CITATION and SRAC-CITATION results. It was found in most cases that the calculated results demonstrate a good agreement with our experiments and published works. Therefore, this analysis benchmarks the code system and will be helpful to enhance further neutronics and thermal hydraulics study of the reactor

  5. Design of A solar Thermophilic Anaerobic Reactor for Small Farms

    NARCIS (Netherlands)

    Mashad, El H.; Loon, van W.K.P.; Zeeman, G.; Bot, G.P.A.; Lettinga, G.

    2004-01-01

    A 10 m(3) completely stirred tank reactor has been designed for anaerobic treatment of liquid cow manure under thermophilic conditions (50degreesC), using a solar heating system mounted on the reactor roof. Simulation models for two systems have been developed. The first system consists of loose

  6. Development of a trickle bed reactor of electro-Fenton process for wastewater treatment

    International Nuclear Information System (INIS)

    Lei, Yangming; Liu, Hong; Shen, Zhemin; Wang, Wenhua

    2013-01-01

    Highlights: • An electrochemical trickle bed reactor was composed of C-PTFE-coated graphite chips. • The trickle bed reactor had a high H 2 O 2 production rate in a dilute electrolyte. • An azo dye was effectively decomposed by the electro-Fenton process in the reactor. -- Abstract: To avoid electrolyte leakage and gas bubbles in the electro-Fenton (E-Fenton) reactors using a gas diffusion cathode, we developed a trickle bed cathode by coating a layer composed of carbon black and polytetrafluoroethylene (C-PTFE) onto graphite chips instead of carbon cloth. The trickle bed cathode was optimized by single-factor and orthogonal experiments, in which carbon black, PTFE, and a surfactant were considered as the determinant of the performance of graphite chips. In the reactor assembled by the trickle bed cathode, H 2 O 2 was generated with a current of 0.3 A and a current efficiency of 60%. This performance was attributed to the fine distribution of electrolyte and air, as well as the effective oxygen transfer from the gas phase to the electrolyte–cathode interface. In terms of H 2 O 2 generation and current efficiency, the developed trickle bed reactor had a performance comparable to that of the conventional E-Fenton reactor using a gas diffusion cathode. Further, 123 mg L −1 of reactive brilliant red X-3B in aqueous solution was decomposed in the optimized trickle bed reactor as E-Fenton reactor. The decolorization ratio reached 97% within 20 min, and the mineralization reached 87% within 3 h

  7. Conceptual design of a moving-ring reactor

    International Nuclear Information System (INIS)

    Smith, A.C.; Carlson, G.A.; Ashworth, C.P.

    1986-01-01

    A design of a prototype moving-ring reactor was completed, and a development plan for a pilot reactor is outlined. The fusion fuel is confined in current-carrying rings of magnetically field-reversed plasma (''compact toroids''). The plasma rings, formed by a coaxial plasma gun, undergo adiabatic magnetic compression to ignition temperature while they are being injected into the reactor's burner section. The cylindrical burner chamber is divided into three ''burn stations.'' Separator coils and a slight axial guide field gradient are used to shuttle the ignited toroids rapidly from one burn station to the next, pausing for one-third of the total burn time at each station. Deuterium-tritium- 3 He ice pellets refuel the rings at a rate that maintains constant radiated power. The fusion power per ring is approx. =105.5 MW. The burn time to reach a fusion energy gain of Q = 30 is 5.9 s

  8. 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

  9. Physical modelling of the composting environment: A review. Part 1: Reactor systems

    International Nuclear Information System (INIS)

    Mason, I.G.; Milke, M.W.

    2005-01-01

    In this paper, laboratory- and pilot-scale reactors used for investigation of the composting process are described and their characteristics and application reviewed. Reactor types were categorised by the present authors as fixed-temperature, self-heating, controlled temperature difference and controlled heat flux, depending upon the means of management of heat flux through vessel walls. The review indicated that fixed-temperature reactors have significant applications in studying reaction rates and other phenomena, but may self-heat to higher temperatures during the process. Self-heating laboratory-scale reactors, although inexpensive and uncomplicated, were shown to typically suffer from disproportionately large losses through the walls, even with substantial insulation present. At pilot scale, however, even moderately insulated self-heating reactors are able to reproduce wall losses similar to those reported for full-scale systems, and a simple technique for estimation of insulation requirements for self-heating reactors is presented. In contrast, controlled temperature difference and controlled heat flux laboratory reactors can provide spatial temperature differentials similar to those in full-scale systems, and can simulate full-scale wall losses. Surface area to volume ratios, a significant factor in terms of heat loss through vessel walls, were estimated by the present authors at 5.0-88.0 m 2 /m 3 for experimental composting reactors and 0.4-3.8 m 2 /m 3 for full-scale systems. Non-thermodynamic factors such as compression, sidewall airflow effects, channelling and mixing may affect simulation performance and are discussed. Further work to investigate wall effects in composting reactors, to obtain more data on horizontal temperature profiles and rates of biological heat production, to incorporate compressive effects into experimental reactors and to investigate experimental systems employing natural ventilation is suggested

  10. CFX-10 and RELAP5-3D simulations of coolant mixing phenomena in RPV of VVER-1000 reactors

    International Nuclear Information System (INIS)

    Terzuoli, F.; Moretti, F.; Melideo, D.; D'Auria, F.; Shkarupa, O.

    2006-01-01

    The present paper deals with numerical analyses of coolant mixing in the reactor pressure vessel of a VVER-1000 reactor, performed with the ANSYS CFX-10 CFD code and with the RELAP5-3D system code. In particular, the attention focused on the 'swirl' effect that has been observed to take place in the downcomer of such kind of reactor, with the aim of assessing the capability of the codes to predict that effect, and to understand the reasons for its occurrence. The results have been compared against experimental data from V1000CT-2 Benchmark. (author)

  11. Annual report of Department of Research Reactor and Tandem Accelerator, JFY2011. Operation, utilization and technical development of JRR-3, JRR-4, NSRR and tandem accelerator

    International Nuclear Information System (INIS)

    Ishii, Tetsuro; Nakamura, Kiyoshi; Kawamata, Satoshi; Ishikuro, Yasuhiro; Kawashima, Kazuhito; Kabumoto, Hiroshi; Nakamura, Takemi; Tamura, Itaru; Kawasaki, Sayuri; Sataka, Masao

    2013-03-01

    The Department of Research Reactors and Tandem Accelerator is in charge of the operation, utilization and technical development of JRR-3(Japan Research Reactor No.3), JRR-4(Japan Research Reactor No.4), NSRR(Nuclear Safety Research Reactor) and Tandem Accelerator. This annual report describes a summary of activities of services and technical developments carried out in the period between April 1, 2011 and March 31, 2012. The activities were categorized into six service/development fields: (1) Recovery from the Great East Japan Earthquake, (2) Operation and maintenance of research reactors and tandem accelerator, (3) Utilization of research reactors and tandem accelerator, (4) Upgrading of utilization techniques of research reactors and tandem accelerator, (5) Safety administration for research reactors and tandem accelerator, (6) International cooperation. Also contained are lists of publications, meetings, granted permissions on lows and regulations concerning atomic energy, number of staff members dispatched to Fukushima for the technical assistance, commendation, outcomes in service and technical developments and so on. (author)

  12. Nuclear reactor

    International Nuclear Information System (INIS)

    Schulze, I.; Gutscher, E.

    1980-01-01

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

  13. Managing severe reactor accidents. A review and evaluation of our knowledge on reactor accidents and accident management

    International Nuclear Information System (INIS)

    Gustavsson, Veine

    2002-11-01

    The report gives a review of the results from the last years research on severe reactor accidents, and an opinion on the possibilities to refine the present strategies for accident management in Swedish and Finnish BWRs. The following aspect of reactor accidents are the major themes of the study: 1. Early pressure relief from hydrogen production; 2. Recriticality in re-flooded, degraded core; 3. Melt-through; 4. Steam explosion after melt-through; 5. Coolability of the melt after after melt-through; 6. Hydrogen fire in the reactor containment; 7. Leaking containment; 8. Hydrogen fire in the reactor building; 9. Long-time developments after a severe accident; 10. Accidents during shutdown for overhaul; 11. Information need for remedial actions. Possibilities for improving the strategies in each of these areas are discussed. The review shows that our knowledge is sufficient in the areas 1, 2, 4, 6, 8. For the other areas, more research is needed

  14. Conceptual design of a mirror reactor for a fusion engineering research facility (FERF)

    International Nuclear Information System (INIS)

    Batzer, T.H.; Burleigh, R.C.; Carlson, G.A.; Dexter, W.L.; Hamilton, G.W.; Harvey, A.R.; Hickman, R.G.; Hoffman, M.A.; Hooper, E.B. Jr.; Moir, R.W.; Nelson, R.L.; Pittenger, L.C.; Smith, B.H.; Taylor, C.E.; Werner, R.W.; Wilcox, T.P.

    1975-01-01

    A conceptual design is presented for a small mirror fusion reactor for a Fusion Engineering Research Facility (FERF). The reactor produces 3.4 MW of fusion power and a useful neutron flux of about 10 14 n.cm -2 .s -1 . Superconducting ''yin-yang'' coils are used, and the plasma is sustained by injection of energetic neutral D 0 and T 0 . Conceptual layouts are given for the reactor, its major components, and supporting facilities. (author)

  15. Long-lived sodium activity from a fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Martini, M; Moioli, P; Romanelli, C M

    1973-06-05

    The /sup 23/Na(n,2n)/sup 22/Na cross section was measured in two different ways; the first in the TRIGA reactor, with the sodium sample surrounded by a natural boron shield, the second in the central channel of the TAPIRO reactor. Irradiated samples were investigated either to measure the /sup 23/ Na(n,2n)/sup 22/Na cross section or to determine the impurities present and their activation cross sections. Unsatisfactory results for the second problem suggested evaluation of the impurity activations by calculation; therefore, a set of group constants for reactions of interest was prepared. (3 figures, 3 tables) (auth)

  16. Reactor core in FBR type reactor

    International Nuclear Information System (INIS)

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

    1989-01-01

    In a reactor core in FBR type reactors, a portion of homogenous fuels constituting the homogenous reactor core is replaced with multi-region fuels in which the enrichment degree of fissile materials is lower nearer to the axial center. This enables to condition the composition such that a reactor core having neutron flux distribution either of a homogenous reactor core or a heterogenous reactor core has substantially identical reactivity. Accordingly, in the transfer from the homogenous reactor core to the axially heterogenous reactor core, the average reactivity in the reactor core is substantially equal in each of the cycles. Further, by replacing a portion of the homogenous fuels with a multi-region fuels, thereby increasing the heat generation near the axial center, it is possiable to reduce the linear power output in the regions above and below thereof and, in addition, to improve the thermal margin in the reactor core. (T.M.)

  17. 3 MW TRIGA Research Reactor facility of BAEC and its Utilization

    Energy Technology Data Exchange (ETDEWEB)

    Molla, N.I.; Bhuiyan, S.I.; Wadud Mondal, M.A.; Ahmed, F.U.; Islam, M.N.; Hossain, S.M.; Ahmed, K.; Zulquarnain, A.; Abedin, Z. [Bangladesh Atomic Energy Commission, Atomic Energy Research Establishment, Dhaka (Bangladesh)

    1999-08-01

    The paper briefly describes the Utilisation of 3 MW TRIGA Research Reactor of BAEC for neutron beam research, neutron activation analysis are isotope production. It includes the installation of the triple axis neutron spectrometer at the radial piercing beam port and a neutron radiography set-up at the tangential beam port and their uses for material analysis and condensed matter research and material testing. Nuclear and magnetic structures of some ferrites have been studied in powder diffraction method in the double axis mode. SANS technique with double crystal diffraction known as Bonse and Hart's method has been adopted in an experiment with alumina sample. The neutron radiography set-up and its use in the detection of corrosion in alumina have been reported. Determination of arsenic concentration in drinking water from tube well via Instrumental Neutron Activation Analysis and production of radioiodine-131 by dry distillation method are presented. Our experience on the removal of N-16 decay tank because of the leakage of coolant and bringing the research reactor back to operational by-passing the decay tank have been focussed. A possible reconfiguration of the existing TRIGA core, without exceeding the safety margins, providing additional irradiation channel and upgrading the neutron flux for increased radioisotope production has been attempted. Cross section library ENDF/B-VI and JENDL3.2, code NJOY94.10, WIMSD package, 3-D code CITATION, PARET and Monte Carlo code MCNP4B2 have been employed to achieve the objective. (author)

  18. Development of Pneumatic Transfer Irradiation Facility (PTS no.3) 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

    2008-04-15

    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. The pneumatic transfer irradiation system (PTS no.3) involving a manual system and an semi-automatic system were reconstructed with new designs of a functional improvement at the HANARO research reactor and NAA laboratory of RI building in 2006. In this technical report, the design, operation and control of these system (PTS no.3) was described. Also the experimental results and the characteristic parameters measured from 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.

  19. Advanced reactor development

    International Nuclear Information System (INIS)

    Till, C.E.

    1989-01-01

    Consideration is given to what the aims of advanced reactor development have to be, if a new generation of nuclear power is really to play an important role in man's energy generation activities in a fragile environment. The background given briefly covers present atmospheric evidence, the current situation in nuclear power, how reactors work and what can go wrong with them, and the present magnitudes of world energy generation. The central part of the paper describes what is currently being done in advanced reactor development and what can be expected from various systems and various elements of it. A vigorous case is made that three elements must be present in any advanced reactor development: (1) breeding; (2) passive safety; and (3) shorter-live nuclear waste. All three are possible. In the right advanced reactor systems the ways of achieving them are known. But R and D is necessary. That is the central argument made in the paper. Not advanced reactor prototype construction at this point, but R and D itself. (author)

  20. Ethanol dehydration to ethylene in a stratified autothermal millisecond reactor.

    Science.gov (United States)

    Skinner, Michael J; Michor, Edward L; Fan, Wei; Tsapatsis, Michael; Bhan, Aditya; Schmidt, Lanny D

    2011-08-22

    The concurrent decomposition and deoxygenation of ethanol was accomplished in a stratified reactor with 50-80 ms contact times. The stratified reactor comprised an upstream oxidation zone that contained Pt-coated Al(2)O(3) beads and a downstream dehydration zone consisting of H-ZSM-5 zeolite films deposited on Al(2)O(3) monoliths. Ethanol conversion, product selectivity, and reactor temperature profiles were measured for a range of fuel:oxygen ratios for two autothermal reactor configurations using two different sacrificial fuel mixtures: a parallel hydrogen-ethanol feed system and a series methane-ethanol feed system. Increasing the amount of oxygen relative to the fuel resulted in a monotonic increase in ethanol conversion in both reaction zones. The majority of the converted carbon was in the form of ethylene, where the ethanol carbon-carbon bonds stayed intact while the oxygen was removed. Over 90% yield of ethylene was achieved by using methane as a sacrificial fuel. These results demonstrate that noble metals can be successfully paired with zeolites to create a stratified autothermal reactor capable of removing oxygen from biomass model compounds in a compact, continuous flow system that can be configured to have multiple feed inputs, depending on process restrictions. Copyright © 2011 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  1. Safety equipment in a reactor

    International Nuclear Information System (INIS)

    Shiratori, Hirozo; Ishiyama, Satoshi; Ugawa, Yukio.

    1976-01-01

    Object: To safely retain, even if fuel should be molten and flown through the bottom of a container in a reactor, the molten fuel to remove heat generation of the fuel to prevent occurrence of a critical trouble. Structure: A reactor container housing a core and coolant has thereunder a separation dome in a central portion thereof and a partitioning plate coaxially and circularly disposed in the periphery of the separation dome, with a tray formed of magnesium oxide being disposed. Further, a cooling path system is provided so as to surround the tray. The cooling path system and the reactor container are surrounded and protected by a reactor wall provided with heat insulating refractory bricks, a coolant pouring system extends through the reactor wall, and the coolant is supplied to the tray. (Furukawa, Y.)

  2. Fusion reactor development: A review

    International Nuclear Information System (INIS)

    Anon.

    1983-01-01

    This paper is a review of the current prospects for fusion reactor development based upon the present status in plasma physics research, fusion technology development and reactor conceptual design for the tokamak magnetic confinement concept. Recent advances in tokamak plasma research and fusion technology development are summarized. The direction and conclusions of tokamak reactor conceptual design are discussed. The status of alternate magnetic confinement concept research is reviewed briefly. A feasible timetable for the development of fusion reactors is presented

  3. Comparison of 'system thermal-hydraulics-3 dimensional reactor kinetics' coupled calculations using the MARS 1D and 3D modules and the MASTER code

    International Nuclear Information System (INIS)

    Jung, J. J.; Joo, H. K.; Lee, W. J.; Ji, S. K.; Jung, B. D.

    2002-01-01

    KAERI has developed the coupled 'system thermal-hydraulics - 3 dimensional reactor kinetics' code, MARS/MASTER since 1998. However, there is a limitation in the existing MARS/MASTER code; that is, to perform the coupled calculations using MARS/MASTER, we have to utilize the hydrodynamic model and the heat structure model of the MARS '3D module'. In some transients, reactor kinetics behavior is strongly multi-dimensional, but core thermal-hydraulic behavior remains in one-dimensional manner. For efficient analysis of such transients, we coupled the MARS 1D module with MASTER. The new feature has been assessed by the 'OECD NEA Main Steam Line Break (MSLB) benchmark exercise III' simulations

  4. Real-Time Quantitative Operando Raman Spectroscopy of a CrOx/Al2O3 Propane Dehydrogenation Catalyst in a Pilot-Scale Reactor

    NARCIS (Netherlands)

    Sattler, Jesper J. H. B.; Mens, Ad M.; Weckhuysen, Bert M.

    2014-01-01

    Combined operando UV/vis-Raman spectroscopy has been used to study the deactivation of CrOx/Al2O3 catalyst extrudates in a pilot scale propane dehydrogenation reactor. For this purpose, UV/vis and Raman optical fiber probes have been designed, constructed and tested. The light absorption measured by

  5. Seismic analysis of a large pool-type LMR [liquid metal reactor

    International Nuclear Information System (INIS)

    Wang, C.Y.; Gvildys, J.

    1989-01-01

    This paper describes the seismic study of a 450-MWe liquid metal reactor (LMR) under 0.3-g SSE ground excitation. Two calculations were performed using the new design configuration. They deal with the seismic response of the reactor vessel, the guard vessel and support skirt, respectively. In both calculations, the stress and displacement fields at important locations of those components are investigated. Assessments are also made on the elastic and inelastic structural capabilities for other beyond-design basis seismic loads. Results of the reactor vessel analysis reveal that the maximum equivalent stress is only about half of the material yield stress. For the guard vessel and support skirt, the stress level is very small. Regarding the analysis if inelastic structural capability, solutions of the Newmark-Hall ductility modification method show that the reactor vessel can withstand seismics with ground ZPAs ranging from 1.015 to 1.31 g, which corresponds to 3.37 to 4.37 times the basic 0.3-g SSE. Thus, the reactor vessel and guard vessel are strong enough to resist seismic loads. 4 refs., 10 figs., 5 tabs

  6. Study of heat transfer in 3D fuel rods of the EPRI-9R reactor modified

    International Nuclear Information System (INIS)

    Affonso, Renato Raoni Werneck; Lava, Deise Diana; Borges, Diogo da Silva; Sampaio, Paulo Augusto Berquo de; Moreira, Maria de Lourdes

    2014-01-01

    This paper aims to conduct a case study of the fuel rods that have the highest and the lowest average power of the EPRI-9R 3D reactor modified , for various positions of the control rods banks. For this, will be addressed the verification of computer code, comparing the results obtained with analytical solutions. This check is important so that, subsequently, it is possible use the program to understand the behavior of the fuel rods and the coolant channel of the EPRI-9R 3D reactor modified. Thus, in view of the scope of this paper, first a brief introducing on the heat transfer is done, including the rod equations and the equation of energy in the channel to allow the analysis of the results

  7. Reactor FaceMap Tool: A modern graphics tool for displaying reactor data

    International Nuclear Information System (INIS)

    Roberts, J.C.

    1991-01-01

    A prominent graphical user interface in reactor physics applications at the Savannah River Site is the reactor facemap display. This is a two dimensional view of a cross section of a reactor. In the past each application which needed a facemap implemented its own version. Thus, none of the code was reused, the facemap implementation was hardware dependent and the user interface was different for each facemap. The Reactor FaceMap Tool was built to solve these problems. Through the use of modern computing technologies such as X Windows, object-oriented programming and client/server technology the Reactor FaceMap Tool has the flexibility to work in many diverse applications and the portability to run on numerous types of hardware

  8. Conceptual design of a moving-ring reactor

    International Nuclear Information System (INIS)

    Smith, A.C. Jr.; Ashworth, C.P.; Abreu, K.E.

    1983-01-01

    A design of a prototype Moving-Ring Reactor has been completed. The fusion fuel is confined in current-carrying rings of magnetically field-reversed plasma (''compact toroids''). The plasma rings, formed by a coaxial plasma gun, undergo adiabatic magnetic compression to ignition temperature while they are being injected into the reactor's burner section. The cylindrical burner chamber is divided into three ''burn stations''. Separator coils and a slight axial guide-field gradient are used to shuttle the ignited toroids rapidly from one burn station to the next, pausing for one third of the total burn time at each station. D-T- 3 He ice pellets refuel the rings at a rate which maintains constant radiated power. The first wall and tritium breeding blanket designs make credible use of helium cooling, SiC and Li 2 O to minimize structural radioactivity. ''Hands-on'' maintenance is possible on all reactor components outside the blanket. The first wall and blanket are designed to shut the reactor down passively in the event of a loss-of-coolant or loss-of-flow accident. Helium removes heat from the first wall, blanket and shield, and is used in a closed-cycle gas turbine to produce electricity. Energy residing in the plasma ring at the end of the burn is recovered via magnetic expansion. Electrostatic direct conversion is not used in this design. The reactor produces a constant net power of 99 MW(e). (author)

  9. Reactor Structure Materials: Corrosion of Reactor Core Internals

    International Nuclear Information System (INIS)

    Van Dyck, S.

    2000-01-01

    The objectives of SCK-CEN's R and D programme on the corrosion of reactor core internals are: (1) to gain mechanistic insight into the Irradition Assisted Stress Corrosion Cracking (IASCC) phenomenon by studying the influence of separate parameters in well controlled experiments; (2) to develop and validate a predictive capability on IASCC by model description and (3) to define and validate countermeasures and monitoring techniques for application in reactors. Progress and achievements in 1999 are described

  10. A WIMS-NESTLE reactor physics model for an RBMK reactor

    International Nuclear Information System (INIS)

    Perry, R.T.; Meriwether, G.H.

    1996-01-01

    This work describes the static neutronic calculations made for a three-dimensional model of an RBMK (Russian) reactor. Future work will involve the use of this neutronic model and a thermal-hydraulic model in coupled calculations. The lattice code, WIMS-D, was used to obtain the cross sections for the static neutronic calculations. The static reactor neutronic calculations were made with NESTLE, a three-dimensional nodal diffusion code. The methods used to establish an RBMK reactor model for use in these codes are discussed, and the cross sections calculated are given

  11. A WIMS-NESTLE reactor physics model for an RBMK reactor

    International Nuclear Information System (INIS)

    Perry, R.T.; Meriwether, G.H.

    1996-01-01

    This work describes the static neutronic calculations made for a three-dimensional model of an RBMK (Russian) reactor. Future work will involve the use of this neutronic model and a thermal-hydraulic model in coupled calculations. The lattice code, WIMS-D, was used to obtain the cross sections for the static neutronic calculations. The static reactor neutronic calculations were made with NESTLE, a three-dimensional nodal diffusion code. The methods used to establish an RBMK reactor model for use in these codes are discussed, and the cross sections calculated are given. (author)

  12. Automatic power control for a pressurized water reactor

    International Nuclear Information System (INIS)

    Hah, Yung Joon

    1994-02-01

    During a normal operation of a pressurized water reactor (PWR), the reactivity is controlled by control rods, boron, and the average temperature of the primary coolant. Especially in load follow operation, the reactivity change is induced by changes in power level and effects of xenon concentration. The control of the core power distribution is concerned, mainly, with the axial power distribution which depends on insertion and withdrawal of the control rods resulting in additional reactivity compensation. The utilization of part strength control element assemblies (PSCEAs) is quite appropriate for a control of the power distribution in the case of Yonggwang Nuclear Unit 3 (YGN Unit 3). However, control of the PSCEAs is not automatic, and changes in the boron concentration by dilution/boration are done manually. Thus, manual control of the PSCEAs and the boron concentration require the operator's experience and knowledge for a successful load follow operation. In this thesis, the new concepts have been proposed to adapt for an automatic power control in a PWR. One of the new concepts is the mode K control, another is a fuzzy power control. The system in mode K control implements a heavy-worth bank dedicated to axial shape control, independent of the existing regulating banks. The heavy bank provides a monotonic relationship between its motion and the axial power shape change, which allows automatic control of the axial power distribution. And the mode K enables precise regulation, by using double closed-loop control of the reactor coolant temperature and the axial power difference. Automatic reactor power control permits the nuclear power plant to accommodate the load follow operations, including frequency control, to respond to the grid requirements. The mode K reactor control concepts were tested using simulation responses of a Korean standardized 1000-MWe PWR which is a reference plant for the YGN Unit 3. The simulation results illustrate that the mode K would be

  13. Activity report on the utilization of research reactors (JRR-3 and JRR-4). Japanese fiscal year, 2007

    International Nuclear Information System (INIS)

    2012-03-01

    In the fiscal year 2007, the research reactor JRR-3 was operated for 7 cycles (cycle operation : 26days/cycle) and the JRR-4 was operated for 92 days. JRR-3 is used for the purposes below; Experimental studies such as neutron scattering, prompt gamma-ray analyses, neutron radiography, Irradiation for activation analyses, radioisotope (RI) productions, fission tracks, Irradiation test of reactor materials, etc. JRR-4 is used for the purposes below; Medical irradiation (Boron Neutron Capture Therapy : BNCT), Prompt gamma-ray analyses, Sensitivity measurement of radiation detectors, Experiment and practice in the nuclear reactor training, Irradiation for activation analyses, RI productions, fission tracks, etc. The volume contains 262 activity reports, which are categorized into the fields of neutron scattering (10 subcategories), neutron radiography, neutron activation analyses, prompt gamma-ray analyses, and others submitted by the users in JAEA and other Organizations. (author)

  14. 3-D thermal hydraulic analysis of transient heat removal from fast reactor core using immersion coolers

    International Nuclear Information System (INIS)

    Chvetsov, I.; Volkov, A.

    2000-01-01

    For advanced fast reactors (EFR, BN-600M, BN-1600, CEFR) the special complementary loop is envisaged in order to ensure the decay heat removal from the core in the case of LOF accidents. This complementary loop includes immersion coolers that are located in the hot reactor plenum. To analyze the transient process in the reactor when immersion coolers come into operation one needs to involve 3-D thermal hydraulics code. Furthermore sometimes the problem becomes more complicated due to necessity of simulation of the thermal hydraulics processes into the core interwrapper space. For example on BN-600M and CEFR reactors it is supposed to ensure the effective removal of decay heat from core subassemblies by specially arranged internal circulation circuit: 'inter-wrapper space'. For thermal hydraulics analysis of the transients in the core and in the whole reactor including hot plenum with immersion coolers and considering heat and mass exchange between the main sodium flow and sodium that moves in the inter-wrapper space the code GRIFIC (the version of GRIF code family) was developed in IPPE. GRIFIC code was tested on experimental data obtained on RAMONA rig under conditions simulating decay heat removal of a reactor with the use of immersion coolers. Comparison has been made of calculated and experimental result, such as integral characteristics (flow rate through the core and water temperature at the core inlet and outlet) and the local temperatures (at thermocouple location) as well. In order to show the capabilities of the code some results of the transient analysis of heat removal from the core of BN-600M - type reactor under loss-of-flow accident are presented. (author)

  15. Power measurement of the RA-3 reactor using the neutron noise technique and 16N

    International Nuclear Information System (INIS)

    Gomez, Angel

    2003-01-01

    This work describes a measurement method based on the neutron noise technique which is used for determining the relation between the power and the currents of two ionization chambers. These chambers are sensitive to the gamma radiation from the 16 N decay produced in the RA-3 reactor core. The power during operation is obtained from the calibration factors by measuring those currents. As this calibration factors depend on the cooler flow that circulates in the reactor core and in the 16 N measuring system, an estimator, that is a function of the ratio of this currents, is proposed in order to detect flow changes. (author)

  16. Calculation of the transmutation rates of Tc-99, I-129 and Cs-135 in the High Flux Reactor, in the Phenix Reactor and in a light water reactor

    International Nuclear Information System (INIS)

    Bultman, J.

    1992-04-01

    Transmutation of long-lived fission products is of interest for the reduction of the possible dose to the population resulting from long-term leakage of nuclear waste from waste disposals. Three isotopes are of special interest: Tc-99, I-129 and Cs-135. Therefore, experiments on transmutation of these isotopes in nuclear reactors are planned. In the present study, the possible transmutation rates and mass reductions are determined for experiments in High Flux Reactor (HFR) located in Petten (Netherlands) and in Phenix (France). Also, rates were determined for a standard Light Water Reactor (LWR). The transmutation rates of the 3 fission products will be much higher in HFR than in Phenix reactor, as both total flux and effective cross sections are higher. For thick targets the effective half lives are approximately 3, 2 and 7 years for Tc-99, I-129 and Cs-135 irradiation respectively in HFR and 22, 16 and 40 years for Tc-99, I-129 and Cs-135 irradiation in Phenix reactor. The transmutation rates in LWR are low. Only the relatively large power of LWR guarantees a large total mass reduction. Especially transmutation of Cs-135 will be very difficult in Phenix and LWR, clearly shown by the very long effective half lives of 40 and 100 years, respectively. (author). 7 refs.; 5 figs.; 7 tabs

  17. Proceedings of 2. Yugoslav symposium on reactor physics, Part 3, Herceg Novi (Yugoslavia), 27-29 Sep 1966; 2. Jugoslovenski simpozijum iz reaktorske fizike, Deo 3, Herceg Novi (Yugoslavia), 27-29 Sep 1966

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1966-07-01

    This Volume 3 of the Proceedings of 2. Yugoslav symposium on reactor physics includes three papers describing the following: model for spatial synthesis of automated control system of the GCR type reactor; model for analysis of hydrodynamic processes at the BHWR type reactors; mathematical model for safety analysis of heavy water power reactor.

  18. BR2 Reactor: Introduction

    International Nuclear Information System (INIS)

    Moons, F.

    2007-01-01

    The irradiations in the BR2 reactor are in collaboration with or at the request of third parties such as the European Commission, the IAEA, research centres and utilities, reactor vendors or fuel manufacturers. The reactor also contributes significantly to the production of radioisotopes for medical and industrial applications, to neutron silicon doping for the semiconductor industry and to scientific irradiations for universities. Along the ongoing programmes on fuel and materials development, several new irradiation devices are in use or in design. Amongst others a loop providing enhanced cooling for novel materials testing reactor fuel, a device for high temperature gas cooled fuel as well as a rig for the irradiation of metallurgical samples in a Pb-Bi environment. A full scale 3-D heterogeneous model of BR2 is available. The model describes the real hyperbolic arrangement of the reactor and includes the detailed 3-D space dependent distribution of the isotopic fuel depletion in the fuel elements. The model is validated on the reactivity measurements of several tens of BR2 operation cycles. The accurate calculations of the axial and radial distributions of the poisoning of the beryllium matrix by 3 He, 6 Li and 3T are verified on the measured reactivity losses used to predict the reactivity behavior for the coming decades. The model calculates the main functionals in reactor physics like: conventional thermal and equivalent fission neutron fluxes, number of displacements per atom, fission rate, thermal power characteristics as heat flux and linear power density, neutron/gamma heating, determination of the fission energy deposited in fuel plates/rods, neutron multiplication factor and fuel burn-up. For each reactor irradiation project, a detailed geometry model of the experimental device and of its neighborhood is developed. Neutron fluxes are predicted within approximately 10 percent in comparison with the dosimetry measurements. Fission rate, heat flux and

  19. Development of a trickle bed reactor of electro-Fenton process for wastewater treatment

    Energy Technology Data Exchange (ETDEWEB)

    Lei, Yangming [Key Laboratory of Reservoir Aquatic Environment, Chongqing Institute of Green and Intelligent Technology, Chinese Academy of Sciences, Chongqing 401122 (China); School of Environmental Science and Engineering, Shanghai Jiao Tong University, Shanghai 200240 (China); Liu, Hong, E-mail: liuhong@cigit.ac.cn [Key Laboratory of Reservoir Aquatic Environment, Chongqing Institute of Green and Intelligent Technology, Chinese Academy of Sciences, Chongqing 401122 (China); Shen, Zhemin, E-mail: zmshen@sjtu.edu.cn [School of Environmental Science and Engineering, Shanghai Jiao Tong University, Shanghai 200240 (China); Wang, Wenhua [School of Environmental Science and Engineering, Shanghai Jiao Tong University, Shanghai 200240 (China)

    2013-10-15

    Highlights: • An electrochemical trickle bed reactor was composed of C-PTFE-coated graphite chips. • The trickle bed reactor had a high H{sub 2}O{sub 2} production rate in a dilute electrolyte. • An azo dye was effectively decomposed by the electro-Fenton process in the reactor. -- Abstract: To avoid electrolyte leakage and gas bubbles in the electro-Fenton (E-Fenton) reactors using a gas diffusion cathode, we developed a trickle bed cathode by coating a layer composed of carbon black and polytetrafluoroethylene (C-PTFE) onto graphite chips instead of carbon cloth. The trickle bed cathode was optimized by single-factor and orthogonal experiments, in which carbon black, PTFE, and a surfactant were considered as the determinant of the performance of graphite chips. In the reactor assembled by the trickle bed cathode, H{sub 2}O{sub 2} was generated with a current of 0.3 A and a current efficiency of 60%. This performance was attributed to the fine distribution of electrolyte and air, as well as the effective oxygen transfer from the gas phase to the electrolyte–cathode interface. In terms of H{sub 2}O{sub 2} generation and current efficiency, the developed trickle bed reactor had a performance comparable to that of the conventional E-Fenton reactor using a gas diffusion cathode. Further, 123 mg L{sup −1} of reactive brilliant red X-3B in aqueous solution was decomposed in the optimized trickle bed reactor as E-Fenton reactor. The decolorization ratio reached 97% within 20 min, and the mineralization reached 87% within 3 h.

  20. Designing a mini subcritical nuclear reactor

    International Nuclear Information System (INIS)

    Escobedo G, C. R.; Vega C, H. R.; Davila H, V. M.

    2015-10-01

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

  1. A novel spiral reactor for biodiesel production in supercritical ethanol

    International Nuclear Information System (INIS)

    Farobie, Obie; Sasanami, Kazuma; Matsumura, Yukihiko

    2015-01-01

    Highlights: • A novel spiral reactor for biodiesel production in supercritical ethanol was proposed. • The spiral reactor employed in this study successfully recovered heat. • The effects of temperature and time on FAEE yield were investigated. • FAEE yield as high as 0.937 mol/mol was obtained at 350 °C after 30 min. • The second-order kinetic model expressed the experimental yield well. - Abstract: A spiral reactor is proposed as a novel reactor design for biodiesel production under supercritical conditions. Since the spiral reactor serves as a heat exchanger, it offers the advantage of reduced apparatus space compared to conventional supercritical equipment. Experimental investigations were carried out at reaction temperatures of 270–400 °C, pressure of 20 MPa, oil-to-ethanol molar ratio of 1:40, and reaction times of 3–30 min. An FAEE yield of 0.937 mol/mol was obtained in a short reaction time of 30 min at 350 °C and oil-to-ethanol molar ratio of 1:40 under a reactor pressure of 20 MPa. The spiral reactor was not only as effective as conventional reactor in terms of transesterification reactor but also was superior in terms of heat recovery. A second-order kinetic model describing the transesterification of canola oil in supercritical ethanol was proposed, and the reaction was observed to follow Arrhenius behavior. The corresponding reaction rate constants and the activation energies as well as pre-exponential factors were determined

  2. 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

  3. Thermo-mechanical design and structural analysis of the first wall for ARIES-III, A 1000 MWeD-3He power reactor

    International Nuclear Information System (INIS)

    Sviatoslavsky, I.; Blanchard, J.P.; Mogahed, E.A.

    1992-01-01

    This paper reports on ARIES III, a conceptual design study of a 1000 MWe D- 3 He tokamak fusion power reactor in which most of the energy comes from charged particle transport, bremsstrahlung and synchrotron radiation, and only a small fraction (∼ 4%) comes form neutrons. This form of energy is deposited as surface heating on the chamber first wall (FW) and divertor elements, while the neutron energy is deposited as bulk nuclear heating within the shield. Since this reactor does not use tritium, there is no breeding blanket. Instead a shield is provided to protect the magnets from neutrons. The Fw is very unique in a D- 3 He reactor, it must be capable of absorbing the high surface heat in a mode suitable for efficient power cycle conversion, it must be able to reflect synchrotron radiation, and it must be able to withstand high current plasma disruptions. The FW is made of a low activation ferritic steel (MHT-9) and is cooled with an organic coolant (HB-40) at a pressure of 2 MPa. The FW has a coating of 0.01 cm tungsten on the MHT-9, followed by 0.15 cm of Be on the plasma side. This is needed for synchrotron radiation reflection and as a melt layer to guard against the thermal effects of a plasma disruption

  4. Repairing liner of the reactor; Reparacion del liner del reactor

    Energy Technology Data Exchange (ETDEWEB)

    Aguilar H, F. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2001-07-15

    Due to the corrosion problems of the aluminum coating of the reactor pool, a periodic inspections program by ultrasound to evaluate the advance grade and the corrosion speed was settled down. This inspections have shown the necessity to repair some areas, in those that the slimming is significant, of not making it can arrive to the water escape of the reactor pool. The objective of the repair is to place patches of plates of 1/4 inch aluminum thickness in the areas of the reactor 'liner', in those that it has been detected by ultrasound a smaller thickness or similar to 3 mm. To carry out this the fuels are move (of the core and those that are decaying) to a temporary storage, the structure of the core is confined in a tank that this placed inside the pool of the reactor, a shield is placed in the thermal column and it is completely extracted the water for to leave uncover the 'liner' of the reactor. (Author)

  5. 3-dimensional finite element modelling of reactor building internal structure for static analysis

    International Nuclear Information System (INIS)

    Joshi, M.H.; Reddy, V.J.; Kushwaha, H.S.; Reddy, G.R.; Karandikar, G.V.

    1991-01-01

    a) Thin shell element gives fairly accurate results when compared to 3-D Brick element for the type of structure and loading in Reactor Building. b) The maximum element size is fixed from model 3(c) i.e. 2.0 m. c) Openings with size smaller than 0.5 m can be neglected without affecting the results very much. d) For any such problem, the methodology described in this paper can be used to take rational decisions which will ensure reasonable accuracy. (author)

  6. Reinforced confinement in a nuclear reactor

    International Nuclear Information System (INIS)

    Norman, H.

    1988-01-01

    The present invention concerns a nuclear reactor containing a reactor core, a swimming pool space that is filled and pressurized with a neutron-absorbing solution, a reactor tank, at least one heat exchanger, at least one inlet line, at least one return line and at least one circulation pump, where the said reactor tank is confined in the said swimming pool space and designed to be cooled with the aid of relatively pure water, which is fed by means of the said at least one circulating pump to the said reactor tank from the said heat exchanger via the said at least one inlet line and is returned to the heat exchanger via the said at least one return line. The problem that is to be solved by the invention is to design a reactor of the above type in such a way that a complete confinement of the primary circuit of the reactor is achieved at relatively low extra cost. This problem is solved by providing the reactor with a special confinement space that confines the heat exchanger, but not the reactor tank, with the confinement space and the swimming pool space being fashioned in the same concrete body

  7. A neutronics feasibility study for the LEU conversion of Poland's Maria research reactor

    International Nuclear Information System (INIS)

    Bretscher, M. M.

    1998-01-01

    The MARIA reactor is a high-flux multipurpose research reactor which is water-cooled and moderated with both beryllium and water. Standard HEU (80% 235 U)fuel assemblies consist of six concentric fuel tubes of a U-Al alloy clad in aluminum. Although the inventory of HEU (80%) fuel is nearly exhausted, a supply of highly-loaded 36%-enriched fuel assemblies is available at the reactor site. Neutronic equilibrium studies have been made to determine the relative performance of fuels with enrichments of 80%, 36% and 19.7%. These studies indicate that LEU (19.7%) densities of about 2.5 gU/cm 3 and 3.8 gU/cm 3 are required to match the performance of the MARIA reactor with 80%-enriched and with 36%-enriched fuels, respectively

  8. The integral fast reactor

    International Nuclear Information System (INIS)

    Till, C.E.

    1987-01-01

    On April 3rd, 1986, two dramatic demonstrations of the inherent capability of sodium-cooled fast reactors to survive unprotected loss of cooling accidents were carried out on the experimental sodium-cooled power reactor, EBR-II, on the Idaho site of Argonne National Laboratory. Transients potentially of the most serious kind, one an unprotected loss of flow, the other an unprotected loss of heat sink, both initiated from full power. In both cases the reactor quietly shut itself down, without damage of any kind. These tests were a part of the on-going development program at Argonne to develop an advanced reactor with significant new inherent safety characteristics. Called the Integral Fast Reactor, or IFR, the basic thrust is to develop everything that is needed for a complete nuclear power system - reactor, closed fuel cycle, and waste processing - as a single optimized entity, and, for simplicity in concept, as an integral part of a single plant. The particular selection of reactor materials emphasizes inherent safety characteristics and also makes possible a simplified closed fuel cycle and waste process improvements

  9. Report on the Survey of the Design Review of New Reactor Applications. Volume 3: Reactor

    International Nuclear Information System (INIS)

    Downey, Steven; Monninger, John; Nevalainen, Janne; Lorin, Aurelie; ); Webster, Philip; Joyer, Philippe; Kawamura, Tomonori; Lankin, Mikhail; Kubanyi, Jozef; Haluska, Ladislav; Persic, Andreja; Reierson, Craig; Kang, Kyungmin; Kim, Walter

    2016-01-01

    At the tenth meeting of the CNRA Working Group on the Regulation of New Reactors (WGRNR) in March 2013, the Working Group agreed to present the responses to the Second Phase, or Design Phase, of the Licensing Process Survey as a multi-volume text. As such, each report will focus on one of the eleven general technical categories covered in the survey. The general technical categories were selected to conform to the topics covered in the International Atomic Energy Agency (IAEA) Safety Guide GS-G-4.1. This document, which is the third report on the results of the Design Phase Survey, focuses on the Reactor. The Reactor category includes the following technical topics: fuel system design, reactor internals and core support, nuclear design and core nuclear performance, thermal and hydraulic design, reactor materials, and functional design of reactivity control system. For each technical topic, the member countries described the information provided by the applicant, the scope and level of detail of the technical review, the technical basis for granting regulatory authorisation, the skill sets required and the level of effort needed to perform the review. Based on a comparison of the information provided by the member countries in response to the survey, the following observations were made: - Although the description of the information provided by the applicant differs in scope and level of detail among the member countries that provided responses, there are similarities in the information that is required. - All of the technical topics covered in the survey are reviewed in some manner by all of the regulatory authorities that provided responses. - Design review strategies most commonly used to confirm that the regulatory requirements have been met include document review and independent verification of calculations, computer codes, or models used to describe the design and performance of the core and the fuel. - It is common to consider operating experience and

  10. Adapting a reactor safety assessment system for specific plants

    International Nuclear Information System (INIS)

    Ballard, T.L.; Cordes, G.A.

    1991-01-01

    The Reactor Safety Assessment System (RSAS) is an expert system being developed by the Idaho National Engineering Laboratory, the University of Maryland (UofM) and US Nuclear Regulatory Commission (NRC) for use in the NRC Operations center. RSAS is designed to help the Reactor Safety Team monitor and project core status during an emergency at a licensed nuclear power plant. Analysis uses a hierarchical plant model based on equipment availability and automatically input parametric plant information. There are 3 families of designs of pressurized water reactors and 75 plants using modified versions of the basic design. In order to make an RSAS model for each power plant, a generic model for a given plant type is used with differences being specified by plant specific files. Graphical displays of this knowledge are flexible enough to handle any plant configuration. A variety of tools have been implemented to make it easy to modify a design to fit a given plant while minimizing chance for error. 3 refs., 4 figs

  11. Transmutation of actinide 237Np with a fusion reactor and a hybrid reactor

    International Nuclear Information System (INIS)

    Feng, K.M.; Huang, J.H.

    1994-01-01

    The use of fusion reactors to transmute fission reactor wastes to stable species is an attractive concept. In this paper, the feasibility of transmutation of the long-lived actinide radioactive waste Np-237 with a fusion reactor and a hybrid reactor has been investigated. A new waste management concept of burning HLW (High Level Waste), utilizing released energy and converting Np-237 into fissile fuel Pu-239 through transmutation has been adopted. The detailed neutronics and depletion calculation of waste inventories was carried out with a modified version of one-dimensional neutron transport and burnup calculation code system BISON1.5 in this study. The transmutation rate of Np with relationship to neutron wall loading, Pu and Np with relationship to neutron wall load, Pu and Np concentration in the transmutation zone have been explored as well as relevant results are also given

  12. Molten salt reactors

    International Nuclear Information System (INIS)

    Bouchter, J.C.; Dufour, P.; Guidez, J.; Simon, N.; Renault, C.

    2014-01-01

    Molten salt reactors are one of the 6 concepts retained for the 4. generation of nuclear reactors. The principle of this reactor is very innovative: the nuclear fuel is dissolved in the coolant which allows the online reprocessing of the fuel and the online recovery of the fission products. A small prototype: the Molten Salt Reactor Experiment (MSRE - 8 MWt) was operating a few years in the sixties in the USA. The passage towards a fast reactor by the suppression of the graphite moderator leads to the concept of Molten Salt Fast Reactor (MSFR) which is presently studied through different European projects such as MOST, ALISIA and EVOL. Worldwide the main topics of research are: the adequate materials resisting to the high level of corrosiveness of the molten salts, fuel salt reprocessing, the 3-side coupling between neutron transport, thermohydraulics and thermo-chemistry, the management of the changing chemical composition of the salt, the enrichment of lithium with Li 7 in the case of the use of lithium fluoride salt and the use of MSFR using U 233 fuel (thorium cycle). The last part of the article presents a preliminary safety analysis of the MSFR. (A.C.)

  13. Advanced reactor development: The LMR integral fast reactor program at Argonne

    International Nuclear Information System (INIS)

    Till, C.E.

    1990-01-01

    Reactor technology for the 21st Century must develop with characteristics that can now be seen to be important for the future, quite different from the things when the fundamental materials and design choices for present reactors were made in the 1950s. Argonne National Laboratory, since 1984, has been developing the Integral Fast Reactor (IFR). This paper will describe the way in which this new reactor concept came about; the technical, public acceptance, and environmental issues that are addressed by the IFR; the technical progress that has been made; and our expectations for this program in the near term. 3 figs

  14. Reactor physics aspects of burning actinides in a nuclear reactor

    International Nuclear Information System (INIS)

    Hage, W.; Schmidt, E.

    1978-01-01

    A short review of the different recycling strategies of actinides other than fuel treated in the literature, is given along with nuclear data requirements for actinide build-up and transmutation studies. The effects of recycling actinides in a nuclear reactor on the flux distribution, the infinite neutron multiplication factor, the reactivity control system, the reactivity coefficients and the delayed neutron fraction are discussed considering a notional LWR or LMFBR as an Actinide Trasmutaton Reactor. Some operational problems of Actinide Transmutation reactors are mentioned, which are caused by the α-decay heat and the neutron sources of Actinide Target Elements

  15. User's guide of DETRAS system-3. Description of the simulated reactor plant

    International Nuclear Information System (INIS)

    Yamaguchi, Yukichi

    2006-12-01

    DETRAS system is a PWR reactor simulator system for operation trainings whose distinguished feature is that it can be operated from the remote place of the simulator site. The document which is the third one of a series of three volumes of the user's guide of DETRAS, describes firstly an outline of the simulated reactor system then a user's interface needed for operation of the simulator of interest and finally a series of procedure for startup of the simulated reactor and shutdown of it from its rated operation state. (author)

  16. Computational analysis of Bangladesh 3 MW TRIGA research reactor using MCNP4C, JENDL-3.3 and ENDF/B-Vl data libraries

    International Nuclear Information System (INIS)

    Huda, M.Q.

    2006-01-01

    The three-dimensional continuous energy Monte Carlo code MCNP4C was used to develop a versatile and accurate full-core model of the 3 MW TRIGA MARK II research reactor at Atomic Energy Research Establishment, Savar, Dhaka, Bangladesh. The model represents in detail all components of the core with literally no physical approximation. All fresh fuel and control elements as well as the vicinity of the core were precisely described. Validation of the JENDL-3.3 and ENDF/BVI continuous energy cross-section data for MCNP4C was performed against some well-known benchmark lattices. For TRIGA analysis, data from JENDL-3.3 and ENDF/B-VI in combination with the JENDL-3.2 and ENDF/B-V data files (for nat Zr, nat Mo, nat Cr, nat Fe, nat Ni, nat Si, and nat Mg) at 300 K evaluations were used. Full S(α, β) scattering functions from ENDF/B-V for Zr in ZrH, H in ZrH and water molecule, and for graphite were used in both cases. The validation of the model was performed against the criticality and reactivity benchmark experiments of the TRIGA reactor. There is ∼20.0% decrease of thermal neutron flux occurs when the thermal library is removed during the calculation. Effect of erbium isotope that is present in the TRIGA fuel was also studied. In addition to the effective multiplication values, the well-known integral parameters: δ 28 , δ 25 , ρ 25 , and C * were calculated and compared for both JENDL3.3 and ENDF/B-VI libraries and were found to be in very good agreement. Results are also reported for most of the analyses performed by JENDL-3.2 and ENDF/B-V data libraries

  17. The Integral Fast Reactor

    International Nuclear Information System (INIS)

    Chang, Y.I.

    1988-01-01

    The Integral Fast Reactor (IFR) is an innovative liquid metal reactor concept being developed at Argonne National Laboratory. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system. This paper describes the key features and potential advantages of the IFR concept, with emphasis on its safety characteristics. 3 refs., 4 figs., 1 tab

  18. Mirror hybrid reactor optimization studies

    International Nuclear Information System (INIS)

    Bender, D.J.

    1976-01-01

    A system model of the mirror hybrid reactor has been developed. The major components of the model include (1) the reactor description, (2) a capital cost analysis, (3) various fuel management schemes, and (4) an economic analysis that includes the hybrid plus its associated fission burner reactors. The results presented describe the optimization of the mirror hybrid reactor, the objective being to minimize the cost of electricity from the hybrid fission-burner reactor complex. We have examined hybrid reactors with two types of blankets, one containing natural uranium, the other thorium. The major difference between the two optimized reactors is that the uranium hybrid is a significant net electrical power producer, whereas the thorium hybrid just about breaks even on electrical power. Our projected costs for fissile fuel production are approximately 50 $/g for 239 Pu and approximately 125 $/g for 233 U

  19. Comparison between TRU burning reactors and commercial fast reactor

    International Nuclear Information System (INIS)

    Fujimura, Koji; Sanda, Toshio; Ogawa, Takashi

    2001-03-01

    Research and development for stabilizing or shortening the radioactive wastes including in spent nuclear fuel are widely conducted in view point of reducing the environmental impact. Especially it is effective way to irradiate and transmute long-lived TRU by fast reactors. Two types of loading way were previously proposed. The former is loading relatively small amount of TRU in all commercial fast reactors and the latter is loading large amount of TRU in a few TRU burning reactors. This study has been intended to contribute to the feasibility studies on commercialized fast reactor cycle system. The transmutation and nuclear characteristics of TRU burning reactors were evaluated and compared with those of conventional transmutation system using commercial type fast reactor based upon the investigation of technical information about TRU burning reactors. Major results are summarized as follows. (1) Investigation of technical information about TRU burning reactors. Based on published reports and papers, technical information about TRU burning reactor concepts transmutation system using convectional commercial type fast reactors were investigated. Transmutation and nuclear characteristics or R and D issue were investigated based on these results. Homogeneously loading of about 5 wt% MAs on core fuels in the conventional commercial type fast reactor may not cause significant impact on the nuclear core characteristics. Transmutation of MAs being produced in about five fast reactors generating the same output is feasible. The helium cooled MA burning fast reactor core concept propose by JAERI attains criticality using particle type nitride fuels which contain more than 60 wt% MA. This reactor could transmute MAs being produced in more than ten 1000 MWe-LWRs. Ultra-long life core concepts attaining more than 30 years operation without refueling by utilizing MA's nuclear characteristics as burnable absorber and fertile nuclides were proposed. Those were pointed out that

  20. 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.

  1. Reactor Simulations for Safeguards with the MCNP Utility for Reactor Evolution Code

    International Nuclear Information System (INIS)

    Shiba, T.; Fallot, M.

    2015-01-01

    To tackle nuclear material proliferation, we conducted several proliferation scenarios using the MURE (MCNP Utility for Reactor Evolution) code. The MURE code, developed by CNRS laboratories, is a precision, open-source code written in C++ that automates the preparation and computation of successive MCNP (Monte Carlo N-Particle) calculations and solves the Bateman equations in between, for burnup or thermal-hydraulics purposes. In addition, MURE has been completed recently with a module for the CHaracterization of Radioactive Sources, called CHARS, which computes the emitted gamma, beta and alpha rays associated to any fuel composition. Reactor simulations could allow knowing how plutonium or other material generation evolves inside reactors in terms of time and amount. The MURE code is appropriate for this purpose and can also provide knowledge on associated particle emissions. Using MURE, we have both developed a cell simulation of a typical CANDU reactor and a detailed model of light water PWR core, which could be used to analyze the composition of fuel assemblies as a function of time or burnup. MURE is also able to provide, thanks to its extension MURE-CHARTS, the emitted gamma rays from fuel assemblies unloaded from the core at any burnup. Diversion cases of Generation IV reactors have been also developed; a design of Very High Temperature Reactor (a Pebble Bed Reactor (PBR), loaded with UOx, PuOx and ThUOx fuels), and a Na-cooled Fast Breeder Reactor (FBR) (with depleted Uranium or Minor Actinides in the blanket). The loading of Protected Plutonium Production (P3) in the FBR was simulated. The simulations of various reactor designs taking into account reactor physics constraints may bring valuable information to inspectors. At this symposium, we propose to show the results of these reactor simulations as examples of the potentiality of reactor simulations for safeguards. (author)

  2. On reactor type comparisons for the next generation of reactors

    International Nuclear Information System (INIS)

    Alesso, H.P.; Majumdar, K.C.

    1991-01-01

    In this paper, we present a broad comparison of studies for a selected set of parameters for different nuclear reactor types including the next generation. This serves as an overview of key parameters which provide a semi-quantitative decision basis for selecting nuclear strategies. Out of a number of advanced reactor designs of the LWR type, gas cooled type, and FBR type, currently on the drawing board, the Advanced Light Water Reactors (ALWR) seem to have some edge over other types of the next generation of reactors for the near-term application. This is based on a number of attributes related to the benefit of the vast operating experience with LWRs coupled with an estimated low risk profile, economics of scale, degree of utilization of passive systems, simplification in the plant design and layout, modular fabrication and manufacturing. 32 refs., 1 fig., 3 tabs

  3. Pressurized water reactors: the EPR project

    International Nuclear Information System (INIS)

    Py, J.P.; Yvon, M.

    2007-01-01

    EPR (originally 'European pressurized water reactor', and now 'evolutionary power reactor') is a model of reactor initially jointly developed by French and German engineers which fulfills the particular safety specifications of both countries but also the European utility requirements jointly elaborated by the main European power companies under the initiative of Electricite de France (EdF). Today, two EPR-based reactors are under development: one is under construction in Finland and the other, Flamanville 3 (France), received its creation permit decree on April 10, 2007. This article presents, first, the main objectives of the EPR, and then, describes the Flamanville 3 reactor: reactor type and general conditions, core and conditions of operation, primary and secondary circuits with their components, main auxiliary and recovery systems, man-machine interface and instrumentation and control system, confinement and serious accidents, arrangement of buildings. (J.S.)

  4. Cermet fuel reactors

    International Nuclear Information System (INIS)

    Cowan, C.L.; Palmer, R.S.; Van Hoomissen, J.E.; Bhattacharyya, S.K.; Barner, J.O.

    1987-09-01

    Cermet fueled nuclear reactors are attractive candidates for high performance space power systems. The cermet fuel consists of tungsten-urania hexagonal fuel blocks characterized by high strength at elevated temperatures, a high thermal conductivity and resultant high thermal shock resistance. Key features of the cermet fueled reactor design are (1) the ability to achieve very high coolant exit temperatures, and (2) thermal shock resistance during rapid power changes, and (3) two barriers to fission product release - the cermet matrix and the fuel element cladding. Additionally, thre is a potential for achieving a long operating life because of (1) the neutronic insensitivity of the fast-spectrum core to the buildup of fission products and (2) the utilization of a high strength refractory metal matrix and structural materials. These materials also provide resistance against compression forces that potentially might compact and/or reconfigure the core. In addition, the neutronic properties of the refractory materials assure that the reactor remains substantially subcritical under conditions of water immersion. It is concluded that cermet fueled reactors can be utilized to meet the power requirements for a broad range of advanced space applications. 4 refs., 4 figs., 3 tabs

  5. TREATMENT OF METHANOLIC WASTEWATER BY ANAEROBIC DOWN-FLOW HANGING SPONGE (ANDHS) REACTOR AND UASB REACTOR

    Science.gov (United States)

    Sumino, Haruhiko; Wada, Keiji; Syutsubo, Kazuaki; Yamaguchi, Takashi; Harada, Hideki; Ohashi, Akiyoshi

    Anaerobic down-flow hanging sponge (AnDHS) reactor and UASB reactor were operated at 30℃ for over 400 days in order to investigate the process performance and the sludge characteristics of treating methanolic wastewater (2 gCOD/L). The settings OLR of AnDHS reactor and of UASB reactor were 5.0 -10.0 kgCOD/m3/d and 5.0 kgCOD/m3/d. The average of the COD removal demonstrated by both reactors were over 90% throughout the experiment. From the results of methane producing activities and the PCR-DGGE method, most methanol was directly converted to methane in both reactors. The conversion was carried out by different methanogens: one closely related to Methanomethylovorans hollandica in the AnDHS retainted sludge and the other closely related to Methanosarcinaceae and Metanosarciales in the UASB retainted sludge.

  6. Feasibility study of a magnetic fusion production reactor

    Science.gov (United States)

    Moir, R. W.

    1986-12-01

    A magnetic fusion reactor can produce 10.8 kg of tritium at a fusion power of only 400 MW —an order of magnitude lower power than that of a fission production reactor. Alternatively, the same fusion reactor can produce 995 kg of plutonium. Either a tokamak or a tandem mirror production plant can be used for this purpose; the cost is estimated at about 1.4 billion (1982 dollars) in either case. (The direct costs are estimated at 1.1 billion.) The production cost is calculated to be 22,000/g for tritium and 260/g for plutonium of quite high purity (1%240Pu). Because of the lack of demonstrated technology, such a plant could not be constructed today without significant risk. However, good progress is being made in fusion technology and, although success in magnetic fusion science and engineering is hard to predict with assurance, it seems possible that the physics basis and much of the needed technology could be demonstrated in facilities now under construction. Most of the remaining technology could be demonstrated in the early 1990s in a fusion test reactor of a few tens of megawatts. If the Magnetic Fusion Energy Program constructs a fusion test reactor of approximately 400 MW of fusion power as a next step in fusion power development, such a facility could be used later as a production reactor in a spinoff application. A construction decision in the late 1980s could result in an operating production reactor in the late 1990s. A magnetic fusion production reactor (MFPR) has four potential advantages over a fission production reactor: (1) no fissile material input is needed; (2) no fissioning exists in the tritium mode and very low fissioning exists in the plutonium mode thus avoiding the meltdown hazard; (3) the cost will probably be lower because of the smaller thermal power required; (4) and no reprocessing plant is needed in the tritium mode. The MFPR also has two disadvantages: (1) it will be more costly to operate because it consumes rather than sells

  7. Calculation models for a nuclear reactor

    International Nuclear Information System (INIS)

    Tashanii, Ahmed Ali

    2010-01-01

    Determination of different parameters of nuclear reactors requires neutron transport calculations. Due to complicity of geometry and material composition of the reactor core, neutron calculations were performed for simplified models of the real arrangement. In frame of the present work two models were used for calculations. First, an elementary cell model was used to prepare cross section data set for a homogenized-core reactor model. The homogenized-core reactor model was then used to perform neutron transport calculation. The nuclear reactor is a tank-shaped thermal reactor. The semi-cylindrical core arrangement consists of aluminum made fuel bundles immersed in water which acts as a moderator as well as a coolant. Each fuel bundle consists of aluminum cladded fuel rods arranged in square lattices. (author)

  8. Evaluation of the trial design studies for an advanced marine reactor, (3)

    International Nuclear Information System (INIS)

    Ambo, Noriaki; Yokomura, Takeyoshi.

    1988-03-01

    JAERI have carried out four core designs for three different type reactors (Semi-Integrated, Integrated and Integrated (self-pressured) type reactors), as the trial designs of an Advanced Marine Reactor for three years (1983 ∼ 1985). This report describes the result of comparison and studies of the core specific characteristics of these four cores, which include core concept, specifications, core life, specific power density, burn-up, reactivity control and etc. In conclusion, it was found that the Integrated type reactor core and the Semi-Integrated type reactor core designs satisfy the conditions of long core life (four years), high specific power density (50 ∼ 61 kw/l) and high burn-up (30,000 ∼ 32,000 MWD/t), so these two cores will be optimum designs based on the present technologies. (author)

  9. Nuclear reactors to come

    International Nuclear Information System (INIS)

    Lung, M.

    2002-01-01

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

  10. Fusion reactor problems

    International Nuclear Information System (INIS)

    Carruthers, R.

    It is pointed out that plasma parameters for a fusion reactor have been fairly accurately defined for many years, and the real plasma physics objective must be to find the means of achieving and maintaining these specifiable parameters. There is good understanding of the generic technological problems: breading blankets and shields, radiation damage, heat transfer and methods of magnet design. The required plasma parameters for fusion self-heated reactors are established at ntausub(E) approximately 2.10 14 cm -3 sec, plasma radius 1.5 to 3 m, wall loading 5 to 10 MW cm -2 , temperature 15 keV. Within this model plasma control by quasi-steady burn as a key problem is studied. It is emphasized that the future programme must interact more closely with engineering studies and should concentrate upon research which is relevant to reactor plasmas. (V.P.)

  11. 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

  12. Development of a three dimension multi-physics code for molten salt fast reactor

    International Nuclear Information System (INIS)

    Cheng Maosong; Dai Zhimin

    2014-01-01

    Molten Salt Reactor (MSR) was selected as one of the six innovative nuclear reactors by the Generation IV International Forum (GIF). The circulating-fuel in the can-type molten salt fast reactor makes the neutronics and thermo-hydraulics of the reactor strongly coupled and different from that of traditional solid-fuel reactors. In the present paper: a new coupling model is presented that physically describes the inherent relations between the neutron flux, the delayed neutron precursor, the heat transfer and the turbulent flow. Based on the model, integrating nuclear data processing, CAD modeling, structured and unstructured mesh technology, data analysis and visualization application, a three dimension steady state simulation code system (MSR3DS) for the can-type molten salt fast reactor is developed and validated. In order to demonstrate the ability of the code, the three dimension distributions of the velocity, the neutron flux, the delayed neutron precursor and the temperature were obtained for the simplified MOlten Salt Advanced Reactor Transmuter (MOSART) using this code. The results indicate that the MSR3DS code can provide a feasible description of multi-physical coupling phenomena in can-type molten salt fast reactor. Furthermore, the code can well predict the flow effect of fuel salt and the transport effect of the turbulent diffusion. (authors)

  13. Dry storage of MTR spent fuel from the Argentine radioisotope production reactor RA-3

    International Nuclear Information System (INIS)

    Di Marco, A.; Gillaume, E.J.; Ruggirello, G.; Zaweruchi, A.

    1996-01-01

    The nuclear fuel elements of the RA-3 reactor consist in 19 rectangular fuel plates held in position by two lateral structural plates. The whole assembly is coupled to the lower nozzles that fits in the reactor core grid. The inner plates are 1.5 mm thick, 70.5 mm wide and 655 mm long and the outer plates are 100 mm longer. The fuel plates are formed by a core of an AI-U alloy co-laminated between two plates of Al. Enrichment is 90% 235 U. After being extracted from the reactor, the fuel elements have been let to cool down in the reactor storage pool and finally moved to the storage facility. This facility is a grid of vertical underground channels connected by a piping system. The system is filled with processed and controlled water. At the present the storage capacity of the facility is near to be depleted and some indications of deterioration of the fuel elements has been detected. Due to the present status of the facility and the spent fuel stored there, a decision has been taken to proceed to modify the present underwater storage to dry storage. The project consist in: a) Decontamination and conditioning of the storage channels to prepare them for dry storage. b) Disassembly of the fuel elements in hot cells in order to can only the active fuel plates in an adequate tight canister. c) The remnant structural pieces will be treated as low level waste. (author). 10 figs

  14. Modelling of the anti-neutrino production and spectra from a Magnox reactor

    Science.gov (United States)

    Mills, Robert W.; Mountford, David J.; Coleman, Jonathon P.; Metelko, Carl; Murdoch, Matthew; Schnellbach, Yan-Jie

    2018-01-01

    The anti-neutrino source properties of a fission reactor are governed by the production and beta decay of the radionuclides present and the summation of their individual anti-neutrino spectra. The fission product radionuclide production changes during reactor operation and different fissioning species give rise to different product distributions. It is thus possible to determine some details of reactor operation, such as power, from the anti-neutrino emission to confirm safeguards records. Also according to some published calculations, it may be feasible to observe different anti-neutrino spectra depending on the fissile contents of the reactor fuel and thus determine the reactor's fissile material inventory during operation which could considerable improve safeguards. In mid-2014 the University of Liverpool deployed a prototype anti-neutrino detector at the Wylfa R1 station in Anglesey, United Kingdom based upon plastic scintillator technology developed for the T2K project. The deployment was used to develop the detector electronics and software until the reactor was finally shutdown in December 2015. To support the development of this detector technology for reactor monitoring and to understand its capabilities, the National Nuclear Laboratory modelled this graphite moderated and natural uranium fuelled reactor with existing codes used to support Magnox reactor operations and waste management. The 3D multi-physics code PANTHER was used to determine the individual powers of each fuel element (8×6152) during the year and a half period of monitoring based upon reactor records. The WIMS/TRAIL/FISPIN code route was then used to determine the radionuclide inventory of each nuclide on a daily basis in each element. These nuclide inventories were then used with the BTSPEC code to determine the anti-neutrino spectra and source strength using JEFF-3.1.1 data. Finally the anti-neutrino source from the reactor for each day during the year and a half of monitored reactor

  15. Modelling of the anti-neutrino production and spectra from a Magnox reactor

    Directory of Open Access Journals (Sweden)

    Mills Robert W

    2018-01-01

    Full Text Available The anti-neutrino source properties of a fission reactor are governed by the production and beta decay of the radionuclides present and the summation of their individual anti-neutrino spectra. The fission product radionuclide production changes during reactor operation and different fissioning species give rise to different product distributions. It is thus possible to determine some details of reactor operation, such as power, from the anti-neutrino emission to confirm safeguards records. Also according to some published calculations, it may be feasible to observe different anti-neutrino spectra depending on the fissile contents of the reactor fuel and thus determine the reactor's fissile material inventory during operation which could considerable improve safeguards. In mid-2014 the University of Liverpool deployed a prototype anti-neutrino detector at the Wylfa R1 station in Anglesey, United Kingdom based upon plastic scintillator technology developed for the T2K project. The deployment was used to develop the detector electronics and software until the reactor was finally shutdown in December 2015. To support the development of this detector technology for reactor monitoring and to understand its capabilities, the National Nuclear Laboratory modelled this graphite moderated and natural uranium fuelled reactor with existing codes used to support Magnox reactor operations and waste management. The 3D multi-physics code PANTHER was used to determine the individual powers of each fuel element (8×6152 during the year and a half period of monitoring based upon reactor records. The WIMS/TRAIL/FISPIN code route was then used to determine the radionuclide inventory of each nuclide on a daily basis in each element. These nuclide inventories were then used with the BTSPEC code to determine the anti-neutrino spectra and source strength using JEFF-3.1.1 data. Finally the anti-neutrino source from the reactor for each day during the year and a half of

  16. The integral fast reactor concept

    International Nuclear Information System (INIS)

    Chang, Yoon I.; Marchaterre, J.F.

    1987-01-01

    The Integral Fast Reactor (IFR) is an innovative liquid metal reactor concept being developed at Argonne National Laboratory. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) an integral fuel cycle, based on pyrometallurgical processing and injection-cast fuel fabrication, with the fuel cycle facility collocated with the reactor, if so desired. This paper gives a review of the IFR concept

  17. Reactor container cooling device

    Energy Technology Data Exchange (ETDEWEB)

    Ando, Koji; Kinoshita, Shoichiro

    1995-11-10

    The device of the present invention efficiently lowers pressure and temperature in a reactor container upon occurrence of a severe accident in a BWR-type reactor and can cool the inside of the container for a long period of time. That is, (1) pipelines on the side of an exhaustion tower of a filter portion in a filter bent device of the reactor container are in communication with pipelines on the side of a steam inlet of a static container cooling device by way of horizontal pipelines, (2) a back flow check valve is disposed to horizontal pipelines, (3) a steam discharge valve for a pressure vessel is disposed closer to the reactor container than the joint portion between the pipelines on the side of the steam inlet and the horizontal pipelines. Upon occurrence of a severe accident, when the pressure vessel should be ruptured and steams containing aerosol in the reactor core should be filled in the reactor container, the inlet valve of the static container cooling device is closed. Steams are flown into the filter bent device of the reactor container, where the aerosols can be removed. (I.S.).

  18. Characteristics of Flameless Combustion in 3D Highly Porous Reactors under Diesel Injection Conditions

    Directory of Open Access Journals (Sweden)

    M. Weclas

    2013-01-01

    Full Text Available The heat release process in a free volume combustion chamber and in porous reactors has been analyzed under Diesel engine-like conditions. The process has been investigated in a wide range of initial pressures and temperatures simulating engine conditions at the moment when fuel injection starts. The resulting pressure history in both porous reactors and in free volumes significantly depends on the initial pressure and temperature. At lower initial temperatures, the process in porous reactors is accelerated. Combustion in a porous reactor is characterized by heat accumulation in the solid phase of the porous structure and results in reduced pressure peaks and lowered combustion temperature. This depends on reactor heat capacity, pore density, specific surface area, pore structure, and heat transport properties. Characteristic modes of a heat release process in a two-dimensional field of initial pressure and temperature have been selected. There are three characteristic regions represented by a single- and multistep oxidation process (with two or three slopes in the reaction curve and characteristic delay time distribution has been selected in five characteristic ranges. There is a clear qualitative similarity of characteristic modes of the heat release process in a free volume and in porous reactors. A quantitative influence of porous reactor features (heat capacity, pore density, pore structure, specific surface area, and fuel distribution in the reactor volume has been clearly indicated.

  19. Refurbishment, Modernization and Ageing Management Program of The 3MW TRIGA Mark-II Research Reactor of Bangladesh

    International Nuclear Information System (INIS)

    Salam, M. A.

    2013-01-01

    The 3 MW TRIGA MK-II research reactor of Bangladesh Atomic Energy Commission (BAEC) achieved its first criticality on 14 September 1986. The reactor has been used for manpower training, radioisotope production and various R and D activities in the field of neutron activation analysis, neutron radiography and neutron scattering. Reactor Operation and Maintenance Unit (ROMU) is responsible for operation and maintenance of the research reactor. During the past twenty seven years ROMU carried out several refurbishments, replacement, modification and modernization activities in the reactor facility. The major tasks carried out under refurbishment program were replacement of the corrosion damaged N-16 decay tank by a new one, replacement of the fouled shell and tube type heat exchanger by a plate type one, modification of the shielding arrangements around the N-16 decay tank and ECCS system and solving the radial beam port-1 leakage problem. All of these refurbishment activities were performed under an annual development project (ADP) funded by Bangladesh government. BAEC research reactor (RR) was operated by analogue console system from its commissioning to July, 2011. Old analog based console has been replaced by digital console on June, 2012. Modernization program for the reactor control console due to obsolescence and unavailability of spare parts of I and C system was vital to restore the safe operation of the reactor. Considering these facts, installation of a digital control console and I and C system based on the state-of-the-art digital technology became necessary. Reactor digital console system installation tasks were performed under another ADP funded project by Bangladesh government. Now the reactor is operating with the digital control system. Besides this, the Neutron Radiography (NR) facility has been modernized by the addition of a digital neutron radiography set-up at the tangential beam port. The Neutron Scattering (NS) facility also has been upgraded

  20. Refurbishment, Modernization and Ageing Management Program of The 3MW TRIGA Mark-II Research Reactor of Bangladesh

    Energy Technology Data Exchange (ETDEWEB)

    Salam, M. A. [Atomic Energy Research Establishment, Dhaka (Bangladesh)

    2013-07-01

    The 3 MW TRIGA MK-II research reactor of Bangladesh Atomic Energy Commission (BAEC) achieved its first criticality on 14 September 1986. The reactor has been used for manpower training, radioisotope production and various R and D activities in the field of neutron activation analysis, neutron radiography and neutron scattering. Reactor Operation and Maintenance Unit (ROMU) is responsible for operation and maintenance of the research reactor. During the past twenty seven years ROMU carried out several refurbishments, replacement, modification and modernization activities in the reactor facility. The major tasks carried out under refurbishment program were replacement of the corrosion damaged N-16 decay tank by a new one, replacement of the fouled shell and tube type heat exchanger by a plate type one, modification of the shielding arrangements around the N-16 decay tank and ECCS system and solving the radial beam port-1 leakage problem. All of these refurbishment activities were performed under an annual development project (ADP) funded by Bangladesh government. BAEC research reactor (RR) was operated by analogue console system from its commissioning to July, 2011. Old analog based console has been replaced by digital console on June, 2012. Modernization program for the reactor control console due to obsolescence and unavailability of spare parts of I and C system was vital to restore the safe operation of the reactor. Considering these facts, installation of a digital control console and I and C system based on the state-of-the-art digital technology became necessary. Reactor digital console system installation tasks were performed under another ADP funded project by Bangladesh government. Now the reactor is operating with the digital control system. Besides this, the Neutron Radiography (NR) facility has been modernized by the addition of a digital neutron radiography set-up at the tangential beam port. The Neutron Scattering (NS) facility also has been upgraded

  1. Nuclear reactor physics course for reactor operators

    International Nuclear Information System (INIS)

    Baeten, P.

    2006-01-01

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

  2. Oklo natural reactor

    International Nuclear Information System (INIS)

    Fujii, Isao

    1985-01-01

    In 1954, Professor Kazuo, Kuroda of Arkansas University in USA published the possibility that spontaneously generated natural nuclear reactors existed in prehistoric age. In 1972, 18 years after that, Commissariat a l'Energie Atomique published that in the Oklo uranium deposit in Gabon, Africa, a natural nuclear reactor was found. This fact was immediately informed to the whole world, but in Japan, its details have not necessarily been well known. The chance of investigating into this fact and visiting the Oklo deposit by the favor of COMUF, the owner of the Oklo deposit, was given, therefore, the state of the natural reactors, which has been known so far, is reported. At present, 12 natural reactors have been found in the vicinity of the Oklo deposit. The natural reactors were generated spontaneously in uranium deposits about 1.7 billion years ago when the isotopic abundance of U-235 was 3 %, and the chain reaction started naturally. When the concentration of U-235 lowered, the reaction stopped naturally. The abnormality in the U-235 abundance in natural uranium was found, and the cause was pursued. The evidence of the existence of natural reactors was shown. (Kako, I.)

  3. Future fuel cycle and reactor strategies. Key issue paper no. 3

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-06-01

    The scope of this paper includes those issues that are expected to remain or become important in the time period from 2015 to 2050. Events in this time frame are difficult to predict with any certainty, so the framework of this paper is necessarily somewhat speculative. The paper includes consideration of all facets of nuclear energy utilization, from ore extraction to final disposal of waste products. The paper first addresses the factors influencing the choice of reactor and fuel cycle. It then goes on to address the quantitatively largest category of reactor types expected to be important during the period; that is, thermal reactors burning uranium and plutonium fuel in various forms. The fast reactor type is then discussed both as stand-alone technology and as technology used in combination with thermal reactors. Thorium fuel use is discussed briefly. This paper is concentrated on the ``medium variant`` energy growth scenario identified in Key Issue Paper No. 1. The effects of either higher or lower growth could, of course, profoundly change the future development of the nuclear power industry. 31 refs, 5 figs, 4 tabs.

  4. Reactor core and initially loaded reactor core of nuclear reactor

    International Nuclear Information System (INIS)

    Koyama, Jun-ichi; Aoyama, Motoo.

    1989-01-01

    In BWR type reactors, improvement for the reactor shutdown margin is an important characteristic condition togehter with power distribution flattening . However, in the reactor core at high burnup degree, the reactor shutdown margin is different depending on the radial position of the reactor core. That is , the reactor shutdown margin is smaller in the outer peripheral region than in the central region of the reactor core. In view of the above, the reactor core is divided radially into a central region and as outer region. The amount of fissionable material of first fuel assemblies newly loaded in the outer region is made less than the amount of the fissionable material of second fuel assemblies newly loaded in the central region, to thereby improve the reactor shutdown margin in the outer region. Further, the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower portion of the first fuel assemblies is made smaller than the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower region of the second fuel assemblies, to thereby obtain a sufficient thermal margin in the central region. (K.M.)

  5. EPR, a GEN 3 Reactor providing a competitive electricity cost

    International Nuclear Information System (INIS)

    Salhi, Othman

    2006-01-01

    Since the very beginning of the development of what was to become the EPR, several European entities were involved. The French and German safety authorities expressed that reinforced safety was compulsory. Additional measures were then included to prevent the occurrence of events likely to damage the core, and reduce the possibility of exposure of operating and maintenance personnel. However, not with standing these safety related features resulting from the requirements of the safety authorities, we will focus today on another group of entities that were key players in EPR development: the Utilities. The Utilities voiced their need for a competitive electricity produced and a competitive nuclear reactor. The tradeoff was then to reach both targets in a unique product: a safer and more competitive NPP. Today, the EPR presents features that enable our clients to compete with the cheapest fossil-based electricity production plants. Increased thermal efficiency is obtained both through a higher steam pressure and through careful optimization of the secondary system thermal cycle

  6. Neutronics of a mixed-flow gas-core reactor

    International Nuclear Information System (INIS)

    Soran, P.D.; Hansen, G.E.

    1977-11-01

    The study was made to investigate the neutronic feasibility of a mixed-flow gas-core reactor. Three reactor concepts were studied: four- and seven-cell radial reactors and a seven-cell scallop reactor. The reactors were fueled with UF 6 (either U-233 or U-235) and various parameters were varied. A four-cell reactor is not practical nor is the U-235 fueled seven-cell radial reactor; however, the 7-cell U-233 radial and scallop reactors can satisfy all design criteria. The mixed flow gas core reactor is a very attractive reactor concept and warrants further investigation

  7. The zero power reactor SUR and its application

    International Nuclear Information System (INIS)

    Wesser, U.

    1986-01-01

    This low-power reactor, rated nominally at 100 milliwatts, has a cylindrical core of 26 cm in diameter and 24 cm high consisting of U 3 O 8 powder in a polyethylene matrix. The fuel is 20 percent enriched and the critical mass about 700 g. The excess reactivity is about 3 mk. The reactivity is controlled by two cadmium sheets in addition to a back-up system that drops the inner reflector. The reactor has no active cooling system. Personnel costs include a supervisor and an operator. The reactor is used for training in Reactor Theory (including use of a neutron chopper), reactor kinetics, nuclear technology, reactor operations and for doctoral thesis research. (author)

  8. A new fluidized bed nuclear reactor

    International Nuclear Information System (INIS)

    Sefidvash, F.

    1986-01-01

    A new nuclear reactor design based on the fluidized bed concept is proposed. A current design utilizes spherical fuel of slightly enriched Zircaloy-clad uranium dioxide fluidized by light water under pressure. The reactor is modular in system; therefore, any size reactor can be constructed from the basic standard modul. The reactor physics calculations show that reactivity increases with porosity to a maximum value and thereafter decreases. This produces inherent safety and eliminates the need for control rods and burnable poisons. The heat transfer calculations show that the maximum power extracted from the reactor core is not limited to the material temperature limits but to the maximum mass flow of coolant, which corresponds to the desired operating porosity. Design simplicity and inherent safety make it an attractive small reactor design. (Author) [pt

  9. Nuclear reactor built, being built, or planned

    International Nuclear Information System (INIS)

    1991-06-01

    This document contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of December 31, 1990. The Office of Scientific and Technical Information, US Department of Energy, gathers this information annually from Washington headquarters and field offices of DOE, from the US Nuclear Regulatory Commission, from the US reactor manufacturers who are the principal nuclear contractors for foreign reactor locations, from US and foreign embassies, and from foreign governmental nuclear departments. The book is divided into three major sections: Section 1 consists of a reactor locator map and reactor tables; Section 2 includes nuclear reactors that are operating, being built, or planned; and Section 3 includes reactors that have been shut down permanently or dismantled. Sections 2 and 3 contain the following classification of reactors: Civilian, Production, Military, Export, and Critical Assembly

  10. The pebble-bed high-temperature reactor as a source of nuclear process heat. Vol. 3

    International Nuclear Information System (INIS)

    Kugeler, K.; Schulten, R.; Kugeler, M.; Niessen, H.F.; Roeth-Kamat, M.; Hohn, H.; Woike, O.; Germer, J.H.

    1974-08-01

    The characteristic questions concerning a process heat reactor with high helium outlet temperatures are dealt with in this volume like e.g. fuel element design, corrosion, and fission product release. Furthermore, some possibilities of the technical realization of the hot-gas ducting and intermediate heat exchangers are described. Important parameters for the design of the reactor such as core power density, helium inlet and outlet temperatures, helium pressure and fuel cycle burn-up and conversion and the effect of these on the primary circuit are investigated. The important question regarding which reactor vessel is to be chosen for nuclear process heat plants is discussed with the aid of the integrated and non-integrated concepts using prestressed concrete, cast iron and cast steel. Thereafter, considerations on the safety of the nuclear plant are given. Finally, mention is made of the availability of the nuclear plant and of the status of development of the HTR technology. (orig.) [de

  11. The integral fast reactor

    International Nuclear Information System (INIS)

    Till, C.E.

    1987-01-01

    On April 3rd, 1986, two demonstrations of the inherent capability of sodium-cooled fast reactors to survive unprotected loss of cooling accidents were carried out on the experimental sodium-cooled power reactor, EBR-II, on the Idaho site of Argonne National Laboratory. Transients potentially of the most serious kind, one an unprotected loss of flow, the other an unprotected loss of heat sink, both initiated from full power. In both cases the reactor quietly shut itself down, without damage of any kind. These tests were a part of the on-going development program at Argonne to develop an advanced reactor with significant new inherent safety characteristics. Called the integral fast reactor, or IFR, the basic thrust is to develop everything that is needed for a complete nuclear power system - reactor, closed fuel cycle, and waste processing - as a single optimized entity, and, for simplicity in concept, as an integral part of a single plant. The particular selection of reactor materials emphasizes inherent safety characteristics also makes possible a simplified close fuel cycle and waste process improvements. The paper describes the IFR concept, the inherent safety, tests, and status of IFR development today

  12. Multiple Irradiation Capsule Experiment (MICE)-3B Irradiation Test of Space Fuel Specimens in the Advanced Test Reactor (ATR) - Close Out Documentation for Naval Reactors (NR) Information

    Energy Technology Data Exchange (ETDEWEB)

    M. Chen; CM Regan; D. Noe

    2006-01-09

    Few data exist for UO{sub 2} or UN within the notional design space for the Prometheus-1 reactor (low fission rate, high temperature, long duration). As such, basic testing is required to validate predictions (and in some cases determine) performance aspects of these fuels. Therefore, the MICE-3B test of UO{sub 2} pellets was designed to provide data on gas release, unrestrained swelling, and restrained swelling at the upper range of fission rates expected for a space reactor. These data would be compared with model predictions and used to determine adequacy of a space reactor design basis relative to fission gas release and swelling of UO{sub 2} fuel and to assess potential pellet-clad interactions. A primary goal of an irradiation test for UN fuel was to assess performance issues currently associated with this fuel type such as gas release, swelling and transient performance. Information learned from this effort may have enabled use of UN fuel for future applications.

  13. Characteristics of irradiation creep in the first wall of a fusion reactor

    International Nuclear Information System (INIS)

    Coghlan, W.A.; Mansur, L.K.

    1981-01-01

    A number of significant differences in the irradiation environment of a fusion reactor are expected with respect to the fission reactor irradiation environment. These differences are expected to affect the characteristics of irradiation creep in the fusion reactor. Special conditions of importance are identified as the (1) large number of defects produced per pka, (2) high helium production rate, (3) cyclic operation, (4) unique stress histories, and (5) low temperature operations. Existing experimental data from the fission reactor environment is analyzed to shed light on irradiation creep under fusion conditions. Theoretical considerations are used to deduce additional characteristics of irradiation creep in the fusion reactor environment for which no experimental data are available

  14. A review of fast reactor program in Japan

    International Nuclear Information System (INIS)

    1996-01-01

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

  15. RA Reactor operation and maintenance (I-IX), part VII, Task 3.08/04, Refurbishment of the RA reactor

    International Nuclear Information System (INIS)

    Zecevic, V.

    1963-12-01

    This volume covers the following reports concerned with the maintenance and repair work of the RA reactor: repair of the technical water system; maintenance of the transportation equipment; vacuuming and drying during refurbishment; repair and decontamination of the distillation device; and the report on participation of the operational dosimetry division in the RA reactor refurbishment activities

  16. A nuclear desalination complex with a