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Sample records for argonaut type reactors

  1. Argonaut type reactor for the best possible Phase Ia training of nuclear plant operators

    International Nuclear Information System (INIS)

    The Argonaut type reactor is an excellent training tool for the training of Electric Utility Nuclear Plant Operators. The training advantages of this type of reactor can best be seen by comparing its design characteristics to a typical large pressurized water reactor and other research/training reactors not necessary for reactor operator training are explained. Some minor modifications of the Argonaut at UCLA would prove valuable and are under consideration. A complete one week Phase Ia training program proposal has been made by UCLA to selected utilities and a summary of this program is presented

  2. Decommissioning of an argonaut type reactor at the Technical University of Catalonia in Barcelona (Spain)

    International Nuclear Information System (INIS)

    The reactor ARGOS is a training nuclear reactor that was active, from 1962 to 1976, at the Technical University of Catalonia (UPC) in Barcelona (Spain). It is an Argonaut type experimental Reactor with 10 kW of maximal thermal power, and was set up by the main Spanish Nuclear Research Centre, presently named CIEMAT, in the period 1958-1962. In 1977, the nuclear installation was halted for technical, economical and administrative reasons. The fuel burn-up of the reactor was 2.7 kWh. In 1992 the fuel was removed from the site and a dismantling project was launched by an academic team of the UPC Nuclear Energy Department. In 1998 the Spanish authorities approved the dismantling plan which was based on the IAEA document Planning and Management of the Decommissioning of Research and Other Small Nuclear Facilities, IAEA 1993. In this plan the University proposed to set up its own dismantling group mainly based on its own academic staff and experimental facilities

  3. The Siemens-Argonaut reactor as a driver zone for a high-temperature reactor cell. Der Siemens-Argonaut-Reaktor als Treiberzone fuer eine Hochtemperaturreaktorzelle

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, H.; Schuerrer, F.; Ninaus, W.; Oswald, K.; Rabitsch, H.; Kreiner, H. (Technische Univ., Graz (Austria). Inst. fuer Theoretische Physik und Reaktorphysik); Neef, R.D. (Kernforschungsanlage Juelich G.m.b.H. (Germany, F.R.). Inst. fuer Reaktorentwicklung)

    1984-12-15

    To enable a validation of neutron physics calculation methods for pebble bed reactors the inner reflector of an Argonaut research reactor was substituted by a full of about 1200 fuel elements of the AVR-Juelich type. The report describes the measuring instruments and the reactor physical layout of the arrangement by the code packages GAMTEREX, ZUT-D.G.L. and MUPO. Comparison of calculated reaction rates with measurements show good agreement. Application of the codes to high-temperature reactors in abnormal states is envisaged. (Author, translated by G.Q.)

  4. Report on the interpretation of critical experiments in the Siemens-Argonaut-Reactor Graz to study water ingress into spherical elements. Ergebnisbericht zur Auslegung kritischer Experimente am Siemens-Argonaut-Reaktor Graz zum Studium des Wassereinbruches im Kugelhaufen

    Energy Technology Data Exchange (ETDEWEB)

    Schuerrer, F. (Technische Univ., Graz (Austria). Inst. fuer Theoretische Physik und Reaktorphysik); Neef, R.D. (Kernforschungsanlage Juelich G.m.b.H. (Germany, F.R.). Inst. fuer Reaktorentwicklung)

    1979-04-15

    The experiments described are of interest in the study of water contamination in HTR fuel elements. The Siemens Argonaut Reactor (SAR) has been considered as a research tool for a simulation experiment. Following a brief description of the SAR, planned programs are discussed in 'dry' and 'wet' cores. Detector foil types and locations are noted. A theoretical model is developed and nuclide concentrations estimated in the various spectral zones. Reactivity calculations have been made and are summarised for various H{sub 2}O percentage concentrations. The discussion is supported by simplified core layout diagrams and graphs of core flux distributions. Neutron diffusion and spectra calculations are referenced to computer programs used by KFA-Juelich, published elsewhere, and include GAM, THERMOS, MUPO and EXTERMINATOR-2. (G.C.)

  5. LMFBR type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kanbe, Mitsuru

    1997-04-04

    An LMFBR type reactor comprises a plurality of reactor cores in a reactor container. Namely, a plurality of pot containing vessels are disposed in the reactor vessel and a plurality of reactor cores are formed in a state where an integrated-type fuel assembly is each inserted to a pot, and a coolant pipeline is connected to each of the pot containing-vessel to cool the reactor core respectively. When fuels are exchanged, the integrated-type fuel assembly is taken out together with the pot from the reactor vessel in a state where the integrated-type fuel assembly is immersed in the coolants in the pot as it is. Accordingly, coolants are supplied to each of the pot containing-vessel connected with the coolant pipeline and circulate while cooling the integrated-type fuel assembly for every pot. Then, when the fuels are exchanged, the integrated type fuel assembly is taken out to the outside of the reactor together with the pot by taking up the pot from the pot-containing vessel. Then, neutron economy is improved to thereby improve reactor power and the breeding ratio. (N.H.)

  6. BWR type nuclear reactor

    International Nuclear Information System (INIS)

    Purpose: To simplify the structure of an emergency core cooling system while suppressing the flow out of coolants upon rapture accidents in a coolant recycling device of BWR type reactors. Constitution: Recirculation pumps are located at a position higher than the reactor core in a pressure vessel, and the lower plenum is bisected vertically by a partition plate. Further, a gas-liquid separator is surrounded with a wall and the water level at the outer side of the wall is made higher than the water level in the inside of the wall. In this structure, coolants are introduced from the upper chamber in the lower plenum into the reactor core, and the steams generated in the reactor core are separated in the gas-liquid separator, whereby the separated liquid is introduced as coolants by way of the inner chamber into the lower chamber of the lower plenum and further sent by way of the outer chamber into the reactor core. Consequently, idle rotation of the recycling pumps due to the flow-in of saturated water is prevented and loss of coolants in the reactor core can also be prevented upon raptures in the pipeway and the driving section of the pump connected to the pressure vessel and in the bottom of the pressure vessel. (Horiuchi, T.)

  7. Argonaute and Argonaute-Bound Small RNAs in Stem Cells

    Directory of Open Access Journals (Sweden)

    Lihong Zhai

    2016-02-01

    Full Text Available Small RNAs are essential for a variety of cellular functions. Argonaute (AGO proteins are associated with all of the different classes of small RNAs, and are indispensable in small RNA-mediated regulatory pathways. AGO proteins have been identified in various types of stem cells in diverse species from plants and animals. This review article highlights recent progress on how AGO proteins and AGO-bound small RNAs regulate the self-renewal and differentiation of distinct stem cell types, including pluripotent, germline, somatic, and cancer stem cells.

  8. LMFBR type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shimizu, Takeshi; Iida, Masaaki; Moriki, Yasuyuki

    1994-10-18

    A reactor core is divided into a plurality of coolants flowrate regions, and electromagnetic pumps exclusively used for each of the flowrate regions are disposed to distribute coolants flowrates in the reactor core. Further, the flowrate of each of the electromagnetic pumps is automatically controlled depending on signals from a temperature detector disposed at the exit of the reactor core, so that the flowrate of the region can be controlled optimally depending on the burning of reactor core fuels. Then, the electromagnetic pumps disposed for every divided region are controlled respectively, so that the coolants flowrate distribution suitable to each of the regions can be attained. Margin for fuel design is decreased, fuels are used effectively, as well as an operation efficiency can be improved. Moreover, since the electromagnetic pump has less flow resistance compared with a mechanical type pump, and flow resistance of the reactor core flowrate control mechanism is eliminated, greater circulating flowrate can be ensured after occurrence of accident in a natural convection using a buoyancy of coolants utilizable for after-heat removal as a driving force. (N.H.).

  9. Pressure tube type reactor

    International Nuclear Information System (INIS)

    Heretofore, a pressure tube type reactor has a problem in that the evaluation for the reactor core performance is complicate and no sufficient consideration is made for the economical property, to increase the size of a calandria tank and make the cost expensive. Then, in the present invention, the inner diameter of a pressure tube is set to greater than 50% of the lattice gap in a square lattice like arrangement, and the difference between the inner and the outer diameters of the calandria tube is set smaller than 20% of the lattice gap. Further, the inner diameter of the pressure tube is set to greater than 40% and the difference between the inner and the outer diameters of the calandria tube is set smaller than 30% of the lattice gap in a triangle lattice arrangement. Then, heavy water-to-fuel volume ratio can be determined appropriately and the value for the coolant void coefficient is made more negative side, to improve the self controllability inherent to the reactor. In particular, when 72 to 90 fuel rods are arranged per one pressure tube, the power density per one fuel rod is can be increased by about twice. Accordingly, the number of the pressure tubes can be reduced about to one-half, thereby enabling to remarkably decrease the diameter of the reactor core and to reduce the size of the calandria, which is economical. (N.H.)

  10. BWR type reactor

    International Nuclear Information System (INIS)

    In a coolant circulation in BWR type reactors, since the mixed stream of steam fluid undergoes a great resistance, the pressure loss due to the flow rate distribution when the coolants flow from the upper plenum into the stand pipe is increased upon passing stand pipe. Also in the spontaneous recycling reactor, pressure loss is still left upon passing the swirling blade of a gas-liquid separator. In view of the above, a plurality of vertical members each having a lower end opened to a gas-liquid two phase boundary and an upper end directly suspended from a steam dryer to the gas-liquid separator. The liquid droplets from the 2-phase boundary heated in the reactor core and formed into a mixed gas-liquid 2-phase stream is directed in the vertical direction accompanied with the steam. The liquid droplets spontaneously fallen by gravity from greater ones successively and the droplets in the steam abutted against the vertical member are fallen as a liquid membrane. Thus, the gas-liquid separation is conducted, the dry steam is directly flown into the steam dryer, thereby capable of providing a gas-liquid separator having gas-liquid separation performance with lower loss than usual. (N.H.)

  11. LMFBR type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kawakami, Hiroto

    1995-02-07

    A reactor container of the present invention has a structure that the reactor container is entirely at the same temperature as that at the inlet of the reactor and, a hot pool is incorporated therein, and the reactor container has is entirely at the same temperature and has substantially uniform temperature follow-up property transiently. Namely, if the temperature at the inlet of the reactor core changes, the temperature of the entire reactor container changes following this change, but no great temperature gradient is caused in the axial direction and no great heat stresses due to axial temperature distribution is caused. Occurrence of thermal stresses caused by the axial temperature distribution can be suppressed to improve the reliability of the reactor container. In addition, since the laying of the reactor inlet pipelines over the inside of the reactor is eliminated, the reactor container is made compact and the heat shielding structures above the reactor and a protection structure of container walls are simplified. Further, secondary coolants are filled to the outside of the reactor container to simplify the shieldings. The combined effects described above can improve economical property and reliability. (N.H.).

  12. LMFBR type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Iwashige, Kengo

    1996-06-21

    In an LMFBR type reactor, partitions are disposed to a coolant channel at positions lower than the free liquid level, and the width of the partitions is adapted to have a predetermined condition. Namely, when low temperature fluid overflowing the wall of the coolant channel, flows down and collided against the free liquid surface in the coolant channel, since the dropping speed thereof is reduced abruptly, large pressure waves are caused by kinetic force of the low temperature fluid. However, if appropriate numbers of partitions having an appropriate shape are formed, the dropping speed of the low temperature fluid is moderated to reduce the pressure waves. In addition, since the pressure waves are dispersed to the circumferential and lateral directions of the coolant flow channel respectively, the propagation of the pressure waves can be prevented effectively. Further, when the flow of the low temperature fluid is changed to the circumferential direction, for example, by earthquakes, since the partitions act as members resisting against the circumferential change of the low temperature fluid, the change of the direction can be suppressed. (N.H.)

  13. Natural convection type reactor

    International Nuclear Information System (INIS)

    In a natural convection type nuclear reactor, recycling flow rate of coolants is increased and the amount of entrained bubbles are increased as the driving force is increased, so that bubbles are not separated completely even if a stagnation region is disposed. Then, a space opened only at the upper portion is disposed at the outer circumference of the upper end of a riser for storing overflown coolants temporarily. The flow of coolants incorporating steam bubbles uprising in the riser turns into the horizontal direction at the upper end of the riser wall and flows into the coolant reservoir. In the coolant reservoir, since the momentum of the coolants is lost and the flow is stagnated, the bubbles are easily released to the upper space. Coolants, after releasing the bubbles, further overflow and descend in the downcomer. Then, the bubbles can be separated undergoing no influence of the driving force caused as the sum of the uprising force in the riser and the water head pressure in the downcomer, to prevent increase of carry under due to increase of the driving force. (N.H.)

  14. PWR type reactor

    International Nuclear Information System (INIS)

    From a PWR with a primary circuit, consisting of a reactor pressure vessel, a steam generator and a reactor coolant pump, hot coolant is removed by means of an auxiliary system containing h.p. pumps for feeding water into the primary circuit and being connected with a pipe, originating at the upper part, which has got at least one isolating value. This is done by opening an outlet in a part of the auxiliary system that has got a lower pressure than the reactor vessel. Preferably a water jet pump is used for mixing with the water of the auxiliary system. (orig.)

  15. New reactor type proposed

    CERN Multimedia

    2003-01-01

    "Russian scientists at the Research Institute of Nuclear Power Engineering in Moscow are hoping to develop a new reactor that will use lead and bismuth as fuel instead of uranium and plutonium" (1/2 page).

  16. The evolutionary journey of Argonaute proteins

    NARCIS (Netherlands)

    Swarts, D.C.; Makarova, K.; Wang, Y.; Nakanishi, K.; Ketting, R.F.; Koonin, E.V.; Patel, D.J.; Oost, van der J.

    2014-01-01

    Argonaute proteins are conserved throughout all domains of life. Recently characterized prokaryotic Argonaute proteins (pAgos) participate in host defense by DNA interference, whereas eukaryotic Argonaute proteins (eAgos) control a wide range of processes by RNA interference. Here we review molecula

  17. On reactor type comparisons for the next generation of reactors

    International Nuclear Information System (INIS)

    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

  18. Research Reactors Types and Utilization

    International Nuclear Information System (INIS)

    A nuclear reactor, in gross terms, is a device in which nuclear chain reactions are initiated, controlled, and sustained at a steady rate. The nuclei of fuel heavy atoms (mostly 235U or 239Pu), when struck by a slow neutron, may split into two or more smaller nuclei as fission products,releasing energy and neutrons in a process called nuclear fission. These newly-born fast neutrons then undergo several successive collisions with relatively low atomic mass material, the moderator, to become thermalized or slow. Normal water, heavy water, graphite and beryllium are typical moderators. These neutrons then trigger further fissions, and so on. When this nuclear chain reaction is controlled, the energy released can be used to heat water, produce steam and drive a turbine that generates electricity. The fission process, and hence the energy release, are controlled by the insertion (or extraction) of control rods through the reactor. These rods are strongly neutron absorbents, and thus only enough neutrons to sustain the chain reaction are left in the core. The energy released, mostly in the form of heat, should be continuously removed, to protect the core from damage. The most significant use of nuclear reactors is as an energy source for the generation of electrical power and for power in some military ships. This is usually accomplished by methods that involve using heat from the nuclear reaction to power steam turbines. Research reactors are used for radioisotope production and for beam experiments with free neutrons. Historically, the first use of nuclear reactors was the production of weapons grade plutonium for nuclear weapons. Currently all commercial nuclear reactors are based on nuclear fission. Fusion power is an experimental technology based on nuclear fusion instead of fission.

  19. Advanced nuclear reactor types and technologies

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-07-01

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

  20. Advanced nuclear reactor types and technologies

    International Nuclear Information System (INIS)

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

  1. Decommissioning of TRIGA Mark II type reactor

    International Nuclear Information System (INIS)

    The first research reactor in Korea, KRR 1, is a TRIGA Mark II type with open pool and fixed core. Its power was 100 kWth at its construction and it was upgraded to 250 kWth. Its construction was started in 1957. The first criticality was reached in 1962 and it had been operated for 36,000 hours. The second reactor, KRR 2, is a TRIGA Mark III type with open pool and movable core. These reactors were shut down in 1995, and the decision was made to decommission both reactors. The aim of the decommissioning activities is to decommission the KRR 2 reactor and decontaminate the residual building structures and site, and to release them as unrestricted areas. The KRR 1 reactor was decided to be preserve as a historical monument. A project was launched for the decommissioning of these reactors in 1997, and approved by the regulatory body in 2000. A total budget for the project was 20.0 million US dollars. It was anticipated that this project would be completed and the site turned over to KEPCO by 2010. However, it was discovered that the pool water of the KRR 1 reactor was leaked into the environment in 2009. As a result, preservation of the KRR 1 reactor as a monument had to be reviewed, and it was decided to fully decommission the KRR 1 reactor. Dismantling of the KRR 1 reactor takes place from 2011 to 2014 with a budget of 3.25 million US dollars. The scope of the work includes licensing of the decommissioning plan change, removal of pool internals including the reactor core, removal of the thermal and thermalizing columns, removal of beam port tubes and the aluminum liner in the reactor tank, removal of the radioactive concrete (the entire concrete structure will not be demolished), sorting the radioactive waste (concrete and soil) and conditioning the radioactive waste for final disposal, and final statuses of the survey and free release of the site and building, and turning over the site to KEPCO. In this paper, the current status of the TRIGA Mark-II type reactor

  2. Decommissioning of TRIGA Mark II type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Dooseong; Jeong, Gyeonghwan; Moon, Jeikwon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-10-15

    The first research reactor in Korea, KRR 1, is a TRIGA Mark II type with open pool and fixed core. Its power was 100 kWth at its construction and it was upgraded to 250 kWth. Its construction was started in 1957. The first criticality was reached in 1962 and it had been operated for 36,000 hours. The second reactor, KRR 2, is a TRIGA Mark III type with open pool and movable core. These reactors were shut down in 1995, and the decision was made to decommission both reactors. The aim of the decommissioning activities is to decommission the KRR 2 reactor and decontaminate the residual building structures and site, and to release them as unrestricted areas. The KRR 1 reactor was decided to be preserve as a historical monument. A project was launched for the decommissioning of these reactors in 1997, and approved by the regulatory body in 2000. A total budget for the project was 20.0 million US dollars. It was anticipated that this project would be completed and the site turned over to KEPCO by 2010. However, it was discovered that the pool water of the KRR 1 reactor was leaked into the environment in 2009. As a result, preservation of the KRR 1 reactor as a monument had to be reviewed, and it was decided to fully decommission the KRR 1 reactor. Dismantling of the KRR 1 reactor takes place from 2011 to 2014 with a budget of 3.25 million US dollars. The scope of the work includes licensing of the decommissioning plan change, removal of pool internals including the reactor core, removal of the thermal and thermalizing columns, removal of beam port tubes and the aluminum liner in the reactor tank, removal of the radioactive concrete (the entire concrete structure will not be demolished), sorting the radioactive waste (concrete and soil) and conditioning the radioactive waste for final disposal, and final statuses of the survey and free release of the site and building, and turning over the site to KEPCO. In this paper, the current status of the TRIGA Mark-II type reactor

  3. Modeling and analysis of Schistosoma Argonaute protein molecular spatial conformation

    Directory of Open Access Journals (Sweden)

    Jianhua Zhang

    2011-08-01

    Conclusions: The information relationship between the structure and function of the Argonaute protein can be initially established with bioinformatics tools and the internet server, and this provides the theoretical basis for further clarifying the function of Schistosoma Argonaute protein.

  4. Moving hot cell for LMFBR type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kanbe, Mitsuru

    1994-09-16

    A moving hot cell for an LMFBR type reactor is made movable on a reactor operation floor between a position just above the reactor container and a position retreated therefrom. Further, it comprises an overhung portion which can incorporate a spent fuel just thereunder, and a crane for moving a fuel assembly between a spent fuel cask and a reactor container. Further, an opening/closing means having a shielding structure is disposed to the bottom portion and the overhung portion thereof, to provide a sealing structure, in which only the receiving port for the spent fuel cask faces to the inner side, and the cask itself is disposed at the outside. Upon exchange of fuels, the movable hot cell is placed just above the reactor to take out the spent fuels, so that a region contaminated with primary sodium is limited within the hot cell. On the other hand, upon maintenance and repair for equipments, the hot cell is moved, thereby enabling to provide a not contaminated reactor operation floor. (N.H.).

  5. Method of operating BWR type reactors

    International Nuclear Information System (INIS)

    Purpose: To enable reactor control depending on any demanded loads by performing control by the insertion of control rods in addition to the control by the regulation of the flow rate of the reactor core water at high power operation of a BWR type reactor. Method: The power is reduced at high power operation by decreasing the flow rate of reactor core water from the starting time for the power reduction and the flow rate is maintained after the time at which it reaches the minimum allowable flow rate. Then, the control rod is started to insert from the above time point to reduce the power to an aimed level. Thus, the insufficiency in the reactivity due to the increase in the xenon concentration can be compensated by the withdrawal of the control rods and the excess reactivity due to the decrease in the xenon concentration can be compensated by the insertion of the control rods, whereby the reactor power can be controlled depending on any demanded loads without deviating from the upper or lower limit for the flow rate of the reactor core water. (Moriyama, K.)

  6. Effects of Argonaute on Gene Expression in Thermus thermophilus.

    Directory of Open Access Journals (Sweden)

    Daan C Swarts

    Full Text Available Eukaryotic Argonaute proteins mediate RNA-guided RNA interference, allowing both regulation of host gene expression and defense against invading mobile genetic elements. Recently, it has become evident that prokaryotic Argonaute homologs mediate DNA-guided DNA interference, and play a role in host defense. Argonaute of the bacterium Thermus thermophilus (TtAgo targets invading plasmid DNA during and after transformation. Using small interfering DNA guides, TtAgo can cleave single and double stranded DNAs. Although TtAgo additionally has been demonstrated to cleave RNA targets complementary to its DNA guide in vitro, RNA targeting by TtAgo has not been demonstrated in vivo.To investigate if TtAgo also has the potential to control RNA levels, we analyzed RNA-seq data derived from cultures of four T. thermophilus strain HB27 variants: wild type, TtAgo knockout (Δago, and either strain transformed with a plasmid. Additionally we determined the effect of TtAgo on expression of plasmid-encoded RNA and plasmid DNA levels.In the absence of exogenous DNA (plasmid, TtAgo presence or absence had no effect on gene expression levels. When plasmid DNA is present, TtAgo reduces plasmid DNA levels 4-fold, and a corresponding reduction of plasmid gene transcript levels was observed. We therefore conclude that TtAgo interferes with plasmid DNA, but not with plasmid-encoded RNA. Interestingly, TtAgo presence stimulates expression of specific endogenous genes, but only when exogenous plasmid DNA was present. Specifically, the presence of TtAgo directly or indirectly stimulates expression of CRISPR loci and associated genes, some of which are involved in CRISPR adaptation. This suggests that TtAgo-mediated interference with plasmid DNA stimulates CRISPR adaptation.

  7. Operation method for BWR type reactor

    International Nuclear Information System (INIS)

    In a BWR type reactor, the number of fuels at low enrichment, among initially loaded fuels, is increased greater than that of fuels to be exchanged, and the number of fuels at low enrichment remained in a reactor core after fuel exchange is decreased to smaller than that of entire control rods. Further, the fuels at low enrichment are disposed to the inner side except for the outermost circumference in the reactor core after fuel exchange. Since fuels of high reactivity are disposed at the outermost circumference in a second cycle, leakage of neutrons is increased and effective breeding factor is decreased. However, since the number of brought over fuels at low enrichment is decreased and the number of fuels at high enrichment is increased, effective average reactor core enrichment degree is increased, to compensate the lowering thereof due to the increase of neutron leakage. Since dispersion effect for the distribution of the enrichment degree can be utilized as much as possible by greatly reducing the number and the enrichment degree of fuels at low enrichment for initially loaded fuels, irrespective of the average enrichment degree and the fueling pattern in a first cycle, a burnup degree upon take-out of initially loaded fuels at ow enrichment degree can be increased to maximum. (N.H.)

  8. The Crystal Structure of Human Argonaute2

    Energy Technology Data Exchange (ETDEWEB)

    Schirle, Nicole T.; MacRae, Ian J. (Scripps)

    2012-07-18

    Argonaute proteins form the functional core of the RNA-induced silencing complexes that mediate RNA silencing in eukaryotes. The 2.3 angstrom resolution crystal structure of human Argonaute2 (Ago2) reveals a bilobed molecule with a central cleft for binding guide and target RNAs. Nucleotides 2 to 6 of a heterogeneous mixture of guide RNAs are positioned in an A-form conformation for base pairing with target messenger RNAs. Between nucleotides 6 and 7, there is a kink that may function in microRNA target recognition or release of sliced RNA products. Tandem tryptophan-binding pockets in the PIWI domain define a likely interaction surface for recruitment of glycine-tryptophan-182 (GW182) or other tryptophan-rich cofactors. These results will enable structure-based approaches for harnessing the untapped therapeutic potential of RNA silencing in humans.

  9. Domain motions of Argonaute, the catalytic engine of RNA interference

    OpenAIRE

    Wall Michael E; Ming Dengming; Sanbonmatsu Kevin Y

    2007-01-01

    Abstract Background The Argonaute protein is the core component of the RNA-induced silencing complex, playing the central role of cleaving the mRNA target. Visual inspection of static crystal structures already has enabled researchers to suggest conformational changes of Argonaute that might occur during RNA interference. We have taken the next step by performing an all-atom normal mode analysis of the Pyrococcus furiosus and Aquifex aeolicus Argonaute crystal structures, allowing us to quant...

  10. Modeling and analysis of Schistosoma Argonaute protein molecular spatial conformation

    Institute of Scientific and Technical Information of China (English)

    Jianhua Zhang; Zhigang Shang; Xiaohui Zhang; Yuntao Zhang

    2011-01-01

    Objective: To analyze the amino acid sequence composition, secondary structure, the spatial conformation of its domain and other characteristics of Argonaute protein. Methods:Bioinformatics tools and the internet server were used. Firstly, the amino acid sequence composition features of the Argonaute protein were analyzed, and the phylogenetic tree was constructed. Secondly, Argonaute protein’s distribution of secondary structure and its physicochemical properties were predicted. Lastly, the protein functional expression form of the domain group was established through the Phyre-based analysis on the spatial conformation of Argonaute protein domains. Results: 593 amino acids were encoded by Argonaute protein, the phylogenetic tree was constructed, and Argonaute protein’s distribution of secondary structure and its physicochemical properties were obtained through analysis. In addition, the functional expression form which comprised the N-terminal PAZ domain and C-terminal Piwi domain for the Argonaute protein was obtained with Phyre. Conclusions: The information relationship between the structure and function of the Argonaute protein can be initially established with bioinformatics tools and the internet server, and this provides the theoretical basis for further clarifying the function of Schistosoma Argonaute protein.

  11. Fast reactor parameter optimization taking into account changes in fuel charge type during reactor operation time

    International Nuclear Information System (INIS)

    The formulation and solution of optimization problem for parameters determining the layout of the central part of sodium cooled power reactor taking into account possible changes in fuel charge type during reactor operation time are performed. The losses under change of fuel composition type for two reactor modifications providing for minimum doubling time for oxide and carbide fuels respectively, are estimated

  12. Study on the Adaptability of Etheriifcation Feedstock to Reactor Type

    Institute of Scientific and Technical Information of China (English)

    Mao Junyi; Yuan Qing; Wang Lei; Huang Tao

    2016-01-01

    A reactive C5 oleifns and methanol etheriifcation kinetic model based on E-R mechanism was established and three different types of reactors including the adiabatic ifxed-bed liquid reactor, the external loop reactor and the mixed-phase reactor were constructed by Aspen Plus. The adaptability of reactive C5 oleifns to these reactors was studied and simulated using various gasoline fractions with different oleifns content. After the theoretical model was validated by the experimental data of the etheriifcation of three C5 light cut fractions from different gasoline sources in different reactors, the simulated isoamylene conversion with reactive C5 olefin contents increasing from 10% to 60% was studied in the three different types of reactors for etheriifcation with methanol, respectively. Test results show that there is an obvious adaptability of the feedstock composition to the reactor type to achieve a high conversion.

  13. Spent fuel situation at the ASTRA Seibersdorf and the TRIGA Vienna research reactors

    International Nuclear Information System (INIS)

    In the past decades Austria operated three research reactors, the 10 MW ASTRA reactor at Seibersdorf, the 250 kW TRIGA reactor at the Atomic Institut Vienna and the 1 kW Argonaut reactor at the Technical University in Graz. Since the shut down on July 31st, 1999 and decommissioning of the ASTRA reactor and the shut down of the ARGONAUT reactor Graz on July 31, 2004 only the TRIGA reactor remains operational. The MTR fuel elements of the ASTRA reactor have been shipped in spring 2001 to Savannah River and the fuel plates from the ARGONAUT reactor Graz in December 2005 under the DOE fuel return programme. (author)

  14. Beyond the Conventional Argonauts Explanations

    DEFF Research Database (Denmark)

    Rezaei, Shahamak

    professional and cultural experiences and their network in their home and host country to start up new firms in their home country. Following Schumpeter (1942) these firms are pr definition the outcome of innovative activities as the bifocality (Honig et al. 2010) of the transnational entrepreneurs allows......The last 10-15 years have witnessed a surge of interest in high-tech transnational entrepreneurship. The research has especially been spurred by the centrality of Indians and Chinese in Silicon Valley as the overlooked innovators. Special attention has been paid to how they used their bifocal...... for novel types of resource combinations and thus new products (i.e. new to the firm and market) and possible new organizational forms (e.g. double location sourcing). The centrality of transnational entrepreneurs within the ICT industry has been thoroughly documented in the US, Taiwan, China and India...

  15. Dynamic behaviour of a CAREM type reactor

    International Nuclear Information System (INIS)

    As complement to CAREM reactor design studies, behaviour analysis were made in a non-stationary regime, with the aim of developing plant systems and determining process variables variation ranges, characteristic of normal operating conditions, specifying alarm values for different variables, as well as for operating policies. Transient accidental scenes analysis were made, concluding that reactor characteristics provide security, maintaining the core integrity. (Author)

  16. Regulated Proteolysis of Arabidopsis Argonaute1

    DEFF Research Database (Denmark)

    Kausika, Swathi Pranavi

    domain as a bait in yeast two-hybrid screen to identify interactors. We found several candidates, of which many bound preferentially to AGO1 and were known to be integral or peripheral membrane proteins. Further investigations into interactors led us to discover that some of them play a role......Argonaute (AGO) proteins are key effectors of RNA Induced Silencing Complex (RISC) that mediates RNA interference or gene silencing. AGO binds small RNA and uses base pairing to small RNA to bring about repression of specific, complementary target RNA by cleavage or translational repression....... These are large multi domain proteins found in both eukaryotes and prokaryotes. They consist of four domains namely N, PAZ, MID and PIWI domain. The PAZ and MID domains bind small RNA 3´ and 5´ends while the PIWI domain harbors the endonucleolytic activity responsible for target RNA cleavage. This study focuses...

  17. Time and myth: the Argonauts in Ljubljana

    Directory of Open Access Journals (Sweden)

    Ana María Sánchez Tarrío

    2012-12-01

    Full Text Available This papers gives a breif overview of research in its initial phases, which examines, from the perspective of the construction of national identity, the traditional theme of the enduring and timeless nature of Classical myths and in particular the ancient myth, which is the subject of Borges’s sharp irony in his short-story “The immortal”. The reception of the Argonauts myth in Slovenian culture, initiated by the work of Janez Vajkard or Johann Weichard Freiherr von Valvasor (1641–1693, offers a significant case-study, which also highlights the relevance of Humanist and Baroque culture in the critical history of European nationalism. Writing before the emergence in Europe of the Hegelian conception of “Volkgeist”, the polymath writer applied a humanistic approach to patriotic themes, revealing Slovene culture, hybrid from its origins, to the wider world. A characteristic feature of his approach was the fusion of earlier textual tradition with contemporary oral material. Both in Slovenia and in the rest of Europe the comparison of the nineteenth century treatment of material dealing with national identity with its earlier treatment and transmission by humanist writers highlights the importance of the 16th and 17th centuries in the configuration of the different national faces of Europe as well as the significant role of common Greek and Latin roots. As a result, the myth of the Argonauts in Ljubljana, against the backdrop of idealistic or essentialist nationalist faiths, has the not inconsiderable virtue of underscoring the contaminatio that is characteristic of the construction of national identity.

  18. Self-operation type power control device for nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kanbe, Mitsuru.

    1993-07-23

    The device of the present invention operates by sensing the temperature change of a reactor core in all of LMFBR type reactors irrespective of the scale of the reactor core power. That is, a region where liquid poison is filled is disposed at the upper portion and a region where sealed gases are filled is disposed at the lower portion of a pipe having both ends thereof being closed. When the pipe is inserted into the reactor core, the inner diameter of the pipe is determined smaller than a predetermined value so that the boundary between the liquid poison and the sealed gases in the pipe is maintained relative to an assumed maximum acceleration. The sealed gas region is disposed at the reactor core region. If the liquid poison is expanded by the elevation of the reactor core exit temperature, it is moved to the lower gas region, to control the reactor power. Since high reliability can be maintained over a long period of time by this method, it is suitable to FBR reactors disposed in such environments that maintenance can not easily be conducted, such as desserts, isolated islands and undeveloped countries. Further, it is also suitable to ultra small sized nuclear reactors disposed at environments that the direction and the magnitude of gravity are different from those on the ground. (I.S.).

  19. Pin-Type Gas Cooled Reactor for Nuclear Electric Propulsion

    Science.gov (United States)

    Wright, Steven A.; Lipinski, Ronald J.

    2003-01-01

    This paper describes a point design for a pin-type Gas-Cooled Reactor concept that uses a fuel pin design similar to the SP100 fuel pin. The Gas-Cooled Reactor is designed to operate at 100 kWe for 7 years plus have a reduced power mode of 20% power for a duration of 5 years. The power system uses a gas-cooled, UN-fueled, pin-type reactor to heat He/Xe gas that flows directly into a recuperated Brayton system to produce electricity. Heat is rejected to space via a thermal radiator that unfolds in space. The reactor contains approximately 154 kg of 93.15 % enriched UN in 313 fuel pins. The fuel is clad with rhenium-lined Nb-1Zr. The pressures vessel and ducting are cooled by the 900 K He/Xe gas inlet flow or by thermal radiation. This permits all pressure boundaries to be made of superalloy metals rather than refractory metals, which greatly reduces the cost and development schedule required by the project. The reactor contains sufficient rhenium (a neutron poison) to make the reactor subcritical under water immersion accidents without the use of internal shutdown rods. The mass of the reactor and reflectors is about 750 kg.

  20. Domain motions of Argonaute, the catalytic engine of RNA interference

    Directory of Open Access Journals (Sweden)

    Wall Michael E

    2007-11-01

    Full Text Available Abstract Background The Argonaute protein is the core component of the RNA-induced silencing complex, playing the central role of cleaving the mRNA target. Visual inspection of static crystal structures already has enabled researchers to suggest conformational changes of Argonaute that might occur during RNA interference. We have taken the next step by performing an all-atom normal mode analysis of the Pyrococcus furiosus and Aquifex aeolicus Argonaute crystal structures, allowing us to quantitatively assess the feasibility of these conformational changes. To perform the analysis, we begin with the energy-minimized X-ray structures. Normal modes are then calculated using an all-atom molecular mechanics force field. Results The analysis reveals low-frequency vibrations that facilitate the accommodation of RNA duplexes – an essential step in target recognition. The Pyrococcus furiosus and Aquifex aeolicus Argonaute proteins both exhibit low-frequency torsion and hinge motions; however, differences in the overall architecture of the proteins cause the detailed dynamics to be significantly different. Conclusion Overall, low-frequency vibrations of Argonaute are consistent with mechanisms within the current reaction cycle model for RNA interference.

  1. Control rod for PWR type reactor

    International Nuclear Information System (INIS)

    Since a silver-indium-cadmium alloy has been used as the absorber for control rods, swelling due to neutron absorption has been caused. On the other hand, a stainless steel cladding tube for the absorber gradually reduces its outer diameter by the pressure of reactor coolants and neutron irradiation and causes contact during working life to often bring about cracking in the cladding tube. Then, the control rod is divided into two independent portions and joined by an intermediate end plug into a single rod, in which the upper portion is made free from pressure and the lower portion is pressurized. Further, a large gap is formed between the lower absorber and the lower cladding tube. Further, chromium or chromium carbide is coated to the outer surface of the upper cladding tube for improving the abrasion resistance. Thus, the cladding tube is made abrasion resistant and it is possible to prevent cracking in the cladding tube due to interaction between the tube and the absorber, inner presurization at the lower portion, reduced diameter for the absorber and the gap of the tube. (N.H.)

  2. Fuel assembly for BWR type reactor

    International Nuclear Information System (INIS)

    In the fuel assembly of the present invention, a means for mounting and securing short fuel rods is improved. Not only long fuel rods but also short fuel rods are disposed in channel of the fuel assembly to improve reactor safety. The short fuel rods are supported by a screw means only at the lower end plug. The present invention prevents the support for the short fuel rod from being unreliable due to the slack of the screw by the pressure of inflowing coolants. That is, coolant abutting portions such as protrusions or concave grooves are disposed at a portion in the channel box where coolants flowing from the lower tie plate, as an uprising stream, cause collision. With such a constitution, a component caused by the pressure of the flowing coolants is formed. The component acts as a rotational moment in the direction of screwing the male threads of the short fuel rod into the end plug screw hole. Accordingly, the screw is not slackened, and the short fuel rods are mounted and secured certainly. (I.S.)

  3. Device for reducing radioactive corrosion product in FBR type reactor

    International Nuclear Information System (INIS)

    The present invention concerns an FBR type reactor using liquid metal as coolants, connecting the reactor core with a heat exchanger by way of cooling system pipeways and recycling the coolant by the driving force of a pump. A bypass circuit is disposed to a portion of a cooling system, and a vessel inserted with fillers is disposed to a portion of the bypass circuit. The coolants are prepared with the same material as that for the reactor core constituent material. The filler suffered from corrosion with sodium coolants and to increase the concentration of the corrosion products in sodium. This suppresses the corrosion of nuclear fuel cans in the reactor core. Accordingly, leaching of radioactive corrosion products such as Mn or Co caused by the reduction in the wall thickness of the fuel can can be suppressed. (I.J.)

  4. Channel-type nuclear reactor with a boiling coolant

    International Nuclear Information System (INIS)

    The invention is aimed at increasing the channel-type reactor safety, in particular, RBMK-type reactors, during accidents resulting in the coolant circulation discontinuation. The reactor core is assembled of vertial technological channels connected in parallel between distributing group collectors and drum-separator. Each technological channel contains a high pressure tube, a fuel assembly with fuel elements and a storage vessel located above the fuel assembly which is filled with water at saturation temperature in the normal operation regime. After dehydration of channels in the course of accident the boiling water from storage vessel is ejected into them. So the device described allows one to reduce the fuel element can temperature in the course of accidents connected with the coolant circulation discontinuation and so to increase the plant safety level

  5. Description of the advanced gas cooled type of reactor (AGR)

    International Nuclear Information System (INIS)

    The present report comprises a technical description of the Advanced Gas cooled Reactor (AGR), a reactor type which has only been built in Great Britain. 14 AGR reactors have been built, located at 6 different sites and each station is supplied with twin-reactors. The Torness AGR plant on the Lothian coastline of Scotland, 60 km east of Edinburgh, has been chosen as the reference plant and is described in some detail. Data on the other 6 stations, Dungeness B, Hinkely Point B, Hunterston G, Hartlepool, Heysham I and Heysham II, are given only in tables with a summary of design data. Where specific data for Torness AGR has not been available, corresponding data from other AGR plans has been used, primarily from Heysham II, which belongs to the same generation of AGR reactors. The information presented is based on the open literature. The report is written as a part of the NKS/RAK-2 subproject 3: 'Reactors in Nordic Surroundings', which comprises a description of nuclear power plants neighbouring the Nordic countries. (au) 11 refs

  6. Evaluation of plate type fuel options for small power reactors

    International Nuclear Information System (INIS)

    Plate type fuels are generally used in research reactor. The utilization of this kind of configuration improves significantly the overall performance fuel. The conception of new fuels for small power reactors based in plate-type configuration needs a complete review of the safety criteria originally used to conduce power and research reactor projects. In this work, a group of safety criteria is established for the utilization of plate-type fuels in small power reactors taking into consideration the characteristics of power and research reactors. The performance characteristics of fuel elements are strongly supported by its materials properties and the adopted configuration for its fissile particles. The present work makes an orientated bibliographic investigation searching the best material properties (structural materials and fuel compounds) related to the performance fuel. Looking for good parafermionic characteristics and manufacturing exequibility associated to existing facilities in national research centres, this work proposes several alternatives of plate type fuels, considering its utilization in small power reactors: dispersions of UO2 in stainless steel, of UO2 in zircaloy, and of U-Mo alloy in zircaloy, and monolithic plates of U-Mo cladded with zircaloy. Given the strong dependency of radiation damage with temperature increase, the safety criteria related to heat transfer were verified for all the alternatives, namely the DNBR; coolant temperature lower than saturation temperature; peak meat temperature to avoid swelling; peak fuel temperature to avoid meat-matrix reaction. It was found that all alternatives meet the safety criteria including the 0.5 mm monolithic U-Mo plate cladded with zircaloy. (author)

  7. Description of the magnox type of gas cooled reactor (MAGNOX)

    International Nuclear Information System (INIS)

    The present report comprises a technical description of the MAGNOX type of reactor as it has been build in Great Britain. The Magnox reactor is gas cooled (CO2) with graphite moderators. The fuels is natural uranium in metallic form, canned with a magnesium alloy called 'Magnox'. The Calder Hall Magnox plant on the Lothian coastline of Scotland, 60 km east of Edinburgh, has been chosen as the reference plant and is described in some detail. Data on the other stations are given in tables with a summary of design data. Special design features are also shortly described. Where specific data for Calder Hall Magnox has not been available, corresponding data from other Magnox plants has been used. The information presented is based on the open literature. The report is written as a part of the NKS/RAK-2 sub-project 3: 'Reactors in Nordic Surroundings', which comprises a description of nuclear power plants neighbouring the Nordic countries. (au)

  8. Liquid film emergency for FRJ-2 type research reactors

    International Nuclear Information System (INIS)

    A new, efficient emergency cooling procedure based on liquid film cooling was developed for FRJ-2 type research in reactors, which allows a higher power generation in the tubular fuel elements used and which represents an improvement of the engineered safeguards of the reactor. The problem of producing coherent liquid films on the outer surfaces of the four concentrically arranged thin fuel tubes without obstructive modifications of the fuel element design was solved by using radial water jets. These jets discharge into the drained fuel elements from the outside therby crossing the upper edges of the fuel tubes. In hydraulic experiments the influence of the geometry, of the jet velocity and of the water viscosity on the water supply to each fuel tube was measured and the conditions were evaluated where by each fuel tube in the reactor obtain sufficient cooling water taking account of variations in the various parameters. (orig./HP)

  9. A Study on Dismantling of Westinghouse Type Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Woo-Tae; Lee, Sang-Guk [KHNP-CRI, Daejeon (Korea, Republic of)

    2014-10-15

    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.

  10. Artificial intelligence applied to fuel management in BWR type reactors

    International Nuclear Information System (INIS)

    In this work two techniques of artificial intelligence, neural networks and genetic algorithms were applied to a practical problem of nuclear fuel management; the determination of the optimal fuel reload for a BWR type reactor. This is an important problem in the design of the operation cycle of the reactor. As a result of the application of these techniques, comparable or even better reloads proposals than those given by expert companies in the subject were obtained. Additionally, two other simpler problems in reactor physics were solved: the determination of the axial power profile and the prediction of the value of some variables of interest at the end of the operation cycle of the reactor. Neural networks and genetic algorithms have been applied to solve many problems of engineering because of their versatility but they have been rarely used in the area of fuel management. The results obtained in this thesis indicates the convenience of undertaking further work on this area and suggest the application of these techniques of artificial intelligence to the solution of other problems in nuclear reactor physics. (Author)

  11. Safety evaluation report related to the renewal of the operating license for the research reactor at the University of Florida. Docket No. 50-83

    International Nuclear Information System (INIS)

    This Safety Evaluation Report for the application filed by the University of Florida (UF) for a renewal of Operating License R-56 to continue to operate its Argonaut-type research reactor has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is owned and operated by the University of Florida and is located on the UF campus in Gainesville, Alachua County, Florida. The staff concludes that the reactor facility can continue to be operated by UF without endangering the health and safety of the public

  12. The JASON reactor: from core removal to fuel reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Beeley, P.; Williams, A.; Lockwood, R. [Defence College of Electromechanical Engineering, Nuclear Dept., HMS SULTAN (United Kingdom); Raymond, B.; Spyrou, N. [Surrey Univ., Dept. of Physical and Electronic Sciences (United Kingdom); Auziere, P. [AREVA NC, Treatment Business Unit, 78 - Velizy (France)

    2007-07-01

    The 10 kW JASON Argonaut reactor was operated at the Royal Naval College, Greenwich, London, between 1962 and 1996. After initial cooling in the core, the MTR type fuel (80% enriched U{sup 235}) was dry stored on site before transport in 1998 to BNFL, Sellafield for interim wet storage. Arrangements for reprocessing of the fuel at AREVA NC, La Hague are now in progress and this paper will describe various aspects of the storage, transfer, monitoring, and the treatment at La Hague plant. The radioactive waste resulting from the processing of these used fuels will be conditioned into a suitable package for return to UK.

  13. Conceptual design of a pool type molten salt breeder reactor

    International Nuclear Information System (INIS)

    The renewed interest in molten salt coolant technology is backed by the 50 years history of molten salt nuclear technology development, mainly in Oak Ridge National Laboratory (ORNL). In Indian context MSBR is found to be one of the options for sustainable nuclear energy generation, especially in the third stage of the nuclear programme. The system can be operated at high temperature which makes high efficiency power conversion and efficient hydrogen generation through thermo-chemical reactions possible. At present development is in progress in BARC on two molten salt reactor concepts, one is pool type and the other is loop type. Here the design of pool type concept with 850MWe power is described. The core is designed to operate in the fast spectrum region so the conversion of 233U breeding is possible from thorium. Preliminary thermal hydraulic analysis is carried out with LiF-ThF4-UF4 as the primary fuel and coolant. The blanket material is also a molten salt, LiF-ThF4. Reactor physics calculations are also carried out for the feasibility studies of the core design of the reactor. FLiNaK is used as the secondary coolant for the calculations. Both forced circulation and natural circulation options are evaluated. (author)

  14. Identification and characterization of Argonaute gene family and meiosis-enriched Argonaute during sporogenesis in maize

    Institute of Scientific and Technical Information of China (English)

    Zuxin Zhang

    2014-01-01

    Argonaute (AGO) proteins play a key role in regulation of gene expression through smal RNA‐directed RNA cleavage and translational repression, and are essential for multiple developmental processes. In the present study, 17 AGO genes of maize (Zea mays L., ZmAGOs) were identified using a Hidden Markov Model and validated by rapid amplifica-tion of cDNA ends assay. Subsequently, quantitative PCR revealed that expressions of these genes were higher in reproductive than in vegetative tissues. AGOs presented five temporal and spatial expression patterns, which were likely modulated by DNA methylation, 50‐untranslated exons and microRNA‐mediated feedback loops. Intriguingly, ZmAGO18b was highly expressed in tassels during meiosis. Furthermore, in situ hybridization and immunofluorescence showed that ZmA- GO18b was enriched in the tapetum and germ cel s in meiotic anthers. We hypothesized that ZmAGOs are highly expressed in reproductive tissues, and that ZmAGO18b is a tapetum and germ cel‐specific member of the AGO family in maize.

  15. An innovative reactor-type biosensor for BOD rapid measurement.

    Science.gov (United States)

    Wang, Jianlong; Zhang, Yixin; Wang, Yeyao; Xu, Runhua; Sun, Zhonghua; Jie, Zhou

    2010-03-15

    Biochemical oxygen demand (BOD) is one of the most important and widely used parameters for characterizing the organic pollution of water and wastewater. In this paper, a novel reactor-type biosensor for rapid measurement of BOD was developed, based on using immobilized microbial cell (IMC) beads as recognition bio-element in a completely mixed reactor which was used as determining chamber, replacing the traditionally used membrane as recognition bio-element. The IMC beads were freely suspended in the aqueous solution, so the mass transfer resistance for dissolved oxygen and organic compounds significantly reduced, and the quantity of the microbial cells used as recognition element can be easily adjusted, in comparison with the traditional membrane-type BOD biosensor, in which exists a unadjustable contradiction between the quantity of biomass and the thickness of the bio-membrane, thus limiting the stability and the detection limit. This novel kind of BOD biosensor significantly increased the sensitivity of the response, the detecting precision and prolonged the life time of the recognition element. The experimental data showed that the most appropriate temperature for biochemical reaction in the reactor was 30 degrees C, and the IMC beads could keep the bioactivity for about 70d at the detecting frequency of 8 times every day. The standard deviation of repeatability and the reproducibility of responses were within +/-6.4% and +/-5.0%, respectively, which are within acceptable bias limits, and meet the requirement of BOD rapid measurement.

  16. Three-dimensional reactor dynamics code for VVER type nuclear reactors. Doctoral thesis

    Energy Technology Data Exchange (ETDEWEB)

    Kyrki-Rajamaeki, R.

    1995-11-17

    A three-dimensional reactor dynamics computer code HEXTRAN has been developed, thoroughly validated, and extensively applied for transient and accident analyses of VVER type nuclear reactors. HEXTRAN models accurately the VVER core with hexagonal fuel assemblies. The code uses advanced mathematical models in spatial and time discretization of neutronics, heat transfer and two-phase flow hydraulics. The dynamic coupling with the thermal hydraulic system code SMABRE allows also the modelling of cooling circuits. Best-estimate or conservative analyses can be performed for different accidents, e.g., RIA, ATWS or local boron dilutions. The usefulness of the three-dimensionality is shown particularly when there are asymmetric or thermal hydraulic disurbances in the core or cooling circuits.

  17. The utility of different reactor types for the research

    International Nuclear Information System (INIS)

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

  18. ARGONAUTE10 and ARGONAUTE1 regulate the termination of floral stem cells through two microRNAs in Arabidopsis.

    Directory of Open Access Journals (Sweden)

    Lijuan Ji

    2011-03-01

    Full Text Available Stem cells are crucial in morphogenesis in plants and animals. Much is known about the mechanisms that maintain stem cell fates or trigger their terminal differentiation. However, little is known about how developmental time impacts stem cell fates. Using Arabidopsis floral stem cells as a model, we show that stem cells can undergo precise temporal regulation governed by mechanisms that are distinct from, but integrated with, those that specify cell fates. We show that two microRNAs, miR172 and miR165/166, through targeting APETALA2 and type III homeodomain-leucine zipper (HD-Zip genes, respectively, regulate the temporal program of floral stem cells. In particular, we reveal a role of the type III HD-Zip genes, previously known to specify lateral organ polarity, in stem cell termination. Both reduction in HD-Zip expression by over-expression of miR165/166 and mis-expression of HD-Zip genes by rendering them resistant to miR165/166 lead to prolonged floral stem cell activity, indicating that the expression of HD-Zip genes needs to be precisely controlled to achieve floral stem cell termination. We also show that both the ubiquitously expressed ARGONAUTE1 (AGO1 gene and its homolog AGO10, which exhibits highly restricted spatial expression patterns, are required to maintain the correct temporal program of floral stem cells. We provide evidence that AGO10, like AGO1, associates with miR172 and miR165/166 in vivo and exhibits "slicer" activity in vitro. Despite the common biological functions and similar biochemical activities, AGO1 and AGO10 exert different effects on miR165/166 in vivo. This work establishes a network of microRNAs and transcription factors governing the temporal program of floral stem cells and sheds light on the relationships among different AGO genes, which tend to exist in gene families in multicellular organisms.

  19. Scale Effects on Magnet Systems of Heliotron-Type Reactors

    Institute of Scientific and Technical Information of China (English)

    S. Imagawa; A. Sagara

    2005-01-01

    For power plants heliotron-type reactors have attractive advantages, such as no current-disruptions, no current-drive, and wide space between helical coils for the maintenance of in-vessel components. However, one disadvantage is that a major radius has to be large enough to obtain large Q-value or to produce sufficient space for blankets. Although the larger radius is considered to increase the construction cost, the influence has not been understood clearly,yet. Scale effects on superconducting magnet systems have been estimated under the conditions of a constant energy confinement time and similar geometrical parameters. Since the necessary magnetic field with a larger radius becomes lower, the increase rate of the weight of the coil support to the major radius is less than the square root. The necessary major radius will be determined mainly by the blanket space. The appropriate major radius will be around 13 m for a reactor similar to the Large Helical Device (LHD).

  20. Installation of a new type of nuclear reactor in Mexico: advantages and disadvantages; Instalacion de un nuevo tipo de reactor nuclear en Mexico: ventajas y desventajas

    Energy Technology Data Exchange (ETDEWEB)

    Jurado P, M.; Martin del Campo M, C. [FI-UNAM, 04510 Mexico D.F. (Mexico)]. e-mail: mjp_green@hotmail.com

    2005-07-01

    In this work the main advantages and disadvantages of the installation of a new type of nuclear reactor different to the BWR type reactor in Mexico are presented. A revision of the advanced reactors is made that are at the moment in operation and of the advanced reactors that are in construction or one has already planned its construction in the short term. Specifically the A BWR and EPR reactors are analyzed. (Author)

  1. LMFBR type reactor and power generation system using the same

    Energy Technology Data Exchange (ETDEWEB)

    Otsubo, Akira.

    1994-02-25

    A reactor core void reactivity of a reactor main body is set to negative or zero. A heat insulation structure is disposed on the inner wall surface of a reactor container. Oxide fuels or nitride fuels are used. A fuel pin cladding tube has a double walled structure having an outer side of stainless steel and an inner side of niobium alloy. Upon imaginary event, boiling is allowed. Even if boiling of coolants should occur by temperature elevation of fuels upon imaginary event, since reactor core fuels comprises oxides or nitrides, they have a heat resistance, further, and since the fuel pin cladding tube has super heat resistance, it has a high temperature strength, so that it is not ruptured and durable to the coolant boiling temperature. Since the reactor core void reactivity is negative or zero, the reactor core is in a subcritical state by the boiling, and the reactor core power is reduced to several % of the rated power. Accordingly, boiling and non-boiling are repeated substantially permanently in the reactor core, during which safety can be kept with no operator's handling. Further, heat generated in the reactor core is gradually removed by an air cooling system for the reactor container. (N.H.).

  2. A multitasking Argonaute: exploring the many facets of C. elegans CSR-1.

    Science.gov (United States)

    Wedeles, Christopher J; Wu, Monica Z; Claycomb, Julie M

    2013-12-01

    While initial studies of small RNA-mediated gene regulatory pathways focused on the cytoplasmic functions of such pathways, identifying roles for Argonaute/small RNA pathways in modulating chromatin and organizing the genome has become a topic of intense research in recent years. Nuclear regulatory mechanisms for Argonaute/small RNA pathways appear to be widespread, in organisms ranging from plants to fission yeast, Caenorhabditis elegans to humans. As the effectors of small RNA-mediated gene regulatory pathways, Argonaute proteins guide the chromatin-directed activities of these pathways. Of particular interest is the C. elegans Argonaute, chromosome segregation and RNAi deficient (CSR-1), which has been implicated in such diverse functions as organizing the holocentromeres of worm chromosomes, modulating germline chromatin, protecting the genome from foreign nucleic acid, regulating histone levels, executing RNAi, and inhibiting translation in conjunction with Pumilio proteins. CSR-1 interacts with small RNAs known as 22G-RNAs, which have complementarity to 25 % of the protein coding genes. This peculiar Argonaute is the only essential C. elegans Argonaute out of 24 family members in total. Here, we summarize the current understanding of CSR-1 functions in the worm, with emphasis on the chromatin-directed activities of this ever-intriguing Argonaute.

  3. A multitasking Argonaute: exploring the many facets of C. elegans CSR-1.

    Science.gov (United States)

    Wedeles, Christopher J; Wu, Monica Z; Claycomb, Julie M

    2013-12-01

    While initial studies of small RNA-mediated gene regulatory pathways focused on the cytoplasmic functions of such pathways, identifying roles for Argonaute/small RNA pathways in modulating chromatin and organizing the genome has become a topic of intense research in recent years. Nuclear regulatory mechanisms for Argonaute/small RNA pathways appear to be widespread, in organisms ranging from plants to fission yeast, Caenorhabditis elegans to humans. As the effectors of small RNA-mediated gene regulatory pathways, Argonaute proteins guide the chromatin-directed activities of these pathways. Of particular interest is the C. elegans Argonaute, chromosome segregation and RNAi deficient (CSR-1), which has been implicated in such diverse functions as organizing the holocentromeres of worm chromosomes, modulating germline chromatin, protecting the genome from foreign nucleic acid, regulating histone levels, executing RNAi, and inhibiting translation in conjunction with Pumilio proteins. CSR-1 interacts with small RNAs known as 22G-RNAs, which have complementarity to 25 % of the protein coding genes. This peculiar Argonaute is the only essential C. elegans Argonaute out of 24 family members in total. Here, we summarize the current understanding of CSR-1 functions in the worm, with emphasis on the chromatin-directed activities of this ever-intriguing Argonaute. PMID:24178449

  4. Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hogerton, John

    1964-01-01

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

  5. Method of controlling the heterogeneous reactor core in FBR type reactors

    International Nuclear Information System (INIS)

    Purpose: To maintain the power distribution of fuel assemblies constant all over the reactor operation period by operating the control rods depending on the power change in blanket fuels. Method: Blanket fuels (internal blanket) are loaded at a central region of a reactor core comprising plutonium enriched region. Further, control rods for the start-up and shutdown of a reactor and fuel compensation and back-up control rods are arranged within the reactor core. The reactor core is surrounded with an axial blanket and a neutron shielding body. 21 fuel compensating control rods are present in the reactor core and 18 rods out of them are arranged at the outer region of the inner blanket. At the initial stage of the reactor operation, the control rods are divided into three blocks and they are inserted into the reactor core by 0%, 21% and 20% respectively required for the compensation of the burning reactivity at the initial stage of the reactor operation and inserted by 2%, 18% and 15% respectively at the initial balanced stage of the reactor core. (Horiuchi, T.)

  6. Neutron dosimetry. Environmental monitoring in a BWR type reactor

    International Nuclear Information System (INIS)

    The measurements carried out on reactor dosimetry are applied mainly to the study on the effects of the radiation in 108 materials of the reactor; little is on the environmental dosimetry outside of the primary container of BWR reactors. In this work the application of a neutron spectrometer formed by plastic detectors of nuclear traces manufactured in the ININ, for the environmental monitoring in penetrations around the primary container of the unit I of the Laguna Verde central is presented. The neutron monitoring carries out with purposes of radiological protection, during the operational tests of the reactor. (Author)

  7. Problems of control of WWER-type pressurized water reactors (PWR's)

    International Nuclear Information System (INIS)

    The problems are dealt with of nuclear power reactor control. Special attention is paid to the reactor of the WWER type, which will play the most important part in the Czechoslovak power system in the near future. The subsystems are described which comprise the systems of reactor control and protection. The possibilities are outlined of using Czechoslovak instrumentation for the control and safety system of the WWER-type PWR. (author)

  8. Safety-evaluation report related to the renewal of the operating license for the research reactor at the Iowa State University (Docket No. 50-116)

    International Nuclear Information System (INIS)

    This Safety Evaluation Report for the application filed by the Iowa State University (ISU) for a renewal of the Class 104 Operating License R-59 to continue to operate its Argonaut-type research reactor has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is owned and operated by the Iowa State University, and is located on the ISU campus in Ames, Story County, Iowa. The staff concludes that the reactor facility can continue to be operated by ISU without endangering the health and safety of the public. The principal matters reviewed are: design, testing, and performance of the reactor components and systems; the expected consequences of credible accidents; the licensee's management organization; the method used for the control of radiological effluents; the licensee's technical specifications; financial data and information; the physical protection program; procedures for training reactor operators; and emergency plans. 11 references, 15 figures, 13 tables

  9. Reactor Materials Program: Mechanical properties of irradiated Types 304 and 304L stainless steel weldment components

    International Nuclear Information System (INIS)

    The vessels (reactor tanks) of the Savannah River Site nuclear production reactors constructed in the 1950's are comprised of Type 304 stainless steel with Type 308 stainless steel weld filler. Irradiation exposure to the reactor tank sidewalls through reactor operation has caused a change in the mechanical properties of these materials. A database of as-irradiated mechanical properties for site-specific materials and irradiation conditions has been produced for reactor tank structural analyses and to quantify the effects of radiation-induced materials degradation for evaluating reactor service life. The data has been collected from the SRL Reactor Materials Program (RMP) irradiations and testing of archival stainless steel weldment components and from previous SRL programs to measure properties of irradiated reactor Thermal Shield weldments and reactor tank (R-tank) sidewall material. Irradiation programs of the RMP are designed to quantify mechanical properties at tank operating temperatures following irradiation to present and future tank wall maximum exposure conditions. The exposure conditions are characterized in terms of fast neutron fluence (En > 0.1 MeV) and displacements per atom (dpa)3. Tensile properties, Charpy-V notch toughness, and elastic-plastic fracture toughness were measured for base, weld, and weld heat-affected zone (HAZ) weldment components from archival piping specimens following a Screening Irradiation in the University of Buffalo Reactor (UBR) and following a Full-Term Irradiation in the High Flux Isotope Reactor (HFIR)

  10. Reactor Materials Program: Mechanical properties of irradiated Types 304 and 304L stainless steel weldment components

    Energy Technology Data Exchange (ETDEWEB)

    Sindelar, R.L.; Caskey, G.R. Jr.

    1991-12-01

    The vessels (reactor tanks) of the Savannah River Site nuclear production reactors constructed in the 1950's are comprised of Type 304 stainless steel with Type 308 stainless steel weld filler. Irradiation exposure to the reactor tank sidewalls through reactor operation has caused a change in the mechanical properties of these materials. A database of as-irradiated mechanical properties for site-specific materials and irradiation conditions has been produced for reactor tank structural analyses and to quantify the effects of radiation-induced materials degradation for evaluating reactor service life. The data has been collected from the SRL Reactor Materials Program (RMP) irradiations and testing of archival stainless steel weldment components and from previous SRL programs to measure properties of irradiated reactor Thermal Shield weldments and reactor tank (R-tank) sidewall material. Irradiation programs of the RMP are designed to quantify mechanical properties at tank operating temperatures following irradiation to present and future tank wall maximum exposure conditions. The exposure conditions are characterized in terms of fast neutron fluence (E{sub n} > 0.1 MeV) and displacements per atom (dpa){sup 3}. Tensile properties, Charpy-V notch toughness, and elastic-plastic fracture toughness were measured for base, weld, and weld heat-affected zone (HAZ) weldment components from archival piping specimens following a Screening Irradiation in the University of Buffalo Reactor (UBR) and following a Full-Term Irradiation in the High Flux Isotope Reactor (HFIR).

  11. Reactor Materials Program: Mechanical properties of irradiated Types 304 and 304L stainless steel weldment components

    Energy Technology Data Exchange (ETDEWEB)

    Sindelar, R.L.; Caskey, G.R. Jr.

    1991-12-01

    The vessels (reactor tanks) of the Savannah River Site nuclear production reactors constructed in the 1950`s are comprised of Type 304 stainless steel with Type 308 stainless steel weld filler. Irradiation exposure to the reactor tank sidewalls through reactor operation has caused a change in the mechanical properties of these materials. A database of as-irradiated mechanical properties for site-specific materials and irradiation conditions has been produced for reactor tank structural analyses and to quantify the effects of radiation-induced materials degradation for evaluating reactor service life. The data has been collected from the SRL Reactor Materials Program (RMP) irradiations and testing of archival stainless steel weldment components and from previous SRL programs to measure properties of irradiated reactor Thermal Shield weldments and reactor tank (R-tank) sidewall material. Irradiation programs of the RMP are designed to quantify mechanical properties at tank operating temperatures following irradiation to present and future tank wall maximum exposure conditions. The exposure conditions are characterized in terms of fast neutron fluence (E{sub n} > 0.1 MeV) and displacements per atom (dpa){sup 3}. Tensile properties, Charpy-V notch toughness, and elastic-plastic fracture toughness were measured for base, weld, and weld heat-affected zone (HAZ) weldment components from archival piping specimens following a Screening Irradiation in the University of Buffalo Reactor (UBR) and following a Full-Term Irradiation in the High Flux Isotope Reactor (HFIR).

  12. Design guide for Category III reactors: pool type reactors. [US DOE

    Energy Technology Data Exchange (ETDEWEB)

    Brynda, W J; Lobner, P R; Powell, R W; Straker, E A

    1978-11-01

    The Department of Energy (DOE) in the ERDA Manual requires that all DOE-owned reactors be sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that gives adequate consideration to health and safety factors. Specific guidance pertinent to the safety of DOE-owned reactors is found in Chapter 0540 of the ERDA Manual. The purpose of this Design Guide is to provide additional guidance to aid the DOE facility contractor in meeting the requirement that the siting, design, construction, modification, operation, maintenance, and decommissioning of DOE-owned reactors be in accordance with generally uniform standards, guides, and codes which are comparable to those applied to similar reactors licensed by the Nuclear Regulatory Commission (NRC). This Design Guide deals principally with the design and functional requirement of Category III reactor structures, components, and systems.

  13. A New Fuel Design for Two Different HW Type Reactors

    Directory of Open Access Journals (Sweden)

    Daniel O. Brasnarof

    2011-01-01

    Full Text Available A new fuel element (called CARA designed for two different heavy water reactors (HWRs is presented. CARA could match fuel requirements of both (one CANDU and one unique Siemens's design Argentine HW reactors. It keeps the heavier fuel mass density and hydraulic flow restriction in both reactors together with improving both thermomechanic and thermalhydraulic, safety margins of present fuels. In addition, the CARA design could be considered as another design line for the next generation of CANDU fuels intended for higher burnup.

  14. Characterization of Argonaute-related small RNA pathways in Caenorhabditis elegans

    OpenAIRE

    Batista, Pedro Jorge de Oliveira Rodrigues

    2010-01-01

    Tese de doutoramento, Biologia (Genética), Universidade de Lisboa, Faculdade de Ciências, 2011 In Small-RNA-mediated pathways, small RNAs engage a protein of the Argonaute family and utilize base-pairing interactions to identify and regulate complementary genetic information. My research has focused on understanding how diverse classes of small RNAs in the model organism Caenorhabditis elegans interact with specific members of the Argonaute protein family to carry out unique bi...

  15. Seismic stability of VGM type high temperature gas cooled reactors

    International Nuclear Information System (INIS)

    The main principles of the design provision of high temperature gas cooled VGM reactors seismic stability and the results of calculations, performed by linear-spectral method are presented. (author). 1 ref., 10 figs

  16. Solid0Core Heat-Pipe Nuclear Batterly Type Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ehud Greenspan

    2008-09-30

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

  17. Steady Thermal Field Simulation of Forced Air-cooled Column-type Air-core Reactor

    Institute of Scientific and Technical Information of China (English)

    DENG Qiu; LI Zhenbiao; YIN Xiaogen; YUAN Zhao

    2013-01-01

    Modeling the steady thermal field of the column-type air-core reactor,and further analyzing its distribution regularity,will help optimizing reactor design as well as improving its quality.The operation mechanism and inner insulation structure of a novel current limiting column-type air-core reactor is introduced in this paper.The finite element model of five encapsulation forced air-cooled column type air-core reactor is constructed using Fluent.Most importantly,this paper present a new method that,the steady thermal field of reactor working under forced air-cooled condition is simulated without arbitrarily defining the convection heat transfer coefficient for the initial condition; The result of the thermal field distribution shows that,the maximum steady temperature rise of forced air-cooled columntype air-core reactor happens approximately 5% to its top.The law of temperature distribution indicates:In the 1/3part of the reactor to its bottom,the temperature will rise rapidly to the increasing of height,yet the gradient rate is gradually decreasing; In the 5 % part of the reactor to its top,the temperature will drop rapidly to the increasing of height; In the part between,the temperature will rise slowly to the increasing of height.The conclusion draws that more thermal withstand capacity should be considered at the 5 % part of the reactor to its top to achieve optimal design solution.

  18. 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.

  19. 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.

  20. Building of Nuclear Ship Engineering Simulation System development of the simulator for the integral type reactor

    International Nuclear Information System (INIS)

    JAERI had carried out the design study of a light-weight and compact integral type reactor of power 100 MWth with passive safety as a power source for the future nuclear ships, and completed an engineering design. To confirm the design and operation performance and to utilize the study of automation of the operations of reactor, we developed a real-time simulator for the integral type reactor. This simulator is a part of Nuclear Ship Engineering Simulation System (NESSY) and on the same hardware as 'Mutsu' simulator which was developed to simulate the first Japanese nuclear ship Mutsu'. Simulation accuracy of 'Mutsu' simulator was verified by comparing the simulation results With data got in the experimental voyage of 'Mutsu'. The simulator for the integral type reactor uses the same programs which were used in 'Mutsu' simulator for the separate type PWR, and the simulated results are approximately consistent with the calculated values using RELAP5/MOD2 (The later points are reported separately). Therefore simulation accuracy of the simulator for the integral type reactor is also expected to be reasonable, though it is necessary to verify by comparing with the real plant data or experimental data in future. We can get the perspectives to use as a real-time engineering simulator and to achieve the above-mentioned aims. This is a report on development of the simulator for the integral type reactor mainly focused on the contents of the analytical programs expressed the structural features of reactor. (author)

  1. Reactor pressure vessels as type B transport containment boundaries

    Energy Technology Data Exchange (ETDEWEB)

    Nickell, R.E. [Applied Science and Technology, Inc., Poway, CA (United States); Griesbach, T.J. [ATI Consulting, Danville, CA (United States)

    1998-07-01

    Transportation risk and personnel exposure, as well as the cost of decommissioning nuclear power plants, can all be reduced significantly through the one-time use of the reactor pressure vessel as a containment boundary for shipping the activated internal components from the reactor site to a burial site. In order to help provide the technical basis for this end-use application, the ASME Board on Nuclear Codes and Standards, through its Subcommittee XI, has prepared a draft nuclear code case that contains requirements for any modifications to the vessel, including materials, design, fabrication, and examination. In particular, the requirements for evaluation of potential brittle fracture as the result of potentially low ambient shipping temperatures combined with hypothetical transportation accident loading are addressed. Existing ASME Code Section XI rules for linear elastic fracture mechanics evaluation of irradiated reactor pressure vessels have been adapted and included in the code case. (authors)

  2. Capital cost evaluation of liquid metal reactor by plant type - comparison of modular type with monolithic type -

    International Nuclear Information System (INIS)

    A preliminary economic comparison study was performed for KALIMER(Korea Advanced LIquid MEtal Reactor)between a modular plant type with 8 150MWe modules and a 1200MWe monolithic plant type. In both cases of FOAK (First-Of-A-Kind) Plant and NOAK (Nth-Of-A-Kind) Plant, the result says that the economics of monolithic plant is superior to its modular plant. In case of NOAK plant comparison, however, the cost difference is not significant. It means that modular plant can compete with monolithic plant in capital cost if it makes efforts of cost reduction and technical progress on the assumption that the same type of NOAK plant will be constructed continuously

  3. Power distribution monitoring and control in the RBMK type reactors

    International Nuclear Information System (INIS)

    Considered are the structures of monitoring and control systems for the RBMK-1000 reactor including three main systems with high independence: the control and safety system (CSS); the system for physical control of energy distribution (SPCED) as well as the Scala system for centralized control (SCC). Main functions and peculiarities of each system are discussed. Main attention is paid to new structural solutions and new equipment components used in these systems. Described are the RBMK operation software and routine of energy distribution control in it. It is noted that the set of reactor control and monitoring systems has a hierarchical structure, the first level of which includes analog systems (CSS and SPCED) normalizing and transmitting detector signals to the systems of the second level based on computers and realizing computer data processing, data representation to the operator, automatic (through CSS) control for energy distribution, diagnostics of equipment condition and local safety with provision for existing reserves with respect to crisis and thermal loading of fuel assemblies. The third level includes a power computer carrying out complex physical and optimization calculations and providing interconnections with the external computer of power system. A typical feature of the complex is the provision of local automatic safety of the reactor from erroneous withdrawal of any control rod. The complex is designed for complete automatization of energy distribution control in reactor in steady and transient operation conditions

  4. Materials and technology problems of WWER type reactors

    International Nuclear Information System (INIS)

    The symposium heard 29 papers all of which were inputted in INIS. The papers dealt with the chemical composition, metallurgy and mechanical properties of steels used for the manufacture of pressure vessels of nuclear reactors. The reliability was assessed of welded joints and the development and elimination of cracks under overlays dealt with. (E.S.)

  5. Construction of real-type simulator reusing the equipments of the Musashi-reactor

    International Nuclear Information System (INIS)

    The Musashi-reactor, the TRIGA-II type used to conduct various research including neutron capture therapy for cancer, was obliged to be shutdown in December 1989 due to the leak of primary coolant at reactor tank. Its decommissioning was decided in May 2003 and all spent fuels were transported to United States Department of Energy (USDOE). In order to contribute education and training of nuclear engineering and research on reactor instrumentation and control, real-type simulator reusing the consoles, control rod drives and other equipment of the Musashi-reactor was developed and mockup core and fuels was prepared as well as digital signal processors with a built-in measured reactor physics data. (T. Tanaka)

  6. Application of the SSYST-3 program system to WWER type nuclear reactors Pt. 1

    International Nuclear Information System (INIS)

    A computer code was developed for the simulation of reactor physical, thermohydraulical and chemical processes taking place in WWER-1000 type nuclear reactors. Two versions of this code, the SSYST-2 and SSYST-3 were compared with special attention to their data handling capabilities. The MULTRAN module of the SSYST-3 used for the calculation of Zircaloy fuel cladding oxidation was tested in detail. Some problems concerning the adaptation of SSYST-3 modules to WWER-type reactors were analyzed. 8 refs.; 4 tabs

  7. The influence of Argonaute proteins on alternative RNA splicing.

    Science.gov (United States)

    Batsché, Eric; Ameyar-Zazoua, Maya

    2015-01-01

    Alternative splicing of precursor RNAs is an important process in multicellular species because it impacts several aspects of gene expression: from the increase of protein repertoire to the level of expression. A large body of evidences demonstrates that factors regulating chromatin and transcription impact the outcomes of alternative splicing. Argonaute (AGO) proteins were known to play key roles in the regulation of gene expression at the post-transcriptional level. More recently, their role in the nucleus of human somatic cells has emerged. Here, we will discuss some of the nuclear functions of AGO, with special emphasis on alternative splicing. The AGO-mediated modulation of alternative splicing is based on several properties of these proteins: their binding to transcripts on chromatin and their interactions with many proteins, especially histone tail-modifying enzymes, HP1γ and splicing factors. AGO proteins may favor a decrease in the RNA-polymerase II kinetics at actively transcribed genes leading to the modulation of alternative splicing decisions. They could also influence alternative splicing through their interaction with core components of the splicing machinery and several splicing factors. We will discuss the modes of AGO recruitment on chromatin at active genes. We suggest that long intragenic antisense transcripts (lincRNA) might be an important feature of genes containing splicing events regulated by AGO.

  8. EFFECT OF PARTICLE TYPE ON CYCLONE FORMATION INSIDE A SOLAR REACTOR

    Directory of Open Access Journals (Sweden)

    Min-Hsiu Chien

    2016-07-01

    Full Text Available Solar reactors featuring a circulating cyclone flow pattern provide enhanced heat transfer and longer residence time increasing conversion efficiency. Cyclone flow also works in reducing particle deposition on solar reactor walls and exit which is particularly important issue in solar cracking reactors to avoid clogging. This paper focuses on the physics of cyclone formation inside a solar cracking reactor and experimentally analyzes the effect of particle entrainment on the flow pattern via two dimensional Particle Image Velocimetry (PIV. The cyclone flow structure in the reactor is reconstructed by capturing images from orientations perpendicular or parallel to the geometrical axis of the reactor. In order to conduct PIV measurements and to reconstruct the cyclone structure inside the solar reactor, the experiment was operated at room temperature with the flow configuration matching that of a solar reactor operating at high temperatures. Two types of seeding particles were tested, namely tri-ethylene glycol (TEG and solid carbon. The effectiveness of the screening flow was evaluated by measuring the quantity of solid particles deposit on the reactor walls. The Stokes flow analysis of each particle species was performed and the cyclone vector fields generated by using different particles are compared.

  9. Dynamic power behavior of a PWR type nuclear reactor

    International Nuclear Information System (INIS)

    A methodology for the power level evaluation (dynamic behavior) in a Pressurized Water Reactor, during a transient is developed, by solving the point kinetic equation related to the control rod insertion effects and fuel or moderator temperature 'feed-back'. A new version of the thermal-hydraulic code COBRA III P/MIT, is used. In this new version was included, as an option, the methodology developed. (E.G.)

  10. Fuel experience at a 37 year old TRIGA type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Boeck, H. [Atominstitut der Oesterreichischen Universitaeten, Wien (Austria)

    1999-07-01

    A survey is given on 37 years of TRIGA fuel experience at the 250 kW TRIGA Mark II reactor Vienna. Approximately 3000 fuel-years of experience have accumulated at this facility with only minor problems. Totally only 8 fuel elements had to be removed permanently from the core. Various inspection methods which have been developed throughout the years are described in this paper. (author)

  11. Concept of magnet systems for LHD-type reactor

    International Nuclear Information System (INIS)

    Heliotron reactors have attractive features for fusion power plants, such as no need for current drive and a wide space between the helical coils for the maintenance of in-vessel components. Their main disadvantage was considered necessarily the large size of their magnet systems. According to the recent reactor studies based on the experimental results in the Large Helical Device (LHD), the major radius of plasma of 14 to 17 m with a central toroidal field of 6 to 4 T is needed to attain the self-ignition condition with a blanket space thicker than 1.1 m. The magnetic stored energy is estimated at 120 to 140 GJ. Although both the major radius and the magnetic energy are about three times as large as ITER, the maximum magnetic field and mechanical stress can be comparable. In the preliminary structural analysis, the maximum stress intensity including the peak stress is less than 1000 MPa that is allowed for strengthened stainless steel. Although the length of the helical coil is longer than 150 m that is about five times as long as the ITER TF coil, cable-in-conduit conductors can be adopted with a parallel winding method of five-in-hand. The concept of the parallel winding is proposed. Consequently, the magnet systems for helical reactors can be realized with small extension of the ITER technology. (author)

  12. Concept of magnet systems for LHD-type reactor

    International Nuclear Information System (INIS)

    Heliotron reactors have attractive features for fusion power plants, such as no need for current drive and a wide space between the helical coils for the maintenance of in-vessel components. Their main disadvantage was considered the necessarily large size of their magnet systems. According to the recent reactor studies based on the experimental results in the Large Helical Device, the major radius of plasma of 14 to 17 m with a central toroidal field of 6 to 4 T is needed to attain the self-ignition condition with a blanket space thicker than 1.1 m. The stored magnetic energy is estimated at 120 to 140 GJ. Although both the major radius and the magnetic energy are about three times as large as ITER, the maximum magnetic field and mechanical stress can be comparable. In the preliminary structural analysis, the maximum stress intensity including the peak stress is less than 1,000 MPa that is allowed for strengthened stainless steel. Although the length of the helical coil is longer than 150 m that is about five times as long as the ITER TF coil, cable-in-conduit conductors can be adopted with a parallel winding method of five-in-hand. The concept of the parallel winding is proposed. Consequently, the magnet systems for helical reactors can be realized with small extension of the ITER technology. (author)

  13. A CANDU-type small/medium power reactor

    International Nuclear Information System (INIS)

    The assembly known as a CANDU power reactor consists of a number of standardized fuel channels or 'power modules'. Each of these channels produces about 5 thermal megawatts on average. Within practical limitations on fuel enrichment and ultimately on economics, the number of these channels is variable between about 50 and approximately 700. Small reactors suffer from inevitable disadvantages in terms of specific cost of design/construction as well as operating cost. Their natural 'niche' for application is in remote off-grid locations. At the same time this niche application imposes new and strict requirements for staff complement, power system reliability, and so on. The distinct advantage of small reactors arises if the market requires installation of several units in a coordinated installation program - a feature well suited to power requirements in Canada's far North. This paper examines several of the performance requirements and constraints for installation of these plants and presents means for designers to overcome the consequent negative feasibility factors.

  14. Characterization of hydrodynamics and mass transfer in two types of tubular electrochemical reactors

    International Nuclear Information System (INIS)

    Highlights: •The flow field of novel vertical-flow tubular electrochemical reactor with mesh electrodes (VTER) and traditional concentric tubular electrochemical reactor with plate electrodes (CTER) were compared. •The relationship between mass transfer coefficients and tube flow velocity and pressure drop in VTER and CTER were obtained. -- Abstract: Electrochemical treatment is an environmentally friendly method of removing pollutants from industrial wastewater. The tubular electrochemical reactor is one kind of electrochemical reactor. The current density distribution on the electrode surface in a traditional concentric tubular reactor is not homogeneous and the working area of the anodes and cathodes is unequal. Therefore, a novel tubular electrochemical reactor based on plug flow fluid orthogonal with mesh plate electrodes is presented. In this work, fluid flow and hydrodynamics of the vertical-flow tubular electrochemical reactor, such as velocity distribution and turbulent intensity distribution using computational fluid dynamics (CFD) method, are studied by comparing them to the traditional one. The electro-oxidation of phenol simulation wastewater treatment was developed to analyze the mass transfer performance of the two types of electrochemical reactors. In the novel tubular electrochemical reactor, due to the presence of mesh electrodes, the velocity distribution tended to be more homogeneous. In fact, the turbulent intensity clearly increased by 200% around the electrode surface. The kinetics of organic compounds removal in the novel tubular electrochemical reactor was also improved. Under the same flow rate, the improvement of the mass transfer coefficient for the novel tubular electrochemical reactor was more than twice that of the traditional tubular electrochemical reactor

  15. Variation and Evolution in the Glutamine-Rich Repeat Region of Drosophila Argonaute-2.

    Science.gov (United States)

    Palmer, William H; Obbard, Darren J

    2016-01-01

    RNA interference pathways mediate biological processes through Argonaute-family proteins, which bind small RNAs as guides to silence complementary target nucleic acids . In insects and crustaceans Argonaute-2 silences viral nucleic acids, and therefore acts as a primary effector of innate antiviral immunity. Although the function of the major Argonaute-2 domains, which are conserved across most Argonaute-family proteins, are known, many invertebrate Argonaute-2 homologs contain a glutamine-rich repeat (GRR) region of unknown function at the N-terminus . Here we combine long-read amplicon sequencing of Drosophila Genetic Reference Panel (DGRP) lines with publicly available sequence data from many insect species to show that this region evolves extremely rapidly and is hyper-variable within species. We identify distinct GRR haplotype groups in Drosophila melanogaster, and suggest that one of these haplotype groups has recently risen to high frequency in a North American population. Finally, we use published data from genome-wide association studies of viral resistance in D. melanogaster to test whether GRR haplotypes are associated with survival after virus challenge. We find a marginally significant association with survival after challenge with Drosophila C Virus in the DGRP, but we were unable to replicate this finding using lines from the Drosophila Synthetic Population Resource panel.

  16. ARGONAUTE1 acts in Arabidopsis root radial pattern formation independently of the SHR/SCR pathway.

    Science.gov (United States)

    Miyashima, Shunsuke; Hashimoto, Takashi; Nakajima, Keiji

    2009-03-01

    The formation of radially symmetric tissue patterns is one of the most basic processes in the development of vascular plants. In Arabidopsis thaliana, plant-specific GRAS-type transcription factors, SHORT-ROOT (SHR) and SCARECROW (SCR), are required for asymmetric cell divisions that separate two ground tissue cell layers, the endodermis and cortex, as well as for endodermal cell fate specification. While loss of SHR or SCR results in a single-layered ground tissue, radially symmetric cellular patterns are still maintained, suggesting that unknown regulatory mechanisms act independently of the SHR/SCR-dependent pathway. In this study, we identified a novel root radial pattern mutant and found that it is a new argonaute1 (ago1) allele. Multiple ago1 mutant alleles contained supernumerary ground tissue cell layers lacking a concentric organization, while expression patterns of SHR and SCR were not affected, revealing a previously unreported role for AGO1 in root ground tissue patterning. Analyses of ago1 scr double mutants demonstrated that the simultaneous loss of the two pathways caused a dramatic reduction in cellular organization and ground tissue identity as compared with the single mutants. Based on these results, we propose that highly symmetric root ground tissue patterns are maintained by the actions of two independent pathways, one using post-transcriptional regulation mediated by AGO1 and the other using the SHR/SCR transcriptional regulator.

  17. Dependence of the characteristics of bubbles on types of sonochemical reactors.

    Science.gov (United States)

    Yasui, Kyuichi; Tuziuti, Toru; Iida, Yasuo

    2005-01-01

    Computer simulations of bubble oscillations in liquid water irradiated by an ultrasonic wave have revealed that the characteristic of bubbles depends on types of sonochemical reactors: a horn-type reactor and a standing-wave type reactor. When the acoustic amplitude is large at 20 kHz, the bubble content is mostly water vapor even at the end of the bubble collapse and the temperature inside a bubble at the collapse is relatively low. On the other hand, when the acoustic amplitude is relatively low, the bubble content is mostly noncondensable gas at the end of the bubble collapse and the bubble temperature is relatively high. In a horn-type sonochemical reactor, the former type of bubbles are dominant because many bubbles exist near the horn-tip where the acoustic amplitude is large, while in a standing-wave type reactor the latter type of bubbles are dominant because the Bjerknes force gathers bubbles at a region where acoustic amplitude is relatively low.

  18. LMFBR type reactor core and its fuel exchange method

    Energy Technology Data Exchange (ETDEWEB)

    Ishibashi, Yoko; Koyama, Jun-ichi; Aoyama, Motoo; Haikawa, Katsumasa; Yamanaka, Akihiro

    1996-08-20

    Upon initial loading, two kinds of fuel assemblies including first fuel assemblies having a highest enrichment degree and second fuel assemblies having a lowest enrichment degree are loaded. The average fuel enrichment degree of an upper region of the first fuel assembly is made greater than that of the lower region. The reactivity of the lower region of the first fuel assembly is made lower than that of the upper portion to reduce power peak. Upon transfer from a first cycle to a second cycle, at least one of the second fuel assemblies is exchanged by the same number of the third fuel assemblies. In this case, an average fuel enrichment degree of the upper region of the third fuel assembly is made greater than that of the lower region to suppress the reactivity in the lower region of the third fuel assembly lower than the reactivity in the upper region thereby reducing the power peak. Thus, the upper power peak over the entire reactor core is moderated thereby capable of ensuring the reactor shut down margin without deteriorating the same. (N.H.)

  19. Gas/liquid separator for BWR type reactor

    International Nuclear Information System (INIS)

    A two phase gas/liquid flow generated at a heating portion of a nuclear reactor is swirled by inlet vanes. The phase gas/liquid flow uprises as a vortex flow in a vortex cylinder, and a liquid phase of a high density gathers at the outer circumference of the vortex cylinder. The liquid phase gathered at the outer circumference is collected at the inlet of a discharge flow channel which protrude into the vortex cylinder and in a three-step structure, and introduced into a recycling liquid phase passing through the discharge flow channel for liquid phase. There is provided a structure that separated liquid collected at the lowermost state in the inlet of the three-step discharge flow channel inlet descends in the discharge flow channel, then uprises in an uprising flow channel and is introduced into the recycling liquid phase by way of a discharge flow channel exit. The height of the discharge flow channel exit is determined equal to that of a liquid level of the recycling liquid phase during rated operation of the reactor. Accordingly, even in a case where the liquid level in the recycling liquid phase is lowered, the liquid level of the uprising flow channel is kept equal to that during rated operation. (I.N.)

  20. Modelling the WWER-type reactor dynamics using a hybrid computer. Part 1

    International Nuclear Information System (INIS)

    Results of simulation studies into reactor and steam generator dynamics of a WWER type power plant are presented. Spatial kinetics of the reactor core is described by a nodal approximation to diffusion equations, xenon poisoning equations and heat transfer equations. The simulation of the reactor model dynamics was performed on a hybrid computer. Models of both a horizontal and a vertical steam generator were developed. The dynamics was investigated over a large range of power by computing the transients on a digital computer. (author)

  1. The Status and Development Potential of Plate-Type Fuels for Research and Test Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Stahl, D.

    1979-03-01

    Recent U.S. Department of State action to restrict the shipment and use of highly enriched uranium for research and test reactors has renewed fuel development activity. The objective of these development activities is to increase the total uranium loading in the fuel meat so that enrichment reduction can be accomplished without significant performance penalties. This report characterizes the status and the potential for development of the currently utilized plate-type fuels for research and test reactors. The report also characterizes the newer high-density fuels which could be utilized in these reactors and indicates the impact of the utilization of both the new and current fuels on enrichment reduction.

  2. Building of Nuclear Ship Engineering Simulation System development of the simulator for the integral type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Takahashi, Teruo; Shimazaki, Junya; Yabuuchi, Noriaki; Fukuhara, Yosifumi; Kusunoki, Takeshi; Ochiai, Masaaki [Department of Nuclear Energy Systems, Tokai Research Establishment, Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan); Nakazawa, Toshio [Department of HTTR Project, Oarai Research Establishment, Japan Atomic Energy Research Institute, Oarai, Ibaraki (Japan)

    2000-03-01

    JAERI had carried out the design study of a light-weight and compact integral type reactor of power 100 MW{sub th} with passive safety as a power source for the future nuclear ships, and completed an engineering design. To confirm the design and operation performance and to utilize the study of automation of the operations of reactor, we developed a real-time simulator for the integral type reactor. This simulator is a part of Nuclear Ship Engineering Simulation System (NESSY) and on the same hardware as 'Mutsu' simulator which was developed to simulate the first Japanese nuclear ship Mutsu'. Simulation accuracy of 'Mutsu' simulator was verified by comparing the simulation results With data got in the experimental voyage of 'Mutsu'. The simulator for the integral type reactor uses the same programs which were used in 'Mutsu' simulator for the separate type PWR, and the simulated results are approximately consistent with the calculated values using RELAP5/MOD2 (The later points are reported separately). Therefore simulation accuracy of the simulator for the integral type reactor is also expected to be reasonable, though it is necessary to verify by comparing with the real plant data or experimental data in future. We can get the perspectives to use as a real-time engineering simulator and to achieve the above-mentioned aims. This is a report on development of the simulator for the integral type reactor mainly focused on the contents of the analytical programs expressed the structural features of reactor. (author)

  3. Study of the Utilization BWR Type Nuclear Power Reactor for Desalination Process

    International Nuclear Information System (INIS)

    The needs of fresh water increased by rapid population growth and industrials expansion, but these demands can not be prepared naturally. Following this case, seawater desalination becomes the primer option which can fulfill the need through the nuclear desalination technology. The coupled nuclear power reactor enables to supply thermal energy for auxiliary equipment and pumps operation. The utilization study of power reactor type BWR coupled with desalination process has been performed. The goal of study is to obtain characteristic data of desalted water specification which desalination system coupling with nuclear power plant produced energy for desalination process. The study is carried out by browsing data and information, and comprehensive review of thermal energy correlation between NPP with desalination process installation. According to reviewing are found that the thermal energy and electric power utilization from the nuclear power reactor are enable to remove the seawater to produce desalted water and also to operate auxiliary equipments. The assessment results is VK-300 reactor prototype, BWR type 250 MW(e) power are cogeneration unit can supplied hot steam temperature 285 °C to the extraction turbine to empower 150 MW electric power, and a part of hot steam 130 °C is use to operate desalination process and remind heat is distribute to the municipal and offices at that region. The coupled of VK-300 reactor power type BWR with desalination installation of MED type enable to produce desalted water with high quality distillate. Based on the economic calculation that the VK-300 reactor power of BWR type produced water distillate capacity is 300.000 m3/hour with cost US$ 0.58/m3. The coupling VK-300 reactor power type BWR with MED desalination plant is competitive economically. (author)

  4. A novel Y-type reactor for selective excitation of atmospheric pressure glow discharge plasma

    Science.gov (United States)

    Xia, Guan-Guang; Wang, Jin-Yun; Huang, Aimin; Suib, Steven L.; Hayashi, Yuji; Matsumoto, Hiroshige

    2001-02-01

    A novel Y-type atmospheric pressure ac glow discharge plasma reactor has been designed and tested in CO reduction with hydrogen and the reverse water-gas shift reaction. The reactor consists of a Y-type quartz tube with an angle of 120°-180° between the two long arms, two metal rod electrodes serving as high voltage terminals and two pieces of aluminum foil which were wrapped outside of the quartz tubes as a ground electrode. Different combinations of input power applied on this three- electrode system can lead to selective plasmas on one side, two sides, or can also generate a stable arc between the two high voltage terminal electrodes. The ability to selectively activate different species with this type of apparatus can help to minimize side reactions in plasmas to obtain desirable products. The Y-type reactor may provide a novel means to study fundamental problems regarding radical reactions.

  5. Fast Traveling-Wave Reactor of the Channel Type

    CERN Document Server

    Rusov, Vitaliy D; Vashchenko, Volodymyr N; Chernezhenko, Sergei A; Kakaev, Andrei A; Pantak, Oksana I

    2015-01-01

    The main aim of this paper is to solve the technological problems of the TWR based on the technical concept described in our priority of invention reference, which makes it impossible, in particular, for the fuel claddings damaging doses of fast neutrons to excess the ~200 dpa limit. Thus the essence of the technical concept is to provide a given neutron flux at the fuel claddings by setting the appropriate speed of the fuel motion relative to the nuclear burning wave. The basic design of the fast uranium-plutonium nuclear traveling-wave reactor with a softened neutron spectrum is developed, which solves the problem of the radiation resistance of the fuel claddings material.

  6. Reactor physics calculations and their experimental validation for conversion and upgrading of a typical swimming pool type research reactor

    International Nuclear Information System (INIS)

    Detailed neutronic analysis of a typical swimming pool type research reactor, Pakistan Research Reactor-1 (PARR-1), was carried out for conversion of its core from 93% highly enriched uranium to 20% low enriched uranium fuel with power upgrading from 5 to 10 MW. Standard computer codes WIMS-D/4 and CITATION were employed to calculate core excess reactivity, power defect, reactivity effect of xenon and samarium, reactivity worth of fuel element, worth of control rods, shutdown margin, reactivity feedback coefficients, neutron flux and power peaking factors. A series of low and high power tests were performed on the newly converted core to determine its performance. A comparison between the calculated and measured results is presented in this article. The agreement is generally good

  7. Argonautes promote male fertility and provide a paternal memory of germline gene expression in C. elegans

    OpenAIRE

    Conine, Colin C.; Moresco, James J.; Gu, Weifeng; Shirayama, Masaki; Conte, Darryl; Yates, John R.; Mello, Craig C

    2013-01-01

    During each life cycle germ cells preserve and pass on both genetic and epigenetic information. In C. elegans, the ALG-3/4 Argonaute proteins are expressed during male gametogenesis and promote male fertility. Here we show that the CSR-1 Argonaute functions with ALG-3/4 to positively regulate target genes required for spermiogenesis. Our findings suggest that ALG-3/4 functions during spermatogenesis to amplify a small-RNA signal that represents an epigenetic memory of male-specific gene expre...

  8. Special features of embrittlement of welded joints in shells of VVER-type reactors

    International Nuclear Information System (INIS)

    At present, the atomic power engineering of Russia and Ukraine is based on water-water energy reactors of the VVER-440 and VVER-1000 type, with the electric power of 440 and 1000 MW, respectively. The majority of the VVER-440 reactors are installed in Russia, and VVER-1000 reactors operate in Ukraine. The reactors' shell (RS) is produced from cylindrical shells and a dished end welded together by circular joints under a flux. The RS of the VVER-440 reactor is produced from 15Kh3MFA steel, and the VVER-1000 reactors are produced from 15Kh2NMFA steel. The shell of the VVER-1000 reactor has an internal austenite coating. The condition of the RS metal is determined mainly by the critical brittleness temperature Tb at which the impact toughness of specimens with a sharp notch reaches 60 J/cm2. The energy reactors, working in western countries, are characterised by a service life of 40 years and discussion is being carried out to extend this lifetime to 60 years. The design service life of the domestic reactors varies from 30 (RS VVER-440) to 40 (RS VVER-1000) years. According to investigations, the service life of the shells of these reactors is restricted by the properties of welded joints which are characterised by higher susceptibility to embrittlement than that of the parent metal, especially due to a higher content in the weld of phosphorus (RS VVER-440) or nickel (RS VVER-1000). Therefore, some experts believe that the actual service life of the RS is shorter than the design life. The accurate evaluation of the service life of welded joints in the RS is very important for the safety of service and also in the economic aspects, because the unjustified decrease of the permissible service life and premature shutdown of units of the nuclear power station result in huge losses

  9. Reactor coolant pump type RUV for Westinghouse Electric Company LLC reactor AP1000 TM

    International Nuclear Information System (INIS)

    The RUV is a reactor coolant pump, specially designed for the Westinghouse Electric Company LLC AP1000 TM reactor. It is a hermetically sealed, wet winding motor pump. The RUV is a very compact, vertical pump/motor unit, designed to fit into the compartment next to the reactor pressure vessel. Each of the two steam generators has two pump casings welded to the channel head by the suction nozzle. The pump/motor unit consists of a pump part, where a semi-axial impeller/diffuser combination is mounted in a one-piece pump casing. Computational Fluid Dynamics methods combined with various hydraulic tests in a 1:2 scale hydraulic test assure full compliance with the specific customer requirements. A short and rigid shaft, supported by a radial bearing, connects the impeller with the high inertia flywheel. This flywheel consists of a one-piece forged stainless steel cylinder, with an option for several smaller heavy metal cylinders inside. The flywheel is located inside the thermal barrier, which forms part of the pressure boundary. A specific arrangement of cooling water circuits guarantees a homogeneous temperature distribution in and around the flywheel, minimizes the friction losses of the flywheel and protects the motor from hot coolant. The driving torque is transmitted by the motor shaft, which itself is supported by two radial bearings. A three-phase, high-voltage squirrel-cage induction motor generates the driving torque. Due to the wet winding concept it is possible to achieve positive effects regarding motor lifetime. The cooling water is forced through the stator windings and the gap between rotor and stator by an auxiliary impeller. Furthermore, this wet winding motor concept has higher efficiency as compared to a canned motor since there are no eddy current losses. As part of the design process and in addition to the hydraulic scale model, a complete half scale model pump was built. It was used to verify the calculations performed like coast

  10. Development of new CRDM of magnetic jack type with enhanced parameters for VVER-1000 type of reactor

    International Nuclear Information System (INIS)

    The linear control rod drive mechanism used in WWER-type reactors was modernized by SKODA NUCLEAR MACHINERY Ltd. The modernization took place in two consecutive stages. The individual tasks comprised by the two stages are listed, and the verification of the results of upgrading is briefly described. (A.K.)

  11. Activity build-up in pressure vessel type reactors

    International Nuclear Information System (INIS)

    A simplified model is presented which permits the calculation of the average activity on the fuel elements of a reactor which operates under continuous refuelling, based on the assumption of crud interchange between fuel element surface and coolant in the form of particulate material only and using the crud specific activity as an empirical parameter determined experimentally. The net activity flux from core to out-of-core components is then calculated in the form of parametric curves depending upon crud specific activity and rate particulate release from fuel surface. The contribution to out-of-core radionuclide inventory arising in the release of activated material from core components is then assessed, and a way to estimate it numerically is presented. This method is based on experimentally determined cobalt-contents of structural materials and crud, and is specially suitable when high-cobalt alloys are present in-core. Activation of crud and release of activated materials are compared and it is shown that it is very likely that the latter may represent a sizable (and even the largest) fraction of the total cobalt activity. The use of the ratio of activities of 59 Fe to 54 Mn as a diagnostic tool for in-situ activation of structural materials is discussed. (author)

  12. Extension and application of the reactor dynamics code DYN3D for Block-type High Temperature Reactors

    International Nuclear Information System (INIS)

    The reactor code DYN3D was developed at the Helmholtz-Zentrum Dresden-Rossendorf to study steady state and transient behavior of Light Water Reactors. Concerning the neutronics part, the multigroup diffusion or SP3 transport equation based on nodal expansion methods is solved both for hexagonal and square fuel element geometry. To deal with Block-type High Temperature Reactor cores DYN3D was extended to a version DYN3D-HTR. A 3D heat conduction model was introduced to include 3D effects of heat transfer and heat conduction and the detailed structure of the fuel element. Homogenized neutronic cross sections were generated by applying a Monte Carlo approach with resolution of each individual TRISO fuel particle. Results of coupled steady state and transient calculations with 12 energy groups are presented. Transient case studies are control rod insertion, a change of the inlet coolant temperature and a change of the coolant gas mass flow rate. It is shown that DYN3D-HTR is an appropriate code system to simulate steady states and short time transients. Furthermore the necessity of the 3D heat conduction model is demonstrated

  13. Fuels and fission products clean up for molten salt reactor of the incinerator type

    Energy Technology Data Exchange (ETDEWEB)

    Ignatiev, V.; Gorbunov, V.; Zakirov, R. [RRC-Karchatov Institute, Moscow (Russian Federation)

    2000-07-01

    The objective of this paper is to discuss the feasibility of molten salt reactor technology for treatment of plutonium, minor actinides and fission products, when the reactor and fission product cleanup unit are planned as an integral system. This contribution summarizes the reasons which led to selection of the salt compositions for the molten salt reactor of the TRU incinerator type (MSB). Special characteristics of behavior of TRUs and fission products during power operation of MSB concepts are presented. The present paper briefly reviews the processing developments underlying the prior molten salt reactor (MSR) programs and relates then to the separation requirements for the MSB concept. Status and development needs in the thermodynamic properties of fluorides and fission product cleanup methods (with emphasis on actinides-lanthanides separation) are discussed. (authors)

  14. Fuels and fission products clean up for molten salt reactor of the incinerator type

    International Nuclear Information System (INIS)

    The objective of this paper is to discuss the feasibility of molten salt reactor technology for treatment of plutonium, minor actinides and fission products, when the reactor and fission product cleanup unit are planned as an integral system. This contribution summarizes the reasons which led to selection of the salt compositions for the molten salt reactor of the TRU incinerator type (MSB). Special characteristics of behavior of TRUs and fission products during power operation of MSB concepts are presented. The present paper briefly reviews the processing developments underlying the prior molten salt reactor (MSR) programs and relates then to the separation requirements for the MSB concept. Status and development needs in the thermodynamic properties of fluorides and fission product cleanup methods (with emphasis on actinides-lanthanides separation) are discussed. (authors)

  15. On the neutron spatial distribution in ionization chamber channels of the WWER type reactors

    International Nuclear Information System (INIS)

    Results of experimental and calculational studies permitting to estimate the neutron flux spatial distribution in ionization chamber channels of the commercial WWER-1000 and WWER-440 reactors and also of the WWER-440 reactor with water biological shield are presented. The integral neutron flux density distribution along the channel cross section approximately at height of the core middle and the corresponding thermal and fast neutron flux density distributions are measured by the activation detectors. It is shown that the difference in fast neutron flux density exceeds that of thermal neutrons. The commercial WWER-1000 type reactor the fast neutron flux density is decreased by the factor of 1.7, and thermal neutron flux density - by the factor of 1.2, for the commercial WWER-440 reactor these values are 1.37 and 1.18, and for the WWER-440 one with water shield - 1.5 and 1.18

  16. Uses of Plutonium Fuel in Pressure-Tube-Type, Heavy-Water-Moderated Thermal Reactors

    International Nuclear Information System (INIS)

    In 1962, a feasibility study was begun in the JAERI on the uses of various nuclear fuels for pressure-tube-type, heavy-water-moderated thermal reactors. This study began with analysis of the use of uranium in heavy-water-moderated thermal reactors such as the CANDU-PHW, CANDU-BLW, SGHW, EL-4, and Ref. 15, D and E lattices, which is designed in the JAERI, from the standpoint of the core design. Then, the ways of using plutonium fuel in the same types were investigated using WATCHTOWER, FLARE and VENUS codes, including: (1) direct substitution of the plutonium from light-water reactors or Magnox reactors, (2) recycle use of the plutonium from heavy-water-moderated reactors, (3) plutonium self-sustaining cycle, and (4) plutonium phoenix fuel. The following conclusions are reported: (1) In the direct substitution of plutonium, somewhat depleted plutonium is more suitable for core design than the plutonium from Magnox reactors or light-water reactors, because the increase in the initial reactivity due to large plutonium absorption cross-section must be prevented. (2) In the plutonium self-sustaining cycle, the fuel burn-up of about 15 000 ∼20000 MWd/t would be expected from natural uranium, and the positive void reactivity which always occurs in the uraniumloaded SGHW or CANDU-BLW lattices is greatly reduced, the latter property giving some margin to bum-out heat flux. (3) It may be concluded from the fuel cycle analysis that the plutonium self-sustaining cycle is equivalent to using slightly enriched uranium (about 1.0 at.%). It may be concluded that the use of plutonium in heavy-water-moderated reactors is technologically feasible and economically advantageous. (author)

  17. Use of Stable Noble Gases as a Predictor of Reactor Fuel Type and Exposure

    International Nuclear Information System (INIS)

    Ensuring spent reactor fuel is not produced to provide weapons-grade plutonium is becoming a major concern as many countries resort to nuclear power as a solution to their energy problems. Proposed solutions range from the development of proliferation resistant fuel to continuous monitoring of the fuel. This paper discusses the use of the stable isotopes of the fissiogenic noble gases, xenon and krypton, for determining the burnup characteristics, fuel type, and the reactor type of the fuel from which the sample was obtained. The gases would be collected on-stack as the fuel is reprocessed, and thus confirm that the fuel is as declared

  18. RMB. The new Brazilian multipurpose research reactor

    International Nuclear Information System (INIS)

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

  19. RMB. The new Brazilian multipurpose research reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-01-15

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

  20. A nuclear desalination complex with a VK-300 boiling type reactor facility

    International Nuclear Information System (INIS)

    RDIPE has developed a detailed design of an enhanced safety nuclear steam supply system (NSSS) with a VK-300 boiling water reactor for combined heat and power generation. The thermal power of the reactor is 750 MW. The maximum electrical power in the condensation mode is 250 MWe. The maximum heat generation capacity of 400 Gcal/h is reached at 150 MWe. This report describes, in brief, the basic technical concepts for the VK-300 NSSS and the power unit, with an emphasis on enhanced safety and good economic performance. With relatively small power, good technical and economic performance of the VK-300 reactor that is a base for the desalination complex is attained through: reduced capital costs of the nuclear plant construction thanks to technical approaches ensuring maximum simplicity of the reactor design and the NSSS layout; a single-circuit power unit configuration (reactor-turbine) excluding expensive equipment with a lot of metal, less pipelines and valves; reduced construction costs of the basic buildings thanks to reduced construction volumes due to rational arrangement concepts; higher reliability of equipment and reduced maintenance and repair costs; longer reactor design service life of up to 60 years; selection of the best reactor and desalination equipment interface pattern. The report considers the potential application of the VK-300 reactor as a source of energy for distillation desalination units. The heat from the reactor is transferred to the desalination unit via an intermediate circuit. Comparison is made between variants of the reactor integration with desalination units of the following types: multi-stage flash (MSF technology); multi-effect distillation horizontal-tube film units of the DOU GTPA type (MED technology). The NDC capacity with the VK-300 reactor, in terms of distillate, will be more than 200,000 m3/day, with the simultaneous output of electric power from the turbine generator buses of around 150 MWe. The variants of the

  1. Evaluation of performance of Son Tek Argonaut acoustic doppler velocity log in tow tank and sea

    Digital Repository Service at National Institute of Oceanography (India)

    Joseph, A.; Madhan, R.; Mascarenhas, A.A.M.Q.; Desai, R.G.P.; VijayKumar, K.; Dias, M.; Tengali, S.; Methar, A.

    Performance of a 500-kHz, 3-beam downward-looking Sontex Argonaut acoustic Doppler velocity log (DVL) based on measurements at tow-tank and sea is addressed. Its accuracy and linearity under tow-tank measurements were largely scattered...

  2. Evaluation of performance of Son Tek Argonaut acoustic doppler velocity log in tow tank and sea

    Digital Repository Service at National Institute of Oceanography (India)

    Joseph, A.; Madhan, R.; Mascarenhas, A.A.M.Q.; Desai; VijayKumar, K.; Dias, M.; Tengali, S.; Methar, A.

    of Performance of SonTek Argonaut Acoustic Doppler Velocity Log in Tow Tank and Sea Antony Joseph, R. Madhan, Antonio Mascaranhas, R.G. Prabhu Desai, Vijaykumar, Mathew Dias, Suryakant Tengali, and Anand Methar National Institute of Oceanography, Dona Paula...

  3. Continuous adsorption and recovery of Cr(VI) in different types of reactors.

    Science.gov (United States)

    Bai, Sudha R; Abraham, T Emilia

    2005-01-01

    This study reports the results of experiments on continuous adsorption and desorption of Cr(VI) ions by a chemically modified and polysulfone-immobilized biomass of the fungus Rhizopus nigricans. A fixed quantity of polymer-entrapped biomass beads corresponding to 2 g of dry biomass powder was employed in packed bed, fluidized bed, and stirred tank reactor for monitoring the continuous removal and recovery of Cr(VI) ions from aqueous solution and synthetic chrome plating effluent. Parameters such as flow rate (5, 10 and 15 mL/min), inlet concentration of Cr(VI) ions (50, 100, 150 and 250 mg/L) and the depth of biosorbent packing (22.8, 11.2 and 4.9 cm) were evaluated for the packed bed reactor. The breakthrough time and the adsorption rates in the packed bed column were found to decrease with increasing flow rate and higher Cr inlet concentrations and to increase with higher depths of sorbent packing. To have a comparative analysis of Cr adsorption efficiency in different types of reactors, the fluidized bed reactor and stirred tank reactor were operated using the same quantities of biosorbent material. For the fluidized bed reactor, Cr(VI) solution of 100 mg/L was pumped at 5 mL/min and fluidized by compressed air at a flow rate of 0.5 kg/cm.(2) The stirred tank reactor had a working volume of 200 mL capacity and the inlet/outlet flow rate was 5 mL/min. The maximum removal efficiency (mg Cr/g biomass) was obtained for the stirred tank reactor (159.26), followed by the fluidized reactor (153.04) and packed bed reactor (123.33). In comparison to the adsorption rate from pure chromate solution, approximately 16% reduction was monitored for synthetic chrome plating effluent in the packed bed. Continuous desorption of bound Cr ions from the reactors was effective with 0.01 N Na(2)CO(3) and nearly 80-94% recoveries have been obtained for all the reactors. PMID:16321053

  4. Thermal-hydraulic experiments of an advanced PIUS-type reactor

    International Nuclear Information System (INIS)

    The author constructed a semi-large scale experimental apparatus for simulating thermal-hydraulic behavior of the PIUS-type reactor with keeping the volumetric scaling ratio to the realistic reactor model. Fundamental experiments such as a steady state operation and a pump trip simulation were reported in ICONE-3(1995). In this paper the authors present two main results. One is a feedback control system using the upper density lock, and a start up simulation based on the non-uniform heating for both the primary loop and the poison loop. The other is a control system of small scale sub-loop attached to the poison loop in order to establish PIUS principle on the realistic operation of the PIUS-type reactor

  5. Conceptual design of magnets with CIC conductors for LHD-type reactors FFHR2m

    International Nuclear Information System (INIS)

    LHD-type reactors have attractive features for fusion power plants, such as no requirement of a current drive and a wide space between the helical coils for the maintenance of in-vessel components. One disadvantage was considered the requirement of a large major radius to attain the self-ignition condition with a sufficient space for blankets. According to the recent reactor studies based on experimental results in LHD, the major radius of plasma is set at 14 to 17 m with the central toroidal field of 6 to 4 T. The stored magnetic energy is estimated at 120 to 130 GJ. Both the major radius and the magnetic energy are three times as large as those for ITER. We intend to summarize the requirements for superconducting magnets of the LHD-type reactors and propose a conceptual design of the magnets with cable-in-conduit (CIC) conductors based on the technology for ITER. (author)

  6. Theoretical study for ICRF sustained LHD type p-{sup 11}B reactor

    Energy Technology Data Exchange (ETDEWEB)

    Watanabe, Tsuguhiro (ed.)

    2003-04-01

    This is a summary of the workshop on 'Theoretical Study for ICRF Sustained LHD Type p-{sup 11}B Reactor' held in National Institute for Fusion Science (NIFS) on July 25, 2002. In the workshop, study of LHD type D-{sup 3}He reactor is also reported. A review concerning the advanced nuclear fusion fuels is also attached. This review was reported at the workshop of last year. The development of the p-{sup 11}B reactor research which uses the LHD magnetic field configuration has been briefly summarized in section 1. In section 2, an integrated report on advanced nuclear fusion fuels is given. Ignition conditions in a D-{sup 3}He helical reactor are summarized in section 3. 0-dimensional particle and power balance equations are solved numerically assuming the ISS95 confinement law including a confinement factor ({gamma}{sub HH}). It is shown that high average beta plasma confinement, a large confinement factor ({gamma}{sub HH} > 3) and the hot ion mode (T{sub i}/T{sub e} > 1.4) are necessary to achieve the ignition of the D-{sup 3}He helical reactor. Characteristics of ICRF sustained p-{sup 11}B reactor are analyzed in section 4. The nuclear fusion reaction rate < {sigma}{upsilon} > is derived assuming a quasilinear plateau distribution function (QPDF) for protons, and an ignition condition of p-{sup 11}B reactor is shown to be possible. The 3 of the presented papers are indexed individually. (J.P.N.)

  7. Controlled thermonuclear fusion in TOKAMAK type reactors, the European example: Joint European Torus (JET)

    International Nuclear Information System (INIS)

    The development of controlled thermonuclear reaction in TOKAMAK type reactors, and the main projects in the world are presented. The main characteristics of the JET (Joint European Torus) program, the perspectives for energy production, and the international cooperation for viable use of the TOKAMAK are analysed. (M.C.K.)

  8. Optimized core design and fuel management of a pebble-bed type nuclear reactor

    NARCIS (Netherlands)

    Boer, B.

    2009-01-01

    The core design of a pebble-bed type Very High Temperature Reactor (VHTR) is optimized, aiming for an increase of the coolant outlet temperature to 1000 C, while retaining its inherent safety features. The VHTR has been selected by the international Generation IV research initiative as one of the si

  9. Effect of reactivity insertion rate on peak power and temperatures in swimming pool type research reactor

    International Nuclear Information System (INIS)

    It is essential to study the reactor behavior under different accidental conditions and take proper measures for its safe operation. We have studied the effect of reactivity insertion, with and without scram conditions, on peak power and temperatures of fuel, cladding and coolant in typical swimming pool type research reactor. The reactivity ranging from 1 $ to 2 $ and insertion times from 0.25 to 1 second have been considered. The computer code PARET has been used and results are presented in this article. (author)

  10. Conceptual Design of Magnets with CIC Conductors for LHD-type Reactors FFHR2m

    OpenAIRE

    Imagawa, Shinsaku; SAGARA, Akio; Kozaki, Yasuji

    2008-01-01

    LHD-type reactors have attractive features for fusion power plants, such as no requirement of a current drive and a wide space between the helical coils for the maintenance of in-vessel components. One disadvantage was considered the requirment of a large major radius to attain the self-ignition condition with a sufficient space for blankets. According to the recent reactor studies based on experimental results in LHD, the major radius of plasma is set at 14 to 17 m with the central toroidal ...

  11. Effects of reactor type and mass transfer on the morphology of CuS and ZnS crystals

    NARCIS (Netherlands)

    Al-Tarazi, M.; Heesink, A. Bert M.; Versteeg, G. F.

    2005-01-01

    For the precipitation of CuS and ZnS, the effects of the reactor/precipitator type, mass transfer and process conditions on crystal morphology were studied. Either H2S gas or a S2- solution were applied. Three different types of reactors have been tested, namely a laminar jet, a bubble column and an

  12. Radiation shielding considerations for the repair and maintenance of a swimming pool-type tokamak reactor

    International Nuclear Information System (INIS)

    The radiation shielding relevant to the repair and maintenance of a swimming pool-type tokamak reactor is considered. The dose rate during the reactor operation can be made low enough for personnel access into the reactor room if a 2m thick water layer is installed above the magnet cryostat. The dose rate 24 h after shutdown is such that the human access is allowed above the magnet cryostat. Sufficient water layer thickness is provided in the inboard space for the operation of automatic welder/cutter while retaining the magnet shielding capability. Some forced cooling is required for the decay heat removal in the first wall. The penetration shield thickness around the neutral beam injector port is estimated to be barely sufficient in terms of the magnet radiation damage. (orig.)

  13. Thermal hydraulics characterization of the core and the reactor vessel type BWR

    International Nuclear Information System (INIS)

    The thermal hydraulics design of a reactor type BWR 5 as the employees in the nuclear power plant of Laguna Verde involves the coupling of at least six control volumes: Pumps jet region, Stratification region, Core region, Vapor dryer region, Humidity separator region and Reactor region. Except by the regions of the core and reactor, these control volumes only are used for design considerations and their importance as operative data source is limited. It is for that is fundamental to complement the thermal hydraulics relations to obtain major data that allow to determine the efficiency of internal components, such as pumps jet, humidity separator and vapor dryer. Like example of the previous thing, calculations are realized on the humidity of the principal vapor during starting, comparing it with the values at the moment incorporated in the data banks of the computers of process of both units. (Author)

  14. Analysis of fuel management pattern of research reactor core of the MTR type design

    International Nuclear Information System (INIS)

    Research reactor core design needs neutronics parameter calculation use computer codes. Research reactor MTR type is very interested because can be used as research and also a radioisotope production. The research reactor in Indonesia right now is already 25 years old. Therefore, it is needed to design a new research reactor as a compact core. Recent research reactor core is not enough to meet criteria acceptance in the UCD which already determined namely thermal neutron flux in the core is 1.0x1015 n/cm2s. so that it is necessary to be redesign the alternative core design. The new research reactor design is a MTR type with 5x5 configuration core, uses U9Mo-Al fuel, 70 cm of high and uses two certainly fuel management pattern. The aim of this research is to achieve neutron flux in the core to meet the criteria acceptance in the UCD. Calculation is done by using WIMSD-B, Batan-FUEL and Batan-3DIFF codes. The neutronic parameters to be achieved by this calculation are the power level of 50 MW thermal and core cycle of 20 days. The neutronics parameter calculation is done for new U-9Mo-Al fuel with variation of densities.The result of calculation showed that the fresh core with 5x5 configuration, 360 gram, 390 gram and 450 gram of fuel loadings have meet safety margin and acceptance criteria in the UCD at the thermal neutron flux is more then 1.0 x 1015 n/cm2s. But for equilibrium core is only the 450 gram of loading meet the acceptance criteria. (author)

  15. SVBR-100 module-type fast reactor of the IV generation for regional power industry

    Science.gov (United States)

    Zrodnikov, A. V.; Toshinsky, G. I.; Komlev, O. G.; Stepanov, V. S.; Klimov, N. N.

    2011-08-01

    In the report the following is presented: basic conceptual provisions of the innovative nuclear power technology (NPT) based on modular fast reactors (FR) SVBR-100, summarized results of calculations of the reactor, analysis of the opportunities of multi-purpose application of such reactor facilities (RF) including export potentials with due account of nonproliferation requirements. The most important features of the proposed NPT analyzed in the report are as follows: (1) integral (monoblock) arrangement of the primary circuit equipment with entire elimination of the primary circuit pipelines and valves that considerably reduces the construction and assembly works period and coupling with high boiling point of lead-bismuth coolant (LBC) deterministically eliminates accidents of the LOCA type, (2) option for 100 MWe power and dimensions of the reactor provide: on the one hand, an opportunity to transport the reactor monoblock in factory-readiness by railway as well as other kinds of transport, on the other hand, core breeding ratio (CBR) exceeds 1 while MOX-fuel is used. The preferable area of application of RF SVBR-100 is regional and small power requiring power-units of electric power in a range of (100-600) MW, which could be used for cogeneration-based district heating while locating them nearby cities as well as for generation of electric power in a mode of load tracking in the regions with low network systems.

  16. In-pile modelling of nuclear fuel element for the MTR type reactors. Pt. 2

    Energy Technology Data Exchange (ETDEWEB)

    Farhadi, Kazem [AEOI, Tehran (Iran, Islamic Republic of). Radiations Application Research School

    2014-06-15

    In part two of the present paper, neutronic properties of the pool-type research reactor core are used to assess the similitude laws derived for out-of-pile modelling of the fuel element. The benchmark reactor used for this purpose is an IAEA 5 MW thermal pool-type research reactor currently in operation. The neutronic properties analysis are based on typical 2 200 m/sec and neutrons having 0.025 eV energy. The non-leakage capability of the system is estimated in terms of diffusion length. Also the slowing down power and the moderating ratio of the modelled methanol coolant are calculated in terms of lethargy of the diffusing medium. It is shown that the Iron which is substituted for Aluminium cladding is a relatively low absorber of neutrons but has a high neutron leakage. Methanol which replaced ordinary water as coolant is not a suitable coolant due to high neutrons absorbing substance. It is concluded that although Iron as a cladding material and methanol as a coolant meet the modelling out-of-pile criteria but are not satisfying neutronic properties. Therefore, use of them as a model clad and coolant are not suggested for research reactors. (orig.)

  17. The use of experimental data in an MTR-type nuclear reactor safety analysis

    Science.gov (United States)

    Day, Simon E.

    Reactivity initiated accidents (RIAs) are a category of events required for research reactor safety analysis. A subset of this is unprotected RIAs in which mechanical systems or human intervention are not credited in the response of the system. Light-water cooled and moderated MTR-type ( i.e., aluminum-clad uranium plate fuel) reactors are self-limiting up to some reactivity insertion limit beyond which fuel damage occurs. This characteristic was studied in the Borax and Spert reactor tests of the 1950s and 1960s in the USA. This thesis considers the use of this experimental data in generic MTR-type reactor safety analysis. The approach presented herein is based on fundamental phenomenological understanding and uses correlations in the reactor test data with suitable account taken for differences in important system parameters. Specifically, a semi-empirical approach is used to quantify the relationship between the power, energy and temperature rise response of the system as well as parametric dependencies on void coefficient and the degree of subcooling. Secondary effects including the dependence on coolant flow are also examined. A rigorous curve fitting approach and error assessment is used to quantify the trends in the experimental data. In addition to the initial power burst stage of an unprotected transient, the longer term stability of the system is considered with a stylized treatment of characteristic power/temperature oscillations (chugging). A bridge from the HEU-based experimental data to the LEU fuel cycle is assessed and outlined based on existing simulation results presented in the literature. A cell-model based parametric study is included. The results are used to construct a practical safety analysis methodology for determining reactivity insertion safety limits for a light-water moderated and cooled MTR-type core.

  18. Hydrogen energy recovery from high strength organic wastewater with ethanol type fermentation using acidogenic EGSB reactor

    Institute of Scientific and Technical Information of China (English)

    REN Nan-qi; GUO Wan-qian; WANG Xiang-jing; ZHANG Lu-si

    2005-01-01

    A lab-scale expanded granular sludge bed (EGSB) reactor was employed to evaluate the feasibility of the hydrogen energy recovery potential from high strength organic wastewater. The results showed that a maxioperation. At the acidogenic phase, COD removal rate was stable at about 15%. In the steady operation peri od, the main liquid end products were ethanol and acetic acid, which represented ethanol type fermentation. Among the liquid end products, the concentration percentage of ethanol and acetic acid amounted to 69.5% ~89.8% and the concentration percentage of ethanol took prominent about 51.7% ~ 59.1%, which is better than the utilization of substrate for the methanogenic bacteria. An ethanol type fermentation pathway was suggested in the operation of enlarged industrial continuous hydrogen bio-producing reactors.

  19. A neural networks based ''trip'' analysis system for PWR-type reactors

    International Nuclear Information System (INIS)

    The analysis short after automatic shutdown (trip) of a PWR-type nuclear reactor takes a considerable amount of time, not only because of the great number of variables involved in transients, but also the various equipment that compose a reactor of this kind. On the other hand, the transients'inter-relationship, intended to the detection of the type of the accident is an arduous task, since some of these accidents (like loss of FEEDWATER and station BLACKOUT, for example), generate transients similar in behavior (as cold leg temperature and steam generators mixture levels, for example). Also, the sequence-of-events analysis is not always sufficient for correctly pin point the causes of the trip. (author)

  20. Assessing optimal fermentation type for bio-hydrogen production in continuous-flow acidogenic reactors.

    Science.gov (United States)

    Ren, N Q; Chua, H; Chan, S Y; Tsang, Y F; Wang, Y J; Sin, N

    2007-07-01

    In this study, the optimal fermentation type and the operating conditions of anaerobic process in continuous-flow acidogenic reactors was investigated for the maximization of bio-hydrogen production using mixed cultures. Butyric acid type fermentation occurred at pH>6, propionic acid type fermentation occurred at pH about 5.5 with E(h) (redox potential) >-278mV, and ethanol-type fermentation occurred at pHClostridium sp., Propionibacterium sp. and Bacteriodes sp., respectively. Ethanol fermentation was optimal type by comparing the operating stabilities and hydrogen production capacities between the fermentation types, which remained stable when the organic loading rate (OLR) reached the highest OLR at 86.1kgCOD/m(3)d. The maximum hydrogen production reached up to 14.99L/d.

  1. A viral suppressor protein inhibits host RNA silencing by hooking up with Argonautes

    OpenAIRE

    Jin, Hailing; Zhu, Jian-Kang

    2010-01-01

    RNA viruses are particularly vulnerable to RNAi-based defenses in the host, and thus have evolved specific proteins, known as viral suppressors of RNA silencing (VSRs), as a counterdefense. In this issue of Genes & Development, Azevedo and colleagues (pp. 904–915) discovered that P38, the VSR of Turnip crinkle virus, uses its glycine/tryptophane (GW) motifs as an ARGONAUTE (AGO) hook to attract and disarm the host's essential effector of RNA silencing. Several GW motif-containing cellular pro...

  2. The criticality problem in reflected slab type reactor in the two-group transport theory

    International Nuclear Information System (INIS)

    The criticality problem in reflected slab type reactor is solved for the first time in the two group neutron transport theory, by singular eingenfunctions expansion, the singular integrals obtained through continuity conditions of angular distributions at the interface are regularized by a recently proposed method. The result is a coupled system of regular integral equations for the expansion coefficients, this system is solved by an ordinary interactive method. Numerical results that can be utilized as a comparative standard for aproximation methods, are presented

  3. Performance of static var compensator control type thyristor controlled reactor and thyristor switched capacitor

    Energy Technology Data Exchange (ETDEWEB)

    Araujo, Josias M. de; Yung, Chou Shaw; Rose, Eber H.; Pantoja, Antonio L.A. [ELETRONORTE, Belem, PA (Brazil); Fouesnant, Thomas; Boissier, Luc

    1994-12-31

    This paper has the objective of presenting the philosophy of Static Var Compensator (SVC) Control as well the necessary adjustments in the project of control system to guarantee suitable performance under different operating conditions. The verification on the performance of the SVC control has been done by Transient Network Analyzer (TNA/CEPEL) studies, commissioning tests and a factory tests. The SVC is the type of Thyristor Controlled Reactor (TCR) and Thyristor Switched Capacitor (TSC). (author) 3 refs., 12 figs.

  4. Regulatory use the classification security systems of I and C in VVER type reactors

    International Nuclear Information System (INIS)

    Presently work the author proposes a classification to the system I and C to the VVER 440 type reactor in categories the regulatory control with a view to establishing the degree to the attention that the regulator should pay to these systems, leaving the importance that have the same ones for the security the installation, during the execution the works that are carried out with this equipment in the stages construction, setting in service and exploitation

  5. A Study of the Energy Efficiency of Hadronic Reactors of Molecular Type

    OpenAIRE

    Aringazin, A. K.; Santilli, R. M.

    2001-01-01

    In this paper, we introduce an estimate of the "commercial efficiency" of Santilli's hadronic reactors of molecular type (Patented and International Patents Pending) which convert a liquid feedstock (such as automotive antifreeze and oil waste, city or farm liquid waste, crude oil, etc.) into the clean burning magnegas plus heat acquired by the liquid feedstock. The "commercial efficiency" is defined as the ratio between the total energy output (energy in magnegas plus heat) and the electric ...

  6. Argonaute-2 regulates the proliferation of adult stem cells in planarian

    Institute of Scientific and Technical Information of China (English)

    Yong-Qin Li; An Zeng; Xiao-Shuai Han; Chen Wang; Ge Li; Zhen-Chao Zhang; Jian-Yong Wang; Yong-Wen Qin; Qing Jing

    2011-01-01

    Dear Editor,Planarian Schmidtea mediterranea has extraordinary regeneration capabilities due to the abundance of adult stem cells (ASCs) known as neoblasts,which make planarian a powerful in vivo system to study ASC biology [ 1 ].The Argonaute (AGO) family proteins are defined by the presence of Piwi-Argonaute-Zwille (PAZ)) and PIWI domains [2],and mediate silencing via cleavage of mRNAs [3] or inhibition of translation [4].The AGO family proteins fall into two subfamilies,one named after Arabidopsis Argonaute and the other after Drosophila PIWI [5].In most organisms investigated so far,PIWI proteins bind Piwi-interacting RNAs (piRNAs) and are functionally involved in the regulation of germ cells [6,7].By contrast,AGO proteins have distinct roles in the smallRNA-mediated gene silencing pathway.In Drosophila,AGO1 has been illustrated to be engaged in the miRNA pathway,while AGO2 plays an important role in the siRNA-mediated gene regulation [4].

  7. Expression of human ARGONAUTE 2 inhibits endogenous microRNA activity in Arabidopsis

    Directory of Open Access Journals (Sweden)

    Ira eDeveson

    2013-04-01

    Full Text Available Plant and animal microRNA (miRNA pathways share many analogous components, the ARGONAUTE (AGO proteins being foremost among them. We sought to ascertain the degree of functional conservation shared by Homo sapiens ARGONAUTE 2 (HsAGO2 and Arabidopsis thaliana ARGONAUTE 1 (AtAGO1, which are the predominant AGO family members involved with miRNA activity in their respective species. Transgenic Arabidopsis plants expressing HsAGO2 were indistinguishable from counterparts over-expressing AtAGO1, each group exhibiting the morphological and molecular hallmarks of miRNA-pathway loss-of-function alleles. However, unlike AtAGO1, HsAGO2 was unable to rescue the ago1-27 allele. We conclude that, despite the evolutionary gulf between them, HsAGO2 is likely capable of interacting with some component/s of the Arabidopsis miRNA pathway, thereby perturbing its operation, although differences have arisen such that HsAGO2 alone is insufficient to confer efficient silencing of miRNA targets in planta.

  8. Load following tests on CANDU-type fuel elements in TRIGA research reactor of INR Pitesti

    International Nuclear Information System (INIS)

    Two load following (LF) tests on CANDU-type fuel elements were performed in the TRIGA Research Reactor of INR Pitesti, where the tests were designed to represent fuel in a CANDU reactor operating in a load following regime. In the first LF test the designated '78R' fuel element successfully experienced 367 power cycles, mostly between 23 and 56 kW/m average linear power. In the second LF test, developed under INR-AECL co-operation, the fuel element designated as 'ME01' withstood 200 power cycles from 27 to 54 kW/m average linear power, as well as additional ramps due to reactor trips and restarts during the test period. This experimental program is ongoing at INR Pitesti. Both LF tests were simulated with finite element computer codes in order to evaluate Stress Corrosion Fatigue (SCF) of the cladding arising from expansion and contraction of the pellets. New LF tests are planned to be performed in order to establish the limits and capabilities for CANDU fuel in LF conditions. This paper presents the results of the LF tests performed in the INR TRIGA Research Reactor compared with the analytical assessment for SCF conditions and their relation to CANDU fuel performance in LF conditions. (author)

  9. Role of VVER-type reactors in large-scale nuclear power of the XXI century

    International Nuclear Information System (INIS)

    Light water reactors (LWR) make over 85% of the world nuclear park and are presently constructed in 12 countries. One of the generally recognized LWR development directions is represented by VVER reactor concept, created and developed in the former Soviet Union. For over 35 years the VVER existence (with gross capacities ranging from 70 to 100 MWe), 58 power units have been built, and 49 are still in operation (13 in Russia and Ukraine each, 6 - in Bulgaria and Slovakia each, 4 - in Hungary and Czech Republic each, 2 - in Finland and 1 - in Armenia). The oldest of operating VVERs -unit 3 of Novovoronezh NPP in Russia - was connected to grid in 1971; the last - Mochovce-2 in Slovakia - was launched in 1999. Geography of VVER reactors is developing quite dynamically. For the first time this reactor type is being built in the countries of Asia: China and Iran, as well as in Cuba. Construction of the first VVER in India is also expected. (author)

  10. A High Operability Supervisory Digital System for TRIGA-Type Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Aronica, O.; Bove, R.; Cappelli, M.; Falconi, L.; Palomba, M.; Santoro, E.; Sepielli, M. [ENEA, UTFISST, Casaccia Research Center, Via Anguillarese, 301 Rome (Italy); Memmi, F. [University of Rome ' Roma Tre' , Department of Electrical Engineering, Via della Vasca Navale, 84 Rome (Italy)

    2011-07-01

    In this work, we propose an outline of a monitoring system to supervise variables coming from a fission nuclear reactor of TRIGA type (1-MW TRIGA reactor RC-1). The system can interface the control room instrumentation and can display the characteristic parameters (e.g. nuclear power, temperatures, flow rates, radiological parameters) in an intuitive, user-friendly way for plant operators. This aim is achieved using the Labview development environment. A front panel of a virtual instrument allows for a direct measure and a check that would not be possible by only reading the output data coming from the instruments of the control room, because of their standards and strict safety regulations. The acquisition system, for signals coming from the reactor, can process data and generate a detailed representation of the results. Statistics resulting from data analysis will be interpreted to optimize reactor management parameters. This system also includes a simulation tool to predict specific performances and investigate critical phenomena, or to optimize overall plant performances. In particular, it allows to have a feedback control and to perform predictive statistical surveys of all main process parameters. (author)

  11. Comparison of decommissioning options for the example of 2 research reactors of type TRIGA

    International Nuclear Information System (INIS)

    For decommissioning of nuclear facilities usually the two decommissioning strategies 'immediate dismantling' or 'deferred dismantling (safe enclosure)' are applied. In general, immediate dismantling is regarded as the more advantageous and more preferable option. Accordingly, immediate dismantling is the mostly selected option. Nevertheless, only in a case by case analysis it can be shown, which decommissioning option is the better one e. g. with respect to technical aspects or to a use of the facility / remaining facility. For two real decommissioning projects of two similar research reactors of TRIGA type GRS with support of the operator, German Cancer Research Center Heidelberg, performed a study on possible advantages of the two different strategies selected. While the first research reactor, TRIGA HD I, was dismantled immediately, the second research reactor, TRIGA HD II, was dismantled after a 20 years period of safe enclosure. As a result, it could be shown, that the selected different decommissioning strategies reflected the special conditions of each both research reactor in best way, so that a clear preference for one of the two decommissioning strategies can not be deduced. The slides of the presentation have been added at the end of the paper. (authors)

  12. Conceptual design study of Pebble Bed Type High Temperature Gas-cooled Reactor with annular core structure

    International Nuclear Information System (INIS)

    This report presents the Conceptual Design Study of Pebble Bed Type High Temperature Gas-cooled Reactor with Annular Core Structure. From this study, it is made clear that the thermal power of the Pebble Bed Type Reactor can be increased to 500MW through introducing the annular core structure without losing the inherent safe characteristics (in the coolant depressurization accident, the fuel temperature does not exceed the temperature where the fuel defect begins.) This thermal power is two times higher than the inherent safe Pebble Bed Type High temperature Gas-cooled Reactor (MHTGR) designed in West Germany. From this result, it is foreseen that the ratio of the plant cost to the reactor power is reduced and the economy of the plant operation is improved. The reactor performances e.g. fuel burnup and fuel temperature are maintained in same level of the MHTGR. (author)

  13. Neutron dosimetry. Environmental monitoring in a BWR type reactor; Dosimetria de neutrones. Monitoreo ambiental en un reactor del tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Tavera D, L.; Camacho L, M.E

    1991-01-15

    The measurements carried out on reactor dosimetry are applied mainly to the study on the effects of the radiation in 108 materials of the reactor; little is on the environmental dosimetry outside of the primary container of BWR reactors. In this work the application of a neutron spectrometer formed by plastic detectors of nuclear traces manufactured in the ININ, for the environmental monitoring in penetrations around the primary container of the unit I of the Laguna Verde central is presented. The neutron monitoring carries out with purposes of radiological protection, during the operational tests of the reactor. (Author)

  14. Development of core fuel management code system for WWER-type reactors

    Institute of Scientific and Technical Information of China (English)

    2006-01-01

    In this article, a core fuel management program for hexagonal pressurized water type WWER reactors (CFMHEX) has been developed, which is based on advanced three-dimensional nodal method and integrated with thermal hydraulic code to realize the coupling of neutronics and thermal-hydraulics. In CFMHEX, all these feedback effects such as burnup, power distribution, moderator density, and control rod insertion are considered. The verification and validation of the code system have been examined through the IAEA WWER-1000-type Kalinin NPP benchmark problem. The numerical results are in good agreement with measurements and are close to those of other international institutes.

  15. Evaluation of Advanced Thermohydraulic System Codes for Design and Safety Analysis of Integral Type Reactors

    International Nuclear Information System (INIS)

    The integral pressurized water reactor (PWR) concept, which incorporates the nuclear steam supply systems within the reactor vessel, is one of the innovative reactor types with high potential for near term deployment. An International Collaborative Standard Problem (ICSP) on Integral PWR Design, Natural Circulation Flow Stability and Thermohydraulic Coupling of Primary System and Containment during Accidents was established in 2010. Oregon State University, which made available the use of its experimental facility built to demonstrate the feasibility of the Multi-application Small Light Water Reactor (MASLWR) design, and sixteen institutes from seven Member States participated in this ICSP. The objective of the ICSP is to assess computer codes for reactor system design and safety analysis. This objective is achieved through the production of experimental data and computer code simulation of experiments. A loss of feedwater transient with subsequent automatic depressurization system blowdown and long term cooling was selected as the reference event since many different modes of natural circulation phenomena, including the coupling of primary system, high pressure containment and cooling pool are expected to occur during this transient. The power maneuvering transient is also tested to examine the stability of natural circulation during the single and two phase conditions. The ICSP was conducted in three phases: pre-test (with designed initial and boundary conditions established before the experiment was conducted), blind (with real initial and boundary conditions after the experiment was conducted) and open simulation (after the observation of real experimental data). Most advanced thermohydraulic system analysis codes such as TRACE, RELAPS and MARS have been assessed against experiments conducted at the MASLWR test facility. The ICSP has provided all participants with the opportunity to evaluate the strengths and weaknesses of their system codes in the transient

  16. Argonaute2 Protein in Rat Spermatogenic Cells Is Localized to Nuage Structures and LAMP2-Positive Vesicles Surrounding Chromatoid Bodies.

    Science.gov (United States)

    Fujii, Yuki; Onohara, Yuko; Fujita, Hideaki; Yokota, Sadaki

    2016-04-01

    Localization of Argonaute2 (AGO2) protein--an essential component for the processing of small interfering RNA (siRNA)-directed RNA interference (RNAi) in RNA-induced silencing complex (RISC) in nuage of rat spermatogenic cells--was evaluated by immunofluorescence microscopy (IFM) and immunoelectron microscopy (IEM). AGO2 was shown, for the first time, to be localized to four previously classified types of nuage: irregularly shaped perinuclear granules (ISPGs), intermitochondrial cement (IMC), satellite bodies (SBs), and chromatoid bodies (CBs). Dual IEM staining for AGO2/Maelstrom (MAEL) protein or AGO2/MIWI protein demonstrated that AGO2 is colocalized with MAEL or MIWI proteins in these types of nuage. Dual IFM and IEM staining of AGO2/lysosomal-associated membrane protein 2 (LAMP2) showed that CB in round spermatids are in contact with and surrounded by LAMP2-positive vesicles, whereas nuage in pachytene spermatocytes are not. Taken together, our findings indicate that: (i) AGO2 in pachytene spermatocytes functions in ISPGs, IMC, and SBs; (ii) AGO2 in round spermatids functions in CBs, and that CBs are associated with lysosomal compartments. PMID:27029769

  17. Computational simulation of fuel burnup estimation for research reactors plate type

    Energy Technology Data Exchange (ETDEWEB)

    Santos, Nadia Rodrigues dos, E-mail: nadiasam@gmail.com [Instituto Federal de Educacao, Ciencia e Tecnologia do Rio de Janeiro (IFRJ), Paracambi, RJ (Brazil); Lima, Zelmo Rodrigues de; Moreira, Maria de Lourdes, E-mail: zrlima@ien.gov.br, E-mail: malu@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2015-07-01

    The aim of this study is to estimate the spatial fuel burnup, through computational simulation, in two research reactors plate type, loaded with dispersion fuel: the benchmark Material Test Research - International Atomic Energy Agency (MTR-IAEA) and a typical multipurpose reactor (MR). The first composed of plates with uranium oxide dispersed in aluminum (UAlx-Al) and a second composed with uranium silicide (U{sub 3}Si{sub 2}) dispersed in aluminum. To develop this work we used the deterministic code, WIMSD-5B, which performs the cell calculation solving the neutron transport equation, and the DF3DQ code, written in FORTRAN, which solves the three-dimensional neutron diffusion equation using the finite difference method. The methodology used was adequate to estimate the spatial fuel burnup , as the results was in accordance with chosen benchmark, given satisfactorily to the proposal presented in this work, even showing the possibility to be applied to other research reactors. For future work are suggested simulations with other WIMS libraries, other settings core and fuel types. Comparisons the WIMSD-5B results with programs often employed in fuel burnup calculations and also others commercial programs, are suggested too. Another proposal is to estimate the fuel burnup, taking into account the thermohydraulics parameters and the Xenon production. (author)

  18. Power pulse tests on CANDU type fuel elements in TRIGA reactor of INR Pitesti

    International Nuclear Information System (INIS)

    Pulse irradiation tests on short fuel elements have been carried out in TRIGA Annular Core Pulse Reactor (TRIGA ACPR) of INR Pitesti to investigate aspects related to the thermal and mechanical behavior of CANDU type fuel elements under short duration and large amplitude power pulse conditions. Short test fuel elements were instrumented with thermocouples for cladding surface temperature measurements and pressure sensor for element internal pressure measurement. Transient histories of reactor power, cooling water pressure, fuel element internal pressure and cladding temperature were recorded during tests. The fuel elements were subjected to total energy deposition from 70 to 280cal g-1 UO2. Rapid fuel pellet expansion due to a power excursion caused radial and longitudinal deformation of the cladding. Cladding failure mechanism and the failure threshold have been established. This paper presents some recent results obtained from these power pulse tests performed in TRIGA ACPR of INR Pitesti. (author)

  19. A Study of the Energy Efficiency of Hadronic Reactors of Molecular Type

    CERN Document Server

    Aringazin, A K

    2001-01-01

    In this paper, we introduce an estimate of the "commercial efficiency" of Santilli's hadronic reactors of molecular type (Patented and International Patents Pending) which convert a liquid feedstock (such as automotive antifreeze and oil waste, city or farm liquid waste, crude oil, etc.) into the clean burning magnegas plus heat acquired by the liquid feedstock. The "commercial efficiency" is defined as the ratio between the total energy output (energy in magnegas plus heat) and the electric energy used for its production, while the "scientific efficiency" is the usual ratio between the total energy output and the total energy input (the sum of the electric energy plus the energy in the liquid feedstock as well as that in the carbon electrodes). A primary purpose of this paper is to show that conventional thermochemistry does indeed predict a commercial efficiency bigger than one, although their values is considerably smaller than the actual efficiency measured in the reactors, thus indicating the applicabili...

  20. Deployable nuclear fleet based on available quantities of uranium and reactor types – the case of fast reactors started up with enriched uranium

    Directory of Open Access Journals (Sweden)

    Baschwitz Anne

    2016-01-01

    Full Text Available International organizations regularly produce global energy demand scenarios. To account for the increasing population and GDP trends, as well as to encompass evolving energy uses while satisfying constraints on greenhouse gas emissions, long-term installed nuclear power capacity scenarios tend to be more ambitious, even after the Fukushima accident. Thus, the amounts of uranium or plutonium needed to deploy such capacities could be limiting factors. This study first considers light-water reactors (LWR, GEN III using enriched uranium, like most of the current reactor technologies. It then examines the contribution of future fast reactors (FR, GEN IV operating with an initial fissile load and then using depleted uranium and recycling their own plutonium. However, as plutonium is only available in limited quantity since it is only produced in nuclear reactors, the possibility of starting up these Generation IV reactors with a fissile load of enriched uranium is also explored. In one of our previous studies, the uranium consumption of a third-generation reactor like an EPR™ was compared with that of a fast reactor started up with enriched uranium (U5-FR. For a reactor lifespan of 60 years, the U5-FR consumes three times less uranium than the EPR and represents a 60% reduction in terms of separative work units (SWU, though its requirements are concentrated over the first few years of operation. The purpose of this study is to investigate the relevance of U5-FRs in a nuclear fleet deployment configuration. Considering several power demand scenarios and assuming different finite quantities of available natural uranium, this paper examines what types of reactors must be deployed to meet the demand. The deployment of light-water reactors only is not sustainable in the long run. Generation IV reactors are therefore essential. Yet when started up with plutonium, the number of reactors that can be deployed is also limited. In a fleet deployment

  1. Flow of kinetic parameters in a typical swimming pool type research reactor

    International Nuclear Information System (INIS)

    Calculations were performed to estimate the variation in kinetic parameters (delayed neutron fraction and prompt neutron generation time) in different core configurations of a typical swimming pool type research reactor. Pakistan research Reactor-1 (PARR-1) was employed for this study. The effect due to burnup of the core was also studied. Calculations were performed with the help of computer codes WIMSD/4 and CITATION. Precursors yield was modified according to the neutron flux averaging only. This is the simple way to calculate the precursor yield for a particular core. The kinetic parameters are different for different core configurations. The βeff decreases with 1.33 x 10-6/% burnup whereas prompt neutron generation time increases with 6.42 x 10-8 s/% burnup. The results were compared with safety analysis report and with published values and were found in good agreement. This study provides the confidence to understand the change in the kinetic parameters of research reactors with core change and also with burnup of the core

  2. Out-of-pile modelling of nuclear fuel elements for MTR type reactors. Pt. 1

    Energy Technology Data Exchange (ETDEWEB)

    Farhadi, Kazem [Nuclear Science and Technology Research Institute, Tehran (Iran, Islamic Republic of). Radiations Application Research School

    2014-03-15

    In the first part of the present paper, for a 5 MW thermal pool-type research reactor, the fuel element is modelled for when undergoing both natural circulation of the coolant and forced convection of the coolant operational conditions. First, the required dimensionless groups were identified and then the pertinent similarity criteria were derived accordingly. The derived similitude laws were modified under the conditions of identical pressure, identical temperature difference and identical coolant and fuel cladding in the model and the prototype. These modifications were done for the system under both natural and forced convections. The effect of varying cladding materials under normal operating conditions of the research reactor were observed via coolant channel thickness. Also the effect of a wider coolant channel on the nature of the coolant fluid was observed. The results obtained indicate that it is not possible to conserve all the dimensionless groups between the model and the prototype and hence achieve an errorless outcome. Among all the liquids available, methanol is the only liquid which nearly satisfies the thermal-hydraulic similitude and must be used in place of ordinary coolant water. This in turn necessitates the coolant channel to be wider and as a consequence the traditional Aluminium cladding in research reactors should be replaced by Iron. The derived scale down criteria can be used for the design of fuel element for the out-of-pile testing. (orig.)

  3. Conceptual design of swimming pool type tokamak power reactor (SPTR-P)

    International Nuclear Information System (INIS)

    A preliminary design study of a tokamak power reactor utilizing the deuterium/tritium/lithium fuel cycle based on a swimming pool type reactor (SPTR) concept is presented. Its primary aim is to investigate the characteristics of the swimming-pool concept in which water replaces much of the steel normally required for shielding. The major design features are: steady state operation, RF wave for plasma heating and current drive, solid tritium breeder material (Li2O), modified austenitic stainless steel as first wall and blanket structural material, pumped limiter for ash exhaust, unified assembling of blanket and vacuum vessel and pressurized water cooling. The huge and heavy solid shield structure protecting superconducting magnets which brings about great difficulties in repair and maintenance is eliminated by submerging the reactor in a water pool. The water plays a role of shielding. In addition the water shield concept reduces radioactive waste disposal and to ease radiation streaming shielding. Key design parameters are: net electric power of 1000 MW, fusion power of 3200 MW, neutron wall loading of 3.3 MW/m2, major radius of 6.9 m, plasma radius of 2.0 m, plasma elongation of 1.6, plasma current of 16 MA, total beta of 7 %, toroidal field on axis of 5.2 T. (author)

  4. Design and production process of bushing-type fuel elements for channel research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Afanasiev, V.L.; Aleksandrov, A.B.; Enin, A.A. [NZHK, Novosibirsk (Russian Federation)

    1998-07-01

    The design of bushing-type fuel elements (FEs) based on the dioxide fuel composition UO{sub 2}+Al for channel research reactors is described. Commercial technological process for bushing-type FEs with up to 0.8 g/cm{sup 3} uranium concentration in the fuel core is presented. This technology is based on fuel core production using powder metallurgy with subsequent chemical treatment of its surface and enclosing into the finished cladding. Commercial technological process for bushing-type FEs with 0.8-3.8 g/cm{sup 3} uranium concentration in the fuel composition is considered. This process is based on fuel core production by means of extrusion technology followed by fuel core enclosing into the cladding. (author)

  5. Chemistry aspects of the source term formation for a severe accident in a CANDU type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Constantin, A.; Constantin, M. [Institute for Nuclear Research, Pitesti (Romania)

    2013-07-15

    The progression of a severe accident in a CANDU type reactor is slow because the core is surrounded by a large quantity of heavy and light water which acts as a heat sink to remove the decay heat. Therefore, the source term formation is a complex and long process involving fission products transport and releasing in the fuel matrix, thermal hydraulics of the transport fluid in the primary heat system and containment, deposition and transport of fission products, chemistry including the interaction with the dousing system, structural materials and paints, etc. The source term is strongly dependent on initial conditions and accident type. The paper presents chemistry aspects for a severe accident in a CANDU type reactor, in terms of the retention in the primary heat system. After releasing from the fuel elements, the fission products suffer a multitude of phenomena before they are partly transferred into the containment region. The most important species involved in the deposition were identified. At the same time, the influence of the break position in the transfer fractions from the primary heat system to the containment was investigated. (orig.)

  6. MR-6 type fuel elements cooling in natural convection conditions after the reactor shut down

    Energy Technology Data Exchange (ETDEWEB)

    Pytel, K.; Bykowski, W.; Moldysz, A. [Institute of Atomic Energy, Otwock Swierk (Poland)

    2002-07-01

    Natural cooling conditions of the nuclear fuel in the channel type reactor after its shut down are commonly determined with relatively high uncertainty. This is not only to he lack of adequate measurements of thermal parameters i.e. the residual power generation, the coolant flow and temperatures, but also due to indeterminate model of convection mechanism. The numerical simulation of natural convection in multitube fuel assembly in the fuel channel leads to various convection modes including evidently chaotic behaviour. To determine the real cooling conditions in the MARIA research reactor a series of experiments has been performed with fuel assembly equipped with a set of thermocouples. After some forced cooling period (the shortest was half an hour after the reactor shut down) the reactor was left with the only natural convection. Two completely different cooling modes have been observed. The MARIA core consists of series of individual fuel channel and so called bypasses, maintaining the hydraulic properties of the fuel channel, connected in parallel. Initially, the convection cells were established trough few so-called bypasses providing a very effective mode of cooling. In this mode the flow charts were identical to those existing in forced cooling mode. After certain period the system switched on the second cooling mode with natural circulation within the individual fuel cells. Higher temperatures and temperature fluctuations were characteristic for this mode approaching 30 deg in amplitude. In almost all the cases the system was switching few times between modes, but eventually remained in the second mode. The switching times were not regular and the process has a chaotic behaviour. (author)

  7. Linear pulse motor type control element drive mechanism for the integral reactor

    International Nuclear Information System (INIS)

    The integral reactor SMART currently under development at Korea Atomic Energy Research Institute is designed with soluble boron free operation and use of nuclear heating for reactor startup. These design features require the Control Element Drive Mechanism (CEDM) for SMART to have fine-step movement capability as well as high reliability for the fine reactivity control. In this paper, design characteristics of a new concept CEDM driven by the Linear Pulse Motor (LPM) which meets the design requirements of the integral reactor SMART are introduced. The primary dimensions of the linear pulse motor are determined by the electro-magnetic analysis and the results are also presented. In parallel with the electro-magnetic analysis, the conceptual design of the CEDM is visualized and checked for interferences among parts by assembling three dimensional (3D) models on the computer. Prototype of LPM with double air-gaps for the CEDM sub-assemblies to lift 100 kg is designed, analysed, manufactured and tested to confirm the validity of the CEDM design concept. A converter and a test facility are manufactured to verify the dynamic performance of the LPM. The mover of the LPM is welded with ferromagnetic material and non-ferromagnetic material to get the magnetic flux path between inner stator and outer stator. The thrust forces of LPM predicted by analytic model have shown good agreement with experimental results from the prototype LPM. It is found that the LPM type CEDM has high force density and simple drive mechanism to reduce volume and satisfy the reactor operating circumstances with high pressure and temperature

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

    International Nuclear Information System (INIS)

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

  9. An analytic study on LBLOCA for CANDU type reactor using MARS-KS/CANDU

    International Nuclear Information System (INIS)

    This study provides the simulation results using MARS-KS/CANDU code for the Large Break LOCA of CANDU type reactor. The purpose of the study is to evaluate the capability of MARS-KS/CANDU for simulating the actual plants (Wolsong 2/3/4). The steady state and the transient analysis results were provided. After the sensitivity study depend on break size, the case that 35% of the inlet header known as the accident that has the most limiting effect on the temperature of the fuel sheath was calculated. In order to evaluate the results, the results were compared with those of CATHENA simulation. (author)

  10. Genomic identification, rapid evolution, and expression of Argonaute genes in the tilapia, Oreochromis niloticus.

    Science.gov (United States)

    Tao, Wenjing; Sun, Lina; Chen, Jinlin; Shi, Hongjuan; Wang, Deshou

    2016-09-01

    Argonaute proteins are key components of the small RNA-induced silencing complex and have multiple roles in RNA-directed regulatory pathways. Argonaute genes can be divided into two subfamilies: the Ago (interacting with microRNA/small interfering RNA) and Piwi subfamilies (interacting with piwi-interacting RNAs (piRNAs)). In the present study, genome-wide analyses firstly yielded the identification of different members of Agos and Piwis in the tilapia, coelacanth, spotted gar, and elephant shark. The additional teleost Ago3b was generated following the fish-specific genome duplication event. Selective pressure analysis on Agos and Piwis between cichlids and other teleosts showed an accelerated evolution of Piwil1 in the cichlid lineages, and the positive selected sites were located in the region of PIWI domain, suggesting that these amino acid substitutions are adapt to targeted cleavage of messenger RNA (mRNA) in cichlids. Ago1 and Ago4 were detected at higher levels at 5 days after hatching (dah) in both ovaries and testes compared with other stages, supporting the previously reported requirement of Ago-mediated pathways to clear the maternal mRNAs during the early embryogenesis. The Piwis were abundantly expressed in tilapia testes, indicating their essential roles in male germline, especially in spermatogenesis. Notable expression of Piwis was also detected in skeletal muscle, indicating that piRNA pathway may not only be confined to development and maintenance of the germline but may also play important roles in somatic tissues. The expression of Piwil1 and Piwil2 was examined by quantitative PCR (qPCR) and in situ hybridization (ISH) to validate the spatial and temporal expression profiles. Taken together, these results present a thorough overview of tilapia Argonaute family and provide a new perspective on the evolution and function of this family in teleosts. PMID:27491892

  11. Conceptual differences between existing and advanced reactors and criteria affecting the development of new types of nuclear power plants world-wide

    International Nuclear Information System (INIS)

    A comparison of the nuclear safety principles and the design and operating parameters between existing and advanced reactors is presented, and criteria affecting the development of new types of nuclear reactor are outlined

  12. Proposal of rectifier type superconducting fault current limiter with non-inductive reactor (SFCL)

    Science.gov (United States)

    Mohammad Salim, Khosru; Muta, Itsuya; Hoshino, Tsutomu; Nakamura, Taketsune; Yamada, Masato

    2004-03-01

    A rectifier type superconducting fault current limiter (SFCL) with non-inductive reactor has been proposed. The concept behind this SFCL is the appearance of high impedance during non-superconducting state of the coil. In a hybrid bridge circuit, two superconducting coils connected in anti-parallel: a trigger coil and a limiting coil. Both the coils are magnetically coupled with each other and have same number of turns. There is almost zero flux inside the core and therefore the total inductance is small during normal operation. At fault time when the trigger coil current reaches to a certain level, the trigger coil changes from superconducting state to normal state. This super-to-normal transition of the trigger coil changes the current ratio of the coils and therefore the flux inside the reactor is no longer zero. So, the equivalent impedance of both the coils increased thus limits the fault current. We have carried out computer simulation using EMTDC and observed the results. A preliminary experiment has already been performed using copper wired reactor with simulated super-to-normal transition resistance and magnetic switches. Both the simulation and preliminary experiment shows good results. The advantage of using hybrid bridge circuit is that the SFCL can also be used as circuit breaker. Two separate bridge circuit can be used for both trigger coil and the limiter coil. In such a case, the trigger coil can be shutdown immediately after the fault to reduce heat and thus reduce the recovery time. Again, at the end of fault when the SFCL needs to re-enter to the grid, turning off the trigger circuit in the two-bridge configuration the inrush current can be reduced. This is because the current only flows through the limiting coil. Another advantage of this type of SFCL is that no voltage sag will appear during load increasing time as long as the load current stays below the trigger current level.

  13. A simple approach for pre-LOCA analysis of MTR type research reactor

    International Nuclear Information System (INIS)

    In this study, it is intended to analyse early phases of a protected loss of coolant accident (LOCA) for TR-2 research reactor at Istanbul, and to show applicability of the present model to the other similar types of research reactors. Even though, there has been substantial amount of experimental and numerical works concerning LOCA of research reactor in the literature, most of the works has been done for the latest phase of accident where the core was totally uncovered and being cooled by natural circulation of air. It is our aim to investigate the transient situation since the time when coolant is beginning to be lost throughout one or more of the main coolant pipes which where supposed to be broken guillotine-like to the time when the core is totally uncovered. The modelling of the problem was separated into two phases: in the first phase when the water level of the pool being decreased in a pre-estimated time-dependent way calculated by using modified Bernoulli equation, the conservation equations are solved by a usual implicit finite difference algorithm. The later phase, when water level reaches to the top level of fuel plates and begins to decrease until the bottom of the core, needs some modifications to the approach used for the first phase. Because, the coolants channels among fuel plates are filled with air when the level goes below, and the fuel plates are being cooled by air above the water level. This complexity is resolved using a moving boundary approach in the numerical solution. A Lagrange type interpolation approximation for the derivatives along with interface conditions is the neighborhood of the air-water interface was imported to the numerical algorithm. For the meshes which are not close to the interface above mentioned usual finite difference scheme to solve conservation equations both for air and water side. The analyse is performed for a nominal channel and for a hot channel

  14. Proliferation Resistance and Material Type considerations within the Collaborative Project for a European Sodium Fast Reactor

    International Nuclear Information System (INIS)

    The collaborative project for a European Sodium Fast Reactor (CP‑ESFR) is an international project where 25 European partners developed Research & Development solutions and concepts for a European sodium fast reactor. The project was funded by the 7. European Union Framework Programme and covered topics such as the reactor architectures and components, the fuel, the fuel element and the fuel cycle, and the safety concepts. Within sub‑project 3, dedicated to safety, a task addressed proliferation resistance considerations. The Generation IV International Forum (GIF) Proliferation Resistance and Physical Protection (PR and PP) Evaluation Methodology has been selected as the general framework for this work, complemented by punctual aspects of the IAEA‑INPRO Proliferation Resistance methodology and other literature studies - in particular for material type characterization. The activity has been carried out taking the GIF PR and PP Evaluation Methodology and its Addendum as the general guideline for identifying potential nuclear material diversion targets. The targets proliferation attractiveness has been analyzed in terms of the suitability of the targets’ nuclear material as the basis for its use in nuclear explosives. To this aim the PR and PP Fissile Material Type measure was supplemented by other literature studies, whose related metrics have been applied to the nuclear material items present in the considered core alternatives. This paper will firstly summarize the main ESFR design aspects relevant for PR following the structure of the GIF PR and PP White Paper template. An analysis on proliferation targets is then discussed, with emphasis on their characterization from a nuclear material point of view. Finally, a high‑level ESFR PR analysis according to the four main proliferation strategies identified by the GIF PR and PP Evaluation Methodology (concealed diversion, concealed misuse, breakout, clandestine production in clandestine facilities) is

  15. The C. elegans CSR-1 Argonaute pathway counteracts epigenetic silencing to promote germline gene expression

    OpenAIRE

    Seth, Meetu; Shirayama, Masaki; Gu, Weifeng; Ishidate, Takao; Conte, Darryl; Mello, Craig C

    2013-01-01

    Organisms can develop adaptive sequence-specific immunity by re-expressing pathogen-specific small RNAs that guide gene silencing. For example, the C. elegans PIWI-Argonaute/piRNA pathway recruits RNA-dependent RNA polymerase RdRP to foreign sequences to amplify a trans-generational small RNA-induced epigenetic silencing signal (termed RNAe). Here we provide evidence that in addition to an adaptive memory of silenced sequences, C. elegans can also develop an opposing adaptive memory of expres...

  16. Irradiation tests on PHWR type fuel elements in TRIGA research reactor of INR Pitesti

    International Nuclear Information System (INIS)

    Nine PHWR type fuel elements with reduced length were irradiated in loop A of the TRIGA Research Reactor of INR Pitesti. The primary objective of the test was to determine the performance of nuclear fuel fabricated at INR Pitesti at high linear powers in pressurized water conditions. Six fuel elements were irradiated with a ramp power history, achieving a maximum power of 45 kW/m during pre-ramp and of 64 kW/m in the ramp. The maximum discharge burnup was of 216 MWh/kgU. Another three fuel elements with reduced length were irradiated with declining power history. At the beginning of irradiation the fuel elements achieved a maximum linear power of 66 kW/m. The maximum fuel power was about 1.3 times the maximum expected in PHWR. The maximum discharge burnup was 205 MWh/kgU. The elements were destructively examined in the hot cells of INR Pitesti. Temperature-sensitive parameters such as UO2 grain growth, fission-gas release and sheath deformations were examined. The tests proved the feasibility of irradiating PHWR type fuel elements at linear powers up to 66 kW/m under pressurized water conditions and demonstrated the possibility of more flexible operation of this fuel in power reactors. This paper presents the results of the investigation. (orig.)

  17. Optimizing the Design of Small Fast Spectrum Battery-Type Nuclear Reactors

    Directory of Open Access Journals (Sweden)

    Staffan Qvist

    2014-07-01

    Full Text Available This study is focused on defining and optimizing the design parameters of inherently safe “battery” type sodium-cooled metallic-fueled nuclear reactor cores that operate on a single stationary fuel loading at full power for 30 years. A total of 29 core designs were developed with varying power and flow conditions, including detailed thermal-hydraulic, structural-mechanical and neutronic analysis. Given set constraints for irradiation damage, primary cycle pressure drop and inherent safety considerations, the attainable power range and performance characteristics of the systems are defined. The optimum power level for a core with a coolant pressure drop limit of 100 kPa and an irradiation damage limit of 200 DPA (displacements per atom is found to be 100 MWt/40 MWe. Raising the power level of an optimized core gives significantly higher attainable power densities and burnup, but severely decreases safety margins and increases the irradiation damage. A fully optimized inherently safe battery-type fast reactor core with an active height and diameter of 150 cm (2.6 m3, a pressure drop limit of 100 kPa and an irradiation damage limit of 300 DPA can be designed to operate at 150 MWt/60 MWe for 30 years, reaching an average discharge burnup of 100 MWd/kg-actinide.

  18. NOx removal using a wet type plasma reactor based on a three-electrode device

    International Nuclear Information System (INIS)

    In this paper, a wet type plasma reactor based on a three electrode device is investigated experimentally in order to remove NO and NOx at low flow rate. First, a comparison of cleaning performances of gas exhaust has been performed when the surface discharge operates in DBD or SD modes. From these previous results, the second part of study has consisted to improve the electrochemical conversion of the wet type plasma reactor by adding a coil between the AC HV power supply and the surface discharge. The parametric study has been performed with 100 ppm of NO content in gas flow at room temperature and atmospheric pressure for a flow rate of 1 L/min. For each electrical parameter tested, an electric characterization and measurement of NOx content via FT-IR has been conducted. The results highlight a better cleaning of gas exhaust when the surface discharge operates in DBD mode. Moreover, the presence of solution promotes the arc transition when the operating mode is SD, resulting a reliability reduction of plasma device. In addition, the measurements show that the insertion of coil in the electrical circuit improves the NOx removal at a given power consumption for the DBD operating mode.

  19. Issues of intergranular embrittlement of VVER-type nuclear reactors pressure vessel materials

    Science.gov (United States)

    Zabusov, O.

    2016-04-01

    In light of worldwide tendency to extension of service life of operating nuclear power plants - VVER-type in the first place - recently a special attention is concentrated on phenomena taking place in reactor pressure vessel materials that are able to lead to increased level of mechanical characteristics degradation (resistibility to brittle fracture) during long term of operation. Formerly the hardening mechanism of degradation (increase in the yield strength under influence of irradiation) mainly had been taken into consideration to assess pressure vessel service life limitations, but when extending the service life up to 60 years and more the non-hardening mechanism (intergranular embrittlement of the steels) must be taken into account as well. In this connection NRC “Kurchatov Institute” has initiated a number of works on investigations of this mechanism contribution to the total embrittlement of reactor pressure vessel steels. The main results of these investigations are described in this article. Results of grain boundary phosphorus concentration measurements in specimens made of first generation of VVER-type pressure vessels materials as well as VVER-1000 surveillance specimens are presented. An assessment of non-hardening mechanism contribution to the total ductile-to- brittle transition temperature shift is given.

  20. Thermal hydraulics characterization of the core and the reactor vessel type BWR; Caracterizacion termohidraulica del nucleo y de la vasija de un reactor tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Zapata Y, M.; Lopez H, L.E. [CFE, Carretera Cardel-Nautla Km. 42.5, Municipio Alto Lucero, Veracruz (Mexico)]. e-mail: marxlenin.zapata@cfe.gob.mx

    2008-07-01

    The thermal hydraulics design of a reactor type BWR 5 as the employees in the nuclear power plant of Laguna Verde involves the coupling of at least six control volumes: Pumps jet region, Stratification region, Core region, Vapor dryer region, Humidity separator region and Reactor region. Except by the regions of the core and reactor, these control volumes only are used for design considerations and their importance as operative data source is limited. It is for that is fundamental to complement the thermal hydraulics relations to obtain major data that allow to determine the efficiency of internal components, such as pumps jet, humidity separator and vapor dryer. Like example of the previous thing, calculations are realized on the humidity of the principal vapor during starting, comparing it with the values at the moment incorporated in the data banks of the computers of process of both units. (Author)

  1. A new impulse in the development of nuclear pool-type reactors for underground heating plant: Designing, running background and possible perspectives

    International Nuclear Information System (INIS)

    This paper considers the concept of energy supply with using ultimately safe pool-type integral nuclear reactors. Safety and reliability of these reactors has already been demonstrated to the public by the long-term operation of this type various research reactors. The reactor and power plant design features, new approach to the nuclear safety, the nuclear upgrading of existing energy system in a small Russian town are considered in the paper

  2. Artificial intelligence applied to fuel management in BWR type reactors; Inteligencia artificial aplicada a la administracion de combustible en reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Ortiz S, J.J

    1998-10-01

    In this work two techniques of artificial intelligence, neural networks and genetic algorithms were applied to a practical problem of nuclear fuel management; the determination of the optimal fuel reload for a BWR type reactor. This is an important problem in the design of the operation cycle of the reactor. As a result of the application of these techniques, comparable or even better reloads proposals than those given by expert companies in the subject were obtained. Additionally, two other simpler problems in reactor physics were solved: the determination of the axial power profile and the prediction of the value of some variables of interest at the end of the operation cycle of the reactor. Neural networks and genetic algorithms have been applied to solve many problems of engineering because of their versatility but they have been rarely used in the area of fuel management. The results obtained in this thesis indicates the convenience of undertaking further work on this area and suggest the application of these techniques of artificial intelligence to the solution of other problems in nuclear reactor physics. (Author)

  3. In-pile creep behavior of type 316 stainless steel at Halden reactor

    Energy Technology Data Exchange (ETDEWEB)

    Takakura, K. [Japan Nuclear Energy Safety Organization (Japan); Sakima, K.; Fujimoto, K. [Takasago RD center, Mitsubishi Heavy Industries (Japan); Kubo, N. [Kobe Shipyard and Machinery Works, Mitsubishi Heavy Industries (Japan); Kido, T. [Nuclear Development Corporation (Japan)

    2011-07-01

    Irradiation Assisted Stress Corrosion Cracking (IASCC) is a matter of great concern as degradation of core internal components in light water nuclear reactor. Japan Nuclear Energy Safety organization (JNES) had been conducting a project related to IASCC as a part of safety research and development study for the aging management and maintenance of the nuclear power plant. Based on the JNES project results, JNES prepared 'IASCC evaluation guide for PWR baffle former bolts (BFBs)' including irradiation creep formula. The purpose of this paper is to describe the background of the creep formula. In order to assess probability of IASCC for stainless steels of BFBs in PWRs, reliable and precise stress evaluation formula for BFB is required, by considering stress change caused by irradiation creep. However, currently there are no in-pile creep data enough to calculate the stress change at PWR irradiation conditions, especially in highly stress region such as around yield stress. JNES and MHI had been conducting an in-pile creep test using the OECD Halden reactor. Type 316 stainless steel for BFB materials are irradiated to about 1 dpa. Tensile type (for lower stress region) and C-ring type specimens (for higher stress region) were used. The irradiation temperatures are 290 and 340 C. The following results were obtained. The steady-state creep rate obtained by tensile type specimens has been evaluated and it agrees well with the literature data, the stress exponent was approximately 1.5, and irradiation creep coefficient was approximately 1*10{sup -6}/dpa/MPa at lower stress region (under 500 MPa). At higher stress region (above 500 MPa), the stress exponent was about 4. There was a little effect of the irradiation temperature on creep rate. Finally, an in-pile creep equation consists of transient creep and steady-state creep covering higher stress region were obtained and included in the IASCC evaluation guide. (authors)

  4. Simplified model for the thermo-hydraulic simulation of the hot channel of a PWR type nuclear reactor

    International Nuclear Information System (INIS)

    The present work deals with the thermal-hydraulic analysis of the hot channel of a standard PWR type reactor utilizing a simplified mathematical model that considers constant the water mass flux during single-phase flow and reduction of the flow when the steam quality is increasing in the channel (two-phase flow). The model has been applied to the Angra-1 reactor and it has proved satisfactory when compared to other ones. (author). 25 refs, 15 figs, 3 tabs

  5. Development of core technology for research reactors using plate type fuels

    International Nuclear Information System (INIS)

    Around 250 research reactors are under operation over the world. However, about 2/3 have been operated more than 30 years and demands for replacements are expected in the near future. The number of expected units is around 110, and around 55 units from 40 countries will be expected to be bid in the world market. In 2007, Netherlands started international bidding process to construct a new 80MW RR (named PALLAS) with the target of commercial operation in 2016, which will replace the existing HFR(45MW). KAERI consortium has been participated in that bid. Most of RRs use plate type fuels as a fuel assembly, Be and Graphite as a reflector. On the other hand, in Korea, the KAERI is operating the HANARO, which uses a rod type fuel assembly and heavy water as a reflector. Hence, core technologies for RRs using plate type fuels are in short. Therefore, core technologies should be secured for exporting a RR. In chapter 2, the conceptual design of PALLAS which use plate type fuels are described including core, cooling system and connected systems, layout of general components. Experimental verification tests for the plate type fuel and second shutdown system and the code verification for nuclear design are explained in Chapter 3 and 4, respectively

  6. Development of in-vessel type control rod drive mechanism for a innovative small reactor (Contract research)

    Energy Technology Data Exchange (ETDEWEB)

    Yoritsune, Tsutomu; Ishida, Toshihisa [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    Although the control rod drive mechanism of an existing large scale light water reactor is generally installed outside the reactor vessel, an in-vessel type control rod drive mechanism (INV-CRDM) is installed inside the reactor vessel. The INV-CRDM contributes to compactness and simplicity of the reactor system, and it can eliminate the possibility of a rod ejection accident. Therefore, INV-CRDM is an important technology adopted in an innovative small reactor. Japan Atomic Energy Research Institute (JAERI) has developed this type of CRDM driven by an electric motor, which can work under high temperature and high pressure water for the advanced marine reactor. On the basis of this research result, a driving motor coil and a bearing were developed to be used under the high temperature steam, severe condition for an innovative small reactor. About the driving motor, we manufactured the driving motor available for high temperature steam and carried out performance test under room temperature atmosphere to confirm the electric characteristic and coolability of the driving coil. With these test results and the past test results under high temperature water, we analyzed and evaluated the electric performance and coolability of the driving coil under high temperature steam. Concerning bearing, we manufactured the test pieces using some candidate material for material characteristic test and carried out the rolling wear test under high temperature steam to select the material. Consequently, we confirmed that performance of the driving coil for the advanced type driving motor, is enough to be used under high temperature steam. And, we evaluated the performance of the bearing and selected the material of the bearing, which can be used under high temperature steam. From these results, we have obtained the prospect that the INV-CRDM can be used for an innovative small reactor under steam atmosphere could be developed. (author)

  7. [Continuous operation of hydrogen bio-production reactor with ethanol-type fermentation].

    Science.gov (United States)

    Ren, Nan-qi; Gong, Man-li; Xing, De-feng

    2004-11-01

    The natural response of a continuous stirred tank reactor (CSTR) for hydrogen bio-production using molasses wastewater as substrate was investigated. Emphasis was placed on assessing the operational controlling strategy on the stable operation of CSTR with high efficiency. It was found that at an initial biomass of 15g/L, an equilibrial microbial community in the ethanol-type fermentation and efficient stable operation of CSTR could be established with following conditions: temperature of 35 degrees C +/- 1 degrees C, COD organic loading rate (OLR) of 40kg/(m3 x d), hydraulic retention time (HRT) of 4h, pH value of 4.6 - 4.9 and oxidation reduction potential (ORP) of -450 - -470mV. Following that, hydrogen production in the reactor was relatively stable. The observed maximal hydrogen bio-production rate was 7.63m3/(m3 x d). The content of hydrogen in the biogas was about 40% - 58%. COD removal rate was between 22% - 26%. The total content of ethanol and acetic acid in the fermentative end products was above 80%.

  8. Base isolation technique for tokamak type fusion reactor using adaptive control

    International Nuclear Information System (INIS)

    In this paper relating to the isolation device of heavy structure such as nuclear fusion reactor, a control rule for reducing the response acceleration and relative displacement simultaneously was formulated, and the aseismic performance was improved by employing the adaptive control method of changing the damping factors of the system adaptively every moment. The control rule was studied by computer simulation, and the aseismic effect was evaluated in an experiment employing a scale model. As a results, the following conclusions were obtained. (1) By employing the control rule presented in this paper, both absolute acceleration and relative displacement can be reduced simultaneously without making the system unstable. (2) By introducing this control rule in a scale model assuming the Tokamak type fusion reactor, the response acceleration can be suppressed down to 78 % and also the relative displacement to 79 % as compared with the conventional aseismic method. (3) The sensitivities of absolute acceleration and relative displacement with respect to the control gain are not equal. However, by employing the relative weighting factor between the absolute acceleration and relative displacement, it is possible to increase the control capability for any kind of objective structures and appliances. (author)

  9. Properties of bio-oil generated by a pyrolysis of forest cedar residuals with the movable Auger-type reactor

    Science.gov (United States)

    Nishimura, Shun; Miyazato, Akio; Ebitani, Kohki

    2016-02-01

    Our research project has developed the new movable reactor for bio-oil production in 2013 on the basis of Auger-type system. This package would be a great impact due to the concept of local production for local consumption in the hilly and mountainous area in not only Japan but also in the world. Herein, we would like to report the properties of the bio-oil generated by the developing Auger-type movable reactor. The synthesized bio-oil possessed C: 46.2 wt%, H: 6.5 wt%, N: wt%, S: alcohol or carboxylic acid functional groups. Thus, it is suggested that the bio-oil generated by the new movable Auger-type reactor has a significant potential as well as the existing bio-oil reported previously.

  10. Propagation of cracks by stress corrosion in conditions of BWR type reactor; Propagacion de grietas por corrosion bajo esfuerzo en condiciones de reactor de agua en ebullicion (BWR)

    Energy Technology Data Exchange (ETDEWEB)

    Merino C, F.J. [ININ, 52045 Estado de Mexico (Mexico); Fuentes C, P. [ITT, Metepec, Estado de Mexico (Mexico)]. E-mail: fjmc@nuclear.inin.mx

    2004-07-01

    In this work, the obtained results when applying the Hydrogen Chemistry to a test tube type Compact Tension (CT), built in austenitic stainless steel 304l, simulating the conditions to those that it operates a Boiling Water Reactor (BWR), temperature 288 C and pressure of 8 MPa are presented. With the application of this water chemistry, seeks to be proven the diminution of the crack propagation speed. (Author)

  11. E-SCAPE: A scale facility for liquid-metal, pool-type reactor thermal hydraulic investigations

    Energy Technology Data Exchange (ETDEWEB)

    Van Tichelen, Katrien, E-mail: kvtichel@sckcen.be [SCK-CEN, Boeretang 200, 2400 Mol (Belgium); Mirelli, Fabio, E-mail: fmirelli@sckcen.be [SCK-CEN, Boeretang 200, 2400 Mol (Belgium); Greco, Matteo, E-mail: mgreco@sckcen.be [SCK-CEN, Boeretang 200, 2400 Mol (Belgium); Viviani, Giorgia, E-mail: giorgiaviviani@gmail.com [University of Pisa, Lungarno Pacinotti 43, 56126 Pisa (Italy)

    2015-08-15

    Highlights: • The E-SCAPE facility is a thermal hydraulic scale model of the MYRRHA fast reactor. • The focus is on mixing and stratification in liquid-metal pool-type reactors. • Forced convection, natural convection and the transition are investigated. • Extensive instrumentation allows validation of computational models. • System thermal hydraulic and CFD models have been used for facility design. - Abstract: MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) is a flexible fast-spectrum research reactor under design at SCK·CEN. MYRRHA is a pool-type reactor with lead bismuth eutectic (LBE) as primary coolant. The proper understanding of the thermal hydraulic phenomena occurring in the reactor pool is an important issue in the design and licensing of the MYRRHA system and liquid-metal cooled reactors by extension. Model experiments are necessary for understanding the physics, for validating experimental tools and to qualify the design for the licensing. The E-SCAPE (European SCAled Pool Experiment) facility at SCK·CEN is a thermal hydraulic 1/6-scale model of the MYRRHA reactor, with an electrical core simulator, cooled by LBE. It provides experimental feedback to the designers on the forced and natural circulation flow patterns. Moreover, it enables to validate the computational methods for their use with LBE. The paper will elaborate on the design of the E-SCAPE facility and its main parameters. Also the experimental matrix and the pre-test analysis using computational fluid dynamics (CFD) and system thermal hydraulics codes will be described.

  12. Modern design and safety analysis of the University of Florida Training Reactor

    International Nuclear Information System (INIS)

    Highlights: • A new safety analysis of the University of Florida Training Reactor is presented. • This analysis uses modern codes and replaces the NRC approved analysis from 1982. • Reduction in engineering margin confirms that the UFTR is a negligible risk reactor. • Safety systems are not required to ensure that safety limits are not breached. • Negligible risk reactors are ideal for testing digital I&C equipment. - Abstract: A comprehensive series of neutronics and thermal hydraulics analyses were conducted to demonstrate the University of Florida Training Reactor (UFTR), an ARGONAUT type research reactor, as a negligible risk reactor that does not require safety-related systems or components to prevent breach of a safety limit. These analyses show that there is no credible UFTR accident that would result in major fuel damage or risk to public health and safety. The analysis was based on two limiting scenarios, whose extremity bound all other accidents of consequence: (1) the large step insertion of positive reactivity and (2) the release of fission products due to mechanical damage to a spent fuel plate. The maximum step insertion of positive reactivity was modeled using PARET/ANL software and shows a maximum peak fuel temperature of 283.2 °C, which is significantly below the failure limit of 530 °C. The exposure to the staff and general public was calculated for the worst-case fission product release scenario using the ORIGEN-S and COMPLY codes and was shown to be 6.5% of the annual limit. Impacts on reactor operations and an Instrumentation & Control System (I&C) upgrade are discussed

  13. High Purity and Yield of Boron Nitride Nanotubes Using Amorphous Boron and a Nozzle-Type Reactor

    Directory of Open Access Journals (Sweden)

    Jaewoo Kim

    2014-08-01

    Full Text Available Enhancement of the production yield of boron nitride nanotubes (BNNTs with high purity was achieved using an amorphous boron-based precursor and a nozzle-type reactor. Use of a mixture of amorphous boron and Fe decreases the milling time for the preparation of the precursor for BNNTs synthesis, as well as the Fe impurity contained in the B/Fe interdiffused precursor nanoparticles by using a simple purification process. We also explored a nozzle-type reactor that increased the production yield of BNNTs compared to a conventional flow-through reactor. By using a nozzle-type reactor with amorphous boron-based precursor, the weight of the BNNTs sample after annealing was increased as much as 2.5-times with much less impurities compared to the case for the flow-through reactor with the crystalline boron-based precursor. Under the same experimental conditions, the yield and quantity of BNNTs were estimated as much as ~70% and ~1.15 g/batch for the former, while they are ~54% and 0.78 g/batch for the latter.

  14. Contribution to fuel depletion study in PWR type reactors, reactor core with three and four regions of enrichment

    International Nuclear Information System (INIS)

    The main methods for calculation of fuel depletion are studied and some approaches to do it are mentioned; the LEOPARD Code is described and full details are given for each subroutine, flow charts are included; the method given by the code for calculation of fuel depletion is described; some imperfections from the IPR's version are listed, and corrected, for instance: the method for burn-up calculation of heavy isotopes; the results of calculations for a reference reactor based on data of the Preliminary Safety Analysis Report (PSAR) for Angra I Nuclear Power Plant are presented and discussed. (author)

  15. Development of poison injection code-COPJET for high pressure liquid poison injection in pressure tube type heavy water reactor

    International Nuclear Information System (INIS)

    Shut Down System-2 (SDS-2) in advanced vertical pressure tube type reactor, provides rapid reactor shutdown by high pressure injection of a neutron absorbing liquid called poison, into the moderator in the calandria. Poison inside the calandria is distributed by poison jets issued from holes provided in the injection tubes. Effectiveness of the system depends on the rate and spread of the poison in the moderator. In this study, a transient one-dimensional (1-D) hydraulic code, COPJET is developed, to predict the performance of system by predicting poison jet length with time. Validation of the COPJET is done with the data available in literature. Thereafter, it is applied for poison jet length prediction of advanced vertical pressure type reactor. (author)

  16. Aging of reactor vessels in LWR type reactors; Envejecimiento de la vasija y de los internos del nuclear de los reactores tipo LWR

    Energy Technology Data Exchange (ETDEWEB)

    Gomez Briceno, D.; Lapena, J.; Serrano, M.

    2004-07-01

    Most of the degradation mechanisms of nuclear components were not included on the design so they have to be treated a posteriori, and that imply a loss of capacity. In this paper the state of the art on the reactor pressure vessel neutron embrittlement and on the irradiation assisted stress corrosion cracking that affects internal components, are explained. Special attention is devoted on the influence of the neutron fluence on IASCC process, on the material alterations promoted by irradiation and their consequences on the susceptibility to this phenomenon. Regarding the reactor pressure vessel degradation, this paper discuss the application of the Master Curve on the structural integrity evaluation of the vessel. Other aspects related to further developments are also mentioned and the importance of a good materials ageing management on the operation of the plant is pointed out. (Author) 12 refs.

  17. Design of a decay tank for a pool type research reactor with a CFD model

    International Nuclear Information System (INIS)

    A conceptual primary cooling system (PCS) was designed for adequate cooling of the core of a research reactor. The primary coolant after passing through the reactor core contains many kinds of radio-nuclides. A decay tank provides a delayed transit time to ensure that the N-16 activity decreases enough before the coolant leaves the decay tank's shielding room. The size of the decay tank should be enlarged to provide sufficient transit time. However, there was a limitation: to minimize the tank size, it should be designed with an internal baffle, which affects the pressure loss in the system and net positive suction head (NPSH) of the PCS pump. Therefore, the decay tank should be optimized for size and the internal baffle. A vertical type decay tank was chosen to optimize the geometrical arrangement of PCS and the vertical internal baffle was installed to minimize the number of internal structures. The preliminary geometry of the tank and the internal baffle were determined to satisfy the required delayed transit time by calculating the maximum velocity and the flow path length of the circular and the annular sections of the tank. The commercially available CFD model, FLUENT, which solves the Navier-Stokes and turbulent models, was used to specifically design the decay tank with the preliminarily calculated geometry and the related flow rate. Several turbulence models, standard k-ε model, renormalization group (RNG) model, and realizable k-ε model, were conducted to isolate the root cause of these differences. By comparing the results of the velocity profile and the characteristics of each model, a detailed design study was simulated using the realizable k-ε model. A user-defined scalar equation was solved to estimate the delayed transit time. The size and the internal baffle that satisfy the required transit time were determined based on the CFD results. (author)

  18. Dynamic structural analysis concerning integrity assessment of a reactor cavity ceiling of type VVER-1000 due to postulated failure of the reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Eisert, P.; Bachmann, P.; Sievers, J. [Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH (GRS) (Germany)

    2007-07-01

    In the framework of the activities concerning the safety of nuclear power plants (NPP) in Middle- and East Europe the behaviour of the reactor cavity bottom ceiling of a NPP of type WER-1000 stressed by loads of postulated failure of the lower head of the reactor pressure vessel (RPV-LH) caused by an assumed core melt accident has been investigated. The investigations are performed in the framework of probabilistic safety analyses (PSA) and include the effects of small leaks in the RPV-LH caused by molten material as well as the total separation of the RPV-LH. The corresponding thermal and mechanical loads are based on results of thermal hydraulic investigations. (orig.)

  19. Design of a plate type fuel based - low power medical reactor for boron neutron capture therapy

    International Nuclear Information System (INIS)

    The interest in the boron neutron capture therapy (BNCT) has been renewed for cancer therapy with some indication of its potential efficacy in recent years. To solve the most important problem that thermal neutrons are attenuated rapidly in tissue due to absorption and scattering, thermal neutron beams are replaced by epithermal neutron beams. Thus, epithermal neutron beams are directed towards a patient's head, during their passage through tissue these neutrons rapidly lose energy by elastic scattering until they end up as thermal neutrons in target tumor volume. The thermal neutrons thus formed, are captured by the 10B atoms which become 11B atoms in the excited state for a very short time 10-12 sec. The 11B atoms then decay producing alpha particles, 7Li recoil nuclei and gamma rays. Tumor cells are killed selectively by the energetic alpha particles and 7Li fission products. We propose a 300kW slab type reactor core having thin and large surface areas so that most of the neutrons emerging from the faces and entering moderator region are fission spectrum neutrons to acquire high intense epithermal neutron beam with high quality. All faces of the slab core, East-West region and North-South region, were considered for epithermal neutron beam collimators. Plate-type U3Si2-Al dispersion fuel having high uranium density is very compatible with composing of a slab type core. The reactor core is loaded with 3.89kg U235 and has the dimension of about 23.46cm width, 31.28cm length and 64.8cm height, with 216 locations to place 204 fuel elements, eight control plates and four safety plates. The general-purpose MCNP 4B code was used to carry out the neutron and photon transport computations. Both keff criticality and fixed source problems were computed. We could reduce at least 7 times long computer time (105 to 140 h in a run) needed to initiate enough neutrons in a run ( 6000 to 8000 cycles in a run with 3000 neutrons per cycle) using the PVM (Parallel Virtual

  20. Research on Precaution and Detection Technology for Flow Blockage of Plate-type Fuel Element in Research Reactors

    Institute of Scientific and Technical Information of China (English)

    DING; Li; QIAO; Ya-xin; ZHANG; Nian-peng; LUO; Bei-bei; HUA; Xiao; JIA; Shu-jie; YAN; Hui-yang

    2013-01-01

    The main aim of this study is to offer the technical support for safety operation and management of research reactors using plate-type fuel assemblies in China,which is performed from analysis of precaution measures for flow blockage and detection methods of accidents.Study shows that most accidents were induced by in-core foreign objects and the swelling of fuel

  1. Temperature coefficient of reactivity of a typical swimming pool type research reactor using low enriched uranium fuel

    International Nuclear Information System (INIS)

    The temperature coefficients of reactivity of a swimming pool type material test research reactor have been calculated using standard computer codes. It is observed that the core reactivity loss due to increase in water temperature and void formation is sensitive to control rod position at criticality. The reactivity decreases more rapidly when the core volume is small. (author)

  2. Radiotoxicity and decay heat power of spent nuclear fuel of VVER type reactors at long-term storage.

    Science.gov (United States)

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

    2005-01-01

    Radiotoxicity and decay heat power of the spent nuclear fuel of VVER-1000 type reactors are calculated during storage time up to 300,000 y. Decay heat power of radioactive waste (radwaste) determines parameters of the heat removal system for the safe storage of spent nuclear fuel. Radiotoxicity determines the radiological hazard of radwaste after its leakage and penetration into the environment.

  3. Study of corrosion resistance of materials for steam generator pipes at NPP with a WWER-type reactor

    International Nuclear Information System (INIS)

    Results of experimental investigation of corrosion processes of different steels and alloys are presented under conditions meeting the regime of their exploitation in the first and second contours of NPP with WWER-type reactor. Recommendations on estimates of the value of corrosion losses of studied materials and value of possible formation of corrosion products are given

  4. A viral suppressor protein inhibits host RNA silencing by hooking up with Argonautes

    KAUST Repository

    Jin, Hailing

    2010-05-01

    RNA viruses are particularly vulnerable to RNAi-based defenses in the host, and thus have evolved specific proteins, known as viral suppressors of RNA silencing (VSRs), as a counterdefense. In this issue of Genes & Development, Azevedo and colleagues (pp. 904-915) discovered that P38, the VSR of Turnip crinkle virus, uses its glycine/tryptophane (GW) motifs as an ARGONAUTE (AGO) hook to attract and disarm the host\\'s essential effector of RNA silencing. Several GW motif-containing cellular proteins are known to be important partners of AGOs in RNA silencing effector complexes in yeast, plants, and animals. The GW motif appears to be a versatile and effective tool for regulating the activities of RNA silencing pathways, and the use of GW mimicry to compete for and inhibit host AGOs may be a strategy used by many pathogens to counteract host RNAi-based defenses. © 2010 by Cold Spring Harbor Laboratory Press.

  5. Completely Autotrophic Nitrogen-removal over Nitrite in Two Types of Reactors

    Institute of Scientific and Technical Information of China (English)

    2005-01-01

    Two lab-scale reactors, susponded-sludge and fluidized-bed, were conducted with the feed of ammonium-rich syntheticwa,tewater devoid of COD. Completely autotrophic nitrogen-removal process was fulfilled in both reactors and the maximum efficiencies of nitrogen removal were achieved, 65% in the suspended-sludge reactor and 73% in the fluidized-bed reactor respectively. Different fromn the steady performance of the fluidized-bed reactor, the suspended-sludge reactor came to deteriorate constantly after a period of stable operation, resulting in almost complete loss of the N-removal ability in the suspending system.Molecular methods such as PCR and FISH were employed for describing the microbial characteristics in two systems. This study suggests that a biofilm system is a suitable configuration for completely autotrophic N-removal with more feasibility and stability than a suspending system.

  6. Inherent safety of advanced nuclear engineering based on BN-800 - type fast reactors

    International Nuclear Information System (INIS)

    Considerations based on the prolonged experience of fast reactor operations exhibiting outlook application of reactors on a basis of BN-800 with sodium coolant are given. Reliability and safety of the block are supported by the probability analysis of safety in the content of engineering project. Conversion on the reactor core with nitride fuel will significantly raise a possibility to conform to safety and nonproliferation of fission materials needs. The suggested optimum variant for reactor core on a basis of nitride fuel is advanced

  7. Commercial scale performance predictions for high-temperature electrolysis plants coupled to three advanced reactor types

    International Nuclear Information System (INIS)

    This paper presents results of system analyses that have been developed to assess the hydrogen-production performance of commercial-scale high-temperature electrolysis (HTE) plants driven by three different advanced reactor - power-cycle combinations: a high-temperature helium-cooled reactor coupled to a direct Brayton power cycle, a supercritical CO2-cooled reactor coupled to a direct recompression cycle, and a sodium-cooled fast reactor coupled to a Rankine cycle. The system analyses were performed using UniSim software. The work described in this report represents a refinement of previous analyses in that the process flow diagrams include realistic representations of the three advanced reactors directly coupled to the power cycles and integrated with the high-temperature electrolysis process loops. In addition, this report includes parametric studies in which the performance of each HTE concept is determined over a wide range of operating conditions. Results of the study indicate that overall thermal-to-hydrogen production efficiencies (based on the low heating value of the produced hydrogen) in the 45 - 50% range can be achieved at reasonable hydrogen production rates with the high-temperature helium-cooled reactor concept, 42 - 44% with the supercritical CO2-cooled reactor and about 33 - 34% with the sodium-cooled reactor. (authors)

  8. Core monitoring and surveillance of VVER-440 type reactors in the Czech Republic and Slovak Republic

    International Nuclear Information System (INIS)

    The SCORPIO-VVER reactor core monitoring system is an advanced redundant software system without actuating members falling in the BT3 class which has been installed at the four Dukovany reactor units and at two units of the Slovak Jaslovske Bohunice V2 NPP. The system is described in detail and its history and experience gained at Dukovany are highlighted. (orig.)

  9. Parametric evaluation of mixed (low and high enriched) fuel core for a swimming pool type research reactor

    International Nuclear Information System (INIS)

    A study has been carried out to evaluate the performance of a swimming pool type research reactor core comprised of mixed (low and high enriched) uranium fuel. The study includes the calculations of core reactivity, worth of control rods and core criticality at the Beginning Of Life (BOL) of the core and for two operating conditions Cold Zero Power (CZP) and Hot Full Power (HFP). Further, to ensure safe and stable operation of the core from nuclear design point of view, average power densities in the fuel region, power peaking factors, axial power distribution in the hot channel and reactivity feed back coefficients have also been calculated. Two group fluxes have also been determined at different irradiation locations. All these calculations have been carried out employing reactor lattice code WIMS-D14 and reactor analysis code CITATION The calculated results show reasonably good agreement with the quoted operational data of the previous LEU cores. (Author)

  10. Construction of real-type simulator reusing the equipment of the Musashi-reactor. The 2nd report

    International Nuclear Information System (INIS)

    Real-time reactor simulator had been developed with reusing control rod drive and operation console, and simulated fuel elements and grid plate of the Musashi reactor. Type of the fuel element and its location in the core were identified through electric circuits. Core characteristics such as excess reactivity, control rod worths and temperature effects were reproduced on a personal computer using the actual operation data of the Musashi Reactor. Operation of control rod, core characteristics, core configuration and instrumentation data were mutually linked and controlled by the interface. Effect of delayed neutrons and simulated reactivity insertion accident could be demonstrated with the application software installed. The simulator was incorporated in the curriculum. (T. Tanaka)

  11. Conceptual design of loop-in-tank type Indian molten salt breeder reactor concept

    International Nuclear Information System (INIS)

    The third stage of Indian nuclear power programme envisages use of thorium as fertile material with 233U, which is proposed to be obtained from reprocessing of spent fuel of Pu/Th based fast reactors in the later part of the second stage of the programme. In India, thorium based reactors have been designed in many configurations, from light water cooled designs to high temperature liquid metal and molten salt cooled options. Another option, which holds promise, is the molten salt-fuelled reactor, which can be configured to give significant breeding ratios. A crucial part for achieving reasonable breeding in such reactors is the need to reprocess the salt continuously, either online or in batch mode. India has recently started carrying out fundamental studies so as to arrive at a conceptual design of Indian Molten Salt Breeder Reactor (IMSBR). (author)

  12. VEERA facility for studies of nuclear safety in VVER type reactors

    International Nuclear Information System (INIS)

    The VEERA facility was built in 1987 for experiments that simulate soluble neutron poison (boric acid) behaviour in a pressurized water reactor (PWR) during the long-term cooling period of loss-of-coolant accidents (LOCAs). The experiments provided insight especially into the processes of concentration, mixing and possible crystallization of boric acid in the core region of a PWR. In 1993 the facility was modified in order to use it for studies of the reflooding phenomenon. The results of the reflood experiments will be used as a data base for testing the capability of the reflood models of different computer codes. The VEERA facility in its original and modified forms is described in this report. Details of the geometry and dimensions of the components are given. This data is needed as a geometrical boundary condition in input deck preparation for thermal hydraulic analysis. The instrumentation and the data acquisition system are described so that the applicability of the facility and the accuracy of the measurements for different types of experiments can be evaluated. Initial and boundary conditions of the experiments and the principal test procedures are also summarized. (orig.) (24 figs., 6 tabs.)

  13. Unlimited cooling capacity of the passive-type emergency core cooling system of the MARS reactor

    International Nuclear Information System (INIS)

    The MARS nuclear plant is a 600 MWth PWR with completely passive core safeguards. The most relevant innovative safety system is the Emergency Core Cooling System (ECCS), which is based on natural circulation, and on a passive-type activation that follows a core flow decrease, whatever was the cause (only one component, 400% redundant, is not static). The main thermal hydraulic transients occurring as a consequence of design basis accidents for the MARS plant were presented at the ICONE 3 Conference. Those transients were analyzed in the first stage, with the aim at pointing out the capability of the innovative ECCS to intervene. So, they included only a short-time analysis (extended for a few hundreds of seconds) and the well known RELAP 5 computer program was used for this purpose. In the present paper, the long-term analyses (extended for several thousands of seconds) of the same transients are shown. These analyses confirmed that the performance of the Emergency Core Cooling System of the MARS reactor is guaranteed also in long-term scenarios

  14. Criticality calculations for a spent fuel storage pool for a BWR type reactor

    International Nuclear Information System (INIS)

    In this work, the methodology for the calculation of the constant of effective multiplication for the arrangement of spent fuel assemblies in the pool of a BWR type reactor is shown. Calculations were done for the pool of spent fuel specified in FSAR and for the assemblies that is thought a conservative composition of high enrichment and without Gadolinium, giving credit to the stainless steel boxes of the frames that keep the assemblies. To carry out this simulation, RECORD and MIXQUIC codes were used. With record code, macroscopic cross sections, two energy groups, for the characteristics of the thought assemblies were obtained. Cross sections, as well as the dimensions of the frames that keep the fuel assemblies were used as input data for MIXQUIC code. With this code, criticality calculations in two dimensions were done, supposing that there is not leak of neutrons along the axial of the main line. Additional calculations, supposing changes in the temperature, distance among fuel assemblies and the thickness of the stainless steel box of the frame were done. The obtained results, including the effect in tolerances due to temperature, weight and thickness, show that the arrangement in the pool, when frames are fully charged, is subcritical by less than 5% in δK. (Author)

  15. Development of the radiation models of a BWR type reactor and it facility in the SUN-RAH

    International Nuclear Information System (INIS)

    This work about generation models, transport in processes and radioactive contamination of areas of a BWR central, is an amplification to the project developed in the UNAM to have a support tool in subjects or electric generation courses. It is planned about the implementation of models of radiation generation in a BWR type reactor for complement the functions developed in the University Simulator of Nucleo electric- Boiling water reactor (SUN-RAH) which it has been implemented in Simulink of MatLab and it has a model for the dynamics of one nucleo electric central that presents the main characteristics of the reactor vessel, the recirculation system, steam lines, turbines, generator, condensers and feeding water, defined by the main processes that intervene in the generation of energy of these plants. By this way the radiation monitoring systems for area and process, operate simultaneously with the processes of energy generation, with that is possible to observe the changes that present with respect to the operation conditions of the plant, and likewise to appreciate the radiation transport process through the components of the reactor, steam lines and turbines, for different operation conditions and possible faults that they could be presented during the reactor operation. (Author)

  16. Effect of conditions of air-lift type reactor work on cadmium adsorption

    Energy Technology Data Exchange (ETDEWEB)

    Filipkowska, Urszula; Szymczyk, Paula Szymczyk; Kuczajowska-Zadrozna, Malgorzata; Joezwiak, Tomasz [University of Warmia and Mazury in Olsztyn, Warszawska (Poland)

    2015-10-15

    We investigated cadmium sorption by activated sludge immobilized in 1.5% sodium alginate with 0.5% polyvinyl alcohol. Experiments were conducted in an air-lift type reactor at the constant concentration of biosorbent reaching 5 d.m./dm{sup 3}, at three flow rates: 0.1, 0.25 and 0.5 V/h, and at three concentrations of the inflowing cadmium solution: 10, 25 and 50mg/dm{sup 3}. Analyses determined adsorption capacity of activated sludge immobilized in alginate as well as reactor's work time depending on flow rate and initial concentration of the solution. Results achieved were described with the use of Thomas model. The highest adsorption capacity of the sorbent (determined from the Thomas model), i.e., 200.2mg/g d.m. was obtained at inflowing solution concentration of 50mg/dm{sup 3} and flow rate of 0.1V/h, whereas the lowest one reached 53.69mg/g d.m. at the respective values of 10mg/dm{sup 3} and 0.1 V/h. Analyses were also carried out to determine the degree of biosorbent adsorption capacity utilization at the assumed effectiveness of cadmium removal - at the breakthrough point (C=0.05*C{sub 0}) and at adsorption capacity depletion point (C−0.9*C0). The study demonstrated that the effectiveness of adsorption capacity utilization was influenced by both the concentration and flow rate of the inflowing solution. The highest degree of sorbent capacity utilization was noted at inflowing solution concentration of 50mg/dm{sup 3} and flow rate of 0.1 V/h, whereas the lowest one at the respective values of 10mg/dm{sup 3} and 0.1 V/h. The course of the process under dynamic conditions was evaluated using coefficients of tangent inclination - a, at point C/C{sub 0}=1/2. A distinct tendency was demonstrated in changes of tangent slope a as affected by the initial concentration of cadmium and flow rate of the solution. The highest values of a coefficient were achieved at the flow rate of 0.1 V/h and initial cadmium concentration of 50mg/dm{sup 3}.

  17. A Pelargonium ARGONAUTE4 gene shows organ-specific expression and differences in RNA and protein levels.

    Science.gov (United States)

    He, Jie; Gray, John; Leisner, Scott

    2010-03-01

    RNAi-induced gene silencing plays a role in plant DNA methylation and defense. While most gene silencing studies have been performed on annuals, little is known about the expression of key components of this process (like ARGONAUTE proteins) in ornamentals. Using a combination of polymerase chain reaction techniques, an ARGONAUTE4 gene, PhAGO4, was isolated from Pelargonium. PhAGO4 encodes a predicted product of 934 amino acids that contains the PAZ and PIWI domains typical of ARGONAUTE (AGO) proteins. Phylogenetic analyses indicate that PhAGO4 clusters with other plant AGO4 proteins. Organ expression patterns of the AGO4 genes in Pelargonium and Arabidopsis show intriguing differences. AGO4 RNA levels decline with leaf age in both Arabidopsis and Pelargonium. In contrast AGO4 RNA levels in roots relative to leaves are higher in Pelargonium than in Arabidopsis. Both Arabidopsis and Pelargonium AGO4 showed higher RNA levels in flowers than leaves or roots. Even though flowers show higher levels of PhAGO4 RNA when compared to leaves and roots, protein gel blot analysis shows that at the protein level, the reverse is true. This suggests that PhAGO4 expression may be regulated at the translational or post-translational level in Pelargonium flowers.

  18. Numerical Analysis of Magnetic Force of Dry-Type Air-Core Reactor

    Institute of Scientific and Technical Information of China (English)

    LIUZhi-gang; GENGYing-san; WANGJian-hua

    2004-01-01

    This paper presents a coupled magnetic-circuit method for computing the magnetic force of air-core reactor under short-time current. The current and the magnetic flux density are computed first and then the magnetic force is obtained. Thus, the dynamic stability performance of air-core reactor can be analyzed at the design stage to reduce experimental cost and shorten the lead-time of product development.

  19. Scaled Facility Design Approach for Pool-Type Lead-Bismuth Eutectic Cooled Small Modular Reactor Utilizing Natural Circulation

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sangrok; Shin, Yong-Hoon; Lee, Jueun; Hwang, Il Soon [Seoul National University, Seoul (Korea, Republic of)

    2015-10-15

    In low carbon era, nuclear energy is the most prominent energy source of electricity. For steady ecofriendly nuclear energy supply, Generation IV reactors which are future nuclear reactor require safety, sustainability, economics and non-proliferation as four criteria. Lead cooled fast reactor (LFR) is one of these reactor type and Generation IV international forum (GIF) adapted three reference LFR systems which are a small and movable systems with long life without refueling, intermediate size and huge electricity generation system for power grid. NUTRECK (Nuclear Transmutation Energy Center of Korea) has been designed reactor called URANUS (Ubiquitous, Rugged, Accident-forgiving, Non-proliferating, and Ultra-lasting Sustainer) which is small modular reactor and using lead-bismuth eutectic coolant. To prove natural circulation capability of URANUS and analyze design based accidents, scaling mock-up experiment facility will be constructed. In this paper, simple specifications of URANUS will be presented. Then based on this feature, scaling law and scaled facility design results are presented. To validate safety feature and thermodynamics characteristic of URANUS, scaled mockup facility of URANUS is designed based on the scaling law. This mockup adapts two area scale factors, core and lower parts of mock-up are scaled for 3D flow experiment. Upper parts are scaled different size to reduce electricity power and LBE tonnage. This hybrid scaling method could distort some thermal-hydraulic parameters, however, key parameters for experiment will be matched for up-scaling. Detailed design of mock-up will be determined through iteration for design optimization.

  20. Core design of NPP Pebble Bed Modular Reactor (PBMR) type using computer code MCNP-5 for beginning of life (BOL)

    International Nuclear Information System (INIS)

    The core design of Nuclear Power Plant for Pebble Bed Modular Reactor (PBMR) type with 70 MWe capacity power in Beginning of Life (BOL) has been performed. The aim of this analysis, to know percent enrichment, temperature distribution and safety value by negative temperature coefficient at type PBMR if reactor power become lower equal to 70 MWe. This analysis was expected become one part of overview project development the power plant with 10.000 MWe of total capacity, spread evenly in territory of Indonesia especially to support of smelter industries. The results showed that, effective multiplication factor (keff) with power 70 MWe critical condition at enrichment 5,626 % is 1,00031±0, 00087, based on enrichment result, a value of the temperature coefficient reactivity is 10,0006 pcm/K. Based on the results of these studies, it can be concluded that the PBMR 70 MWe design is theoretically safe. (author)

  1. Radiological impact of plutonium recycle in the fuel cycle of LWR type reactors: professional exposure during mormal operation

    International Nuclear Information System (INIS)

    The radiological impact of the fuel cycle of light water type reactors using enriched uranium may be changed by plutonium recycle. The impact on human population and on the persons professionally exposed may be different according to the different steps of the fuel cycle. This report analyses the differential radiological impact on the different types of personnel involed in the fuel cycle. Each step of the fuel cycle is separately studied (fuel fabrication, reactor operation, fuel reprocessing), as also the transport of the radioactive materials between the different steps. For the whole fuel cycle, one estimates that, with regard to the fuel cycle using enriched uranium, the plutonium recycle involves a small increase of the professional exposure

  2. Control of fermentation types in continuous-flow acidogenic reactors: effects of pH and redox potential

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    The experiments were carried out in continuous-flow acidogenic reactors with molasses used as sub strate to study the effects of pH and redox potential on fermentation types. The conditions for each fermentation type were investigated at different experimental stages of start-up, pH-regulating and redox potential-regulating.The experiments confirmed that butyric acid-type fermentation would occur at pH > 6, the propionic acid-type fermentation at pH about 5.5 with Eh > - 278 mV, and the ethanol-type fermentation at pH < 4.5. A higher redox potential will lead to propionic acid-type fermentation because propionogens are facultative anaerobic bacteria.

  3. DICER-ARGONAUTE2 complex in continuous fluorogenic assays of RNA interference enzymes.

    Directory of Open Access Journals (Sweden)

    Mark A Bernard

    Full Text Available Mechanistic studies of RNA processing in the RNA-Induced Silencing Complex (RISC have been hindered by lack of methods for continuous monitoring of enzymatic activity. "Quencherless" fluorogenic substrates of RNAi enzymes enable continuous monitoring of enzymatic reactions for detailed kinetics studies. Recombinant RISC enzymes cleave the fluorogenic substrates targeting human thymidylate synthase (TYMS and hypoxia-inducible factor 1-α subunit (HIF1A. Using fluorogenic dsRNA DICER substrates and fluorogenic siRNA, DICER+ARGONAUTE2 mixtures exhibit synergistic enzymatic activity relative to either enzyme alone, and addition of TRBP does not enhance the apparent activity. Titration of AGO2 and DICER in enzyme assays suggests that AGO2 and DICER form a functional high-affinity complex in equimolar ratio. DICER and DICER+AGO2 exhibit Michaelis-Menten kinetics with DICER substrates. However, AGO2 cannot process the fluorogenic siRNA without DICER enzyme, suggesting that AGO2 cannot self-load siRNA into its active site. The DICER+AGO2 combination processes the fluorogenic siRNA substrate (Km=74 nM with substrate inhibition kinetics (Ki=105 nM, demonstrating experimentally that siRNA binds two different sites that affect Dicing and AGO2-loading reactions in RISC. This result suggests that siRNA (product of DICER bound in the active site of DICER may undergo direct transfer (as AGO2 substrate to the active site of AGO2 in the DICER+AGO2 complex. Competitive substrate assays indicate that DICER+AGO2 cleavage of fluorogenic siRNA is specific, since unlabeled siRNA and DICER substrates serve as competing substrates that cause a concentration-dependent decrease in fluorescent rates. Competitive substrate assays of a series of DICER substrates in vitro were correlated with cell-based assays of HIF1A mRNA knockdown (log-log slope=0.29, suggesting that improved DICER substrate designs with 10-fold greater processing by the DICER+AGO2 complex can provide a

  4. Radiotoxicity and decay heat power of spent nuclear fuel of VVER type reactors at long-term storage.

    Science.gov (United States)

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

    2005-01-01

    Radiotoxicity and decay heat power of the spent nuclear fuel of VVER-1000 type reactors are calculated during storage time up to 300,000 y. Decay heat power of radioactive waste (radwaste) determines parameters of the heat removal system for the safe storage of spent nuclear fuel. Radiotoxicity determines the radiological hazard of radwaste after its leakage and penetration into the environment. PMID:16381764

  5. R and D relative to the serious accidents in the PWR type reactors: assessment and perspectives

    International Nuclear Information System (INIS)

    This document presents the current state of the research relative to the grave accidents realized in France and abroad. It aims at giving the most exhaustive possible and objective vision of this original field of research. He allows to contribute to the identification and to the hierarchical organization of the needs of R and D, this hierarchical organization in front of, naturally, to be completed by a strong lighting on needs in terms of safety analyses associated with the different risks and the physical phenomena, in particular with the support of probability evaluations of safety level 2, whose the level of sharpness must be sufficient not to hide, by construction, physical phenomena of which the limited knowledge leads to important uncertainties. Let us note that neither the safety analyses, nor the E.P.S. 2 are presented in this document. This report presents the physical phenomena which can arise during a grave accident, in the reactor vessel and in the reactor containment, their chain and the means allowing to ease the effects. The corresponding scenarios are presented to the chapter 2. The chapter 3 is dedicated to the progress of the accident in the reactor vessel; the degradation of the core in reactor vessel (3.1), the behavior of the corium in bottom of reactor vessel (3.2) the break of the reactor vessel (3.3) and the fusion in pressure (3.4) are thus handled there. The chapter 4 concerns the phenomena which can lead to a premature failure of the containment, namely the direct heating of gases of the containment (4.1), the hydrogen risk (4.2) and the vapor explosion (4.3). The phenomenon which can lead to a delayed failure from the containment, namely the interaction corium-concrete, is approached on the chapter 5. The chapter 6 is dedicated to the problems connected to the keeping back and to the corium cooling in reactor vessel and out of reactor vessel, namely the keeping back in reactor vessel by re-flooding of the primary circuit or by re

  6. Pyrolysis of aseptic packages (tetrapak) in a laboratory screw type reactor and secondary thermal/catalytic tar decomposition.

    Science.gov (United States)

    Haydary, J; Susa, D; Dudáš, J

    2013-05-01

    Pyrolysis of aseptic packages (tetrapak cartons) in a laboratory apparatus using a flow screw type reactor and a secondary catalytic reactor for tar cracking was studied. The pyrolysis experiments were realized at temperatures ranging from 650 °C to 850 °C aimed at maximizing of the amount of the gas product and reducing its tar content. Distribution of tetrapak into the product yields at different conditions was obtained. The presence of H2, CO, CH4, CO2 and light hydrocarbons, HCx, in the gas product was observed. The Aluminum foil was easily separated from the solid product. The rest part of char was characterized by proximate and elemental analysis and calorimetric measurements. The total organic carbon in the tar product was estimated by elemental analysis of tars. Two types of catalysts (dolomite and red clay marked AFRC) were used for catalytic thermal tar decomposition. Three series of experiments (without catalyst in a secondary cracking reactor, with dolomite and with AFRC) at temperatures of 650, 700, 750, 800 and 850 °C were carried out. Both types of catalysts have significantly affected the content of tars and other components in pyrolytic gases. The effect of catalyst on the tetrapack distribution into the product yield on the composition of gas and on the total organic carbon in the tar product is presented in this work. PMID:23428565

  7. Properties of bio-oil generated by a pyrolysis of forest cedar residuals with the movable Auger-type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nishimura, Shun; Ebitani, Kohki, E-mail: ebitani@jaist.ac.jp [School of Materials Science, Japan Advanced Institute of Science and Technology, 1-1 Asahidai, Nomi, Ishikawa 923-1292 (Japan); Miyazato, Akio [Nanotechnology Center, Japan Advanced Institute of Science and Technology, 1-1 Asahidai, Nomi, Ishikawa 923-1292 (Japan)

    2016-02-01

    Our research project has developed the new movable reactor for bio-oil production in 2013 on the basis of Auger-type system. This package would be a great impact due to the concept of local production for local consumption in the hilly and mountainous area in not only Japan but also in the world. Herein, we would like to report the properties of the bio-oil generated by the developing Auger-type movable reactor. The synthesized bio-oil possessed C: 46.2 wt%, H: 6.5 wt%, N: wt%, S: <0.1 wt%, O: 46.8 wt% and H{sub 2}O: 18.4 wt%, and served a good calorific value of 18.1 MJ/kg. The spectroscopic and mass analyses such as FT-IR, GC-MS, {sup 13}C-NMR and FT-ICR MS supported that the bio-oil was composed by the fine mixtures of methoxy phenols and variety of alcohol or carboxylic acid functional groups. Thus, it is suggested that the bio-oil generated by the new movable Auger-type reactor has a significant potential as well as the existing bio-oil reported previously.

  8. Online failed fuel identification using delayed neutron detector signals in pool type reactors

    International Nuclear Information System (INIS)

    In todays world, nuclear reactors are at the forefront of modern day innovation and reactor designs are increasingly incorporating cutting edge technology. It is of utmost importance to detect failure or defects in any part of a nuclear reactor for healthy operation of reactor as well as the safety aspects of the environment. Despite careful fabrication and manufacturing of fuel pins, there is a chance of clad failure. After fuel pin clad rupture takes place, it allows fission products to enter in to sodium pool. There are some potential consequences due to this such as Total Instantaneous Blockage (TIB) of coolant and primary component contamination. At present, the failed fuel detection techniques such as cover gas monitoring (alarming the operator), delayed neutron detection (DND-automatic trip) and standalone failed fuel localization module (FFLM) are exercised in various reactors. The first technique is a quantitative measurement of increase in the cover gas activity background whereas DND system causes automatic trip on detecting certain level of activity during clad wet rupture. FFLM is subsequently used to identify the failed fuel subassembly. The later although accurate, but mainly suffers from downtime and reduction in power during identification process. The proposed scheme, reported in this paper, reduces the operation of FFLM by predicting the faulty sector and therefore reducing reactor down time and thermal shocks. The neutron evolution pattern gets modulated because fission products are the delay neutron precursors. When they travel along with coolant to Intermediate heat Exchangers, experienced three effects i.e. delay; decay and dilution which make the neutron pulse frequency vary depending on the location of failed fuel sub assembly. This paper discusses the method that is followed to study the frequency domain properties, so that it is possible to detect exact fuel subassembly failure online, before the reactor automatically trips. (author)

  9. Evaluation of pressure transitories in BWR type reactors using the BWRDYN code; Evaluacion de transitorios de presion en reactores tipo BWR usando el codigo BWRDYN

    Energy Technology Data Exchange (ETDEWEB)

    Rodriguez P, J.A. [ESIME, Unidad Profesional Azcapotzalco, Av. de las Granjas 682, 02550 Mexico D.F. (Mexico)]. e-mail: jrodriguez@ipn.mx

    2007-07-01

    Several simulations of pressure transitory for a nucleo electric power station with BWR/4 type reactor were carried out. The simulated pressure transitories were made for the Peach Bottom 2 Nucleo electric central. Also, it was carried out for the same Plant the simulation of the turbine shot with derivation to the main condenser, of the reference case (benchmark) outlined by the Organization for the Cooperation and the Economic Development and of the Commission Regulatory in Nuclear matter of the United States of America. As tool to carry out the simulations of the transitory ones, the BWRDYN code developed by the Japan Energy Research Institute was used. Among the main suppositions and models that it includes the BWRDYN code its can be mentioned: a) that of punctual kinetics that calculates the neutron flow; for the calculation of the fuel temperature, this it is divided in nodes in the radial and axial directions, the wrapper is considered like a region in the radial direction; c) the pressure is supposed that it is uniform inside the reactor vessel; and d) the thermal hydraulic pattern of the reactor vessel is divided in five regions and the core is divided in several nodes to take into account the distribution of holes in the axial direction. The modeling of the control systems of the feeding water system is also included, of the pressure regulator and of the recirculation system. The systems of what is known as plant balance are also modeled. The numeric results of the simulations provide valuable information of the behavior of the nucleo electric central. The obtained results of the simulation of the reference case agree acceptably with the measurements data, when comparing them with the measurements made in the Peach Bottom 2 Central. The obtained results of each simulation are fundamental to evaluate the transitory one, as well as to delineate the sequence and the impact of diverse events that they happen during the same one transitory. In the case of the

  10. Mechanical, chemical and radiological characterization of the graphite of the UNGG reactors type; Caracterisation mecanique, chimique et radiologique du graphite des reacteurs de la filiere UNGG

    Energy Technology Data Exchange (ETDEWEB)

    Bresard, I.; Bonal, J.P

    2000-07-01

    In the framework of UNGG reactors type dismantling procedures, the characterization of the graphite, used as moderator, has to be realized. This paper presents the mechanical, chemical and radiological characterizations, the properties measured and gives some results in the case of the Bugey 1 reactor. (A.L.B.)

  11. Nuclear research reactors

    International Nuclear Information System (INIS)

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

  12. Thermal Hydraulic Characteristics of Fuel Defects in Plate Type Nuclear Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bodey, Isaac T [ORNL

    2014-05-01

    Turbulent flow coupled with heat transfer is investigated for a High Flux Isotope Reactor (HFIR) fuel plate. The Reynolds Averaged Navier-Stokes Models are used for fluid dynamics and the transfer of heat from a thermal nuclear fuel plate using the Multi-physics code COMSOL. Simulation outcomes are compared with experimental data from the Advanced Neutron Source Reactor Thermal Hydraulic Test Loop. The computational results for the High Flux Isotope Reactor core system provide a more physically accurate simulation of this system by modeling the turbulent flow field in conjunction with the diffusion of thermal energy within the solid and fluid phases of the model domain. Recommendations are made regarding Nusselt number correlations and material properties for future thermal hydraulic modeling efforts

  13. A Primary Sequence Analysis of the ARGONAUTE Protein Family in Plants.

    Science.gov (United States)

    Rodríguez-Leal, Daniel; Castillo-Cobián, Amanda; Rodríguez-Arévalo, Isaac; Vielle-Calzada, Jean-Philippe

    2016-01-01

    Small RNA (sRNA)-mediated gene silencing represents a conserved regulatory mechanism controlling a wide diversity of developmental processes through interactions of sRNAs with proteins of the ARGONAUTE (AGO) family. On the basis of a large phylogenetic analysis that includes 206 AGO genes belonging to 23 plant species, AGO genes group into four clades corresponding to the phylogenetic distribution proposed for the ten family members of Arabidopsis thaliana. A primary analysis of the corresponding protein sequences resulted in 50 sequences of amino acids (blocks) conserved across their linear length. Protein members of the AGO4/6/8/9 and AGO1/10 clades are more conserved than members of the AGO5 and AGO2/3/7 clades. In addition to blocks containing components of the PIWI, PAZ, and DUF1785 domains, members of the AGO2/3/7 and AGO4/6/8/9 clades possess other consensus block sequences that are exclusive of members within these clades, suggesting unforeseen functional specialization revealed by their primary sequence. We also show that AGO proteins of animal and plant kingdoms share linear sequences of blocks that include motifs involved in posttranslational modifications such as those regulating AGO2 in humans and the PIWI protein AUBERGINE in Drosophila. Our results open possibilities for exploring new structural and functional aspects related to the evolution of AGO proteins within the plant kingdom, and their convergence with analogous proteins in mammals and invertebrates. PMID:27635128

  14. Argonaute-2-null embryonic stem cells are retarded in self-renewal and differentiation

    Indian Academy of Sciences (India)

    P Chandra Shekar; Adnan Naim; D Partha Sarathi; Satish Kumar

    2011-09-01

    RNA interference (RNAi) pathways regulate self-renewal and differentiation of embryonic stem (ES) cells. Argonaute 2 (Ago2) is a vital component of RNA-induced silencing complex (RISC) and the only Ago protein with slicer activity. We generated Ago2-deficient ES cells by conditional gene targeting. Ago2-deficient ES cells are defective in the small-RNA-mediated gene silencing and are significantly compromised in biogenesis of mature microRNA. The self-renewal rate of Ago2-deficient ES cells is affected due to failure of silencing of Cdkn1a by ES-cell-specific microRNAs (miRNA) in the absence of Ago2. Interestingly, unlike Dicer- and Dgcr8-deficient ES cells, they differentiate to all three germ layers both in vivo and in vitro. However, early differentiation of Ago2-deficient ES cells is delayed by 2–4 days as indicated by persistence of higher levels of self-renewal/ pluripotency markers during differentiation. Further, appearance of morphological and differentiation markers is also delayed during the differentiation. In this study we show that Ago2 is essential for normal self-renewal and differentiation. Also, our data suggest that self-renewal and differentiation of ES cells are regulated by both siRNA and miRNA pathways.

  15. Novel Insights into Guide RNA 5′-Nucleoside/Tide Binding by Human Argonaute 2

    Directory of Open Access Journals (Sweden)

    Munishikha Kalia

    2015-12-01

    Full Text Available The human Argonaute 2 (hAgo2 protein is a key player of RNA interference (RNAi. Upon complex formation with small non-coding RNAs, the protein initially interacts with the 5′-end of a given guide RNA through multiple interactions within the MID domain. This interaction has been reported to show a strong bias for U and A over C and G at the 5′-position. Performing molecular dynamics simulations of binary hAgo2/OH–guide–RNA complexes, we show that hAgo2 is a highly flexible protein capable of binding to guide strands with all four possible 5′-bases. Especially, in the case of C and G this is associated with rather large individual conformational rearrangements affecting the MID, PAZ and even the N-terminal domains to different degrees. Moreover, a 5′-G induces domain motions in the protein, which trigger a previously unreported interaction between the 5′-base and the L2 linker domain. Combining our in silico analyses with biochemical studies of recombinant hAgo2, we find that, contrary to previous observations, hAgo2 is capable of functionally accommodating guide strands regardless of the 5′-base.

  16. Anatomy of RISC: how do small RNAs and chaperones activate Argonaute proteins?

    Science.gov (United States)

    Nakanishi, Kotaro

    2016-09-01

    RNA silencing is a eukaryote-specific phenomenon in which microRNAs and small interfering RNAs degrade messenger RNAs containing a complementary sequence. To this end, these small RNAs need to be loaded onto an Argonaute protein (AGO protein) to form the effector complex referred to as RNA-induced silencing complex (RISC). RISC assembly undergoes multiple and sequential steps with the aid of Hsc70/Hsp90 chaperone machinery. The molecular mechanisms for this assembly process remain unclear, despite their significance for the development of gene silencing techniques and RNA interference-based therapeutics. This review dissects the currently available structures of AGO proteins and proposes models and hypotheses for RISC assembly, covering the conformation of unloaded AGO proteins, the chaperone-assisted duplex loading, and the slicer-dependent and slicer-independent duplex separation. The differences in the properties of RISC between prokaryotes and eukaryotes will also be clarified. WIREs RNA 2016, 7:637-660. doi: 10.1002/wrna.1356 For further resources related to this article, please visit the WIREs website. PMID:27184117

  17. Transcriptional regulation mechanism mediated by miRNA-DNA•DNA triplex structure stabilized by Argonaute.

    Science.gov (United States)

    Toscano-Garibay, Julia D; Aquino-Jarquin, Guillermo

    2014-11-01

    Transcription regulation depends on interactions between repressor or activator proteins with promoter sequences, while post-transcriptional regulation typically relies on microRNA (miRNA) interaction with sequences in 5' and 3'-Untranslated regions (UTRs) of messenger RNA (mRNA). However, several pieces of evidence suggest that miRNA:Argonaute (AGO) complexes may also suppress transcription through RNA interference (RNAi) components and epigenetic mechanisms. However, recent observations suggest that miRNA-induced transcriptional silencing could be exerted by an unknown mechanism independent of chromatin modifiers. The RNA-DNA•DNA triplex structure has emerged as an important RNA tertiary motif in which successive non-canonical base pairs form between a DNA-DNA duplex and a third strand. Frequently, promoters have Purine (PU)-rich tracts, and some Triplex-forming oligonucleotides (TFOs) targeting these regulatory regions have been shown to inhibit transcription selectively. Here, we summarize observations suggesting that miRNAs exert regulation over promoter regions through miRNA-DNA•DNA triplex structure formation stabilized by AGO proteins which represents a plausible model of RNA-mediated Transcriptional gene silencing (TGS). PMID:25086339

  18. A Primary Sequence Analysis of the ARGONAUTE Protein Family in Plants

    Science.gov (United States)

    Rodríguez-Leal, Daniel; Castillo-Cobián, Amanda; Rodríguez-Arévalo, Isaac; Vielle-Calzada, Jean-Philippe

    2016-01-01

    Small RNA (sRNA)-mediated gene silencing represents a conserved regulatory mechanism controlling a wide diversity of developmental processes through interactions of sRNAs with proteins of the ARGONAUTE (AGO) family. On the basis of a large phylogenetic analysis that includes 206 AGO genes belonging to 23 plant species, AGO genes group into four clades corresponding to the phylogenetic distribution proposed for the ten family members of Arabidopsis thaliana. A primary analysis of the corresponding protein sequences resulted in 50 sequences of amino acids (blocks) conserved across their linear length. Protein members of the AGO4/6/8/9 and AGO1/10 clades are more conserved than members of the AGO5 and AGO2/3/7 clades. In addition to blocks containing components of the PIWI, PAZ, and DUF1785 domains, members of the AGO2/3/7 and AGO4/6/8/9 clades possess other consensus block sequences that are exclusive of members within these clades, suggesting unforeseen functional specialization revealed by their primary sequence. We also show that AGO proteins of animal and plant kingdoms share linear sequences of blocks that include motifs involved in posttranslational modifications such as those regulating AGO2 in humans and the PIWI protein AUBERGINE in Drosophila. Our results open possibilities for exploring new structural and functional aspects related to the evolution of AGO proteins within the plant kingdom, and their convergence with analogous proteins in mammals and invertebrates. PMID:27635128

  19. Bacterial Expression of Mouse Argonaute 2 for Functional and Mutational Studies

    Directory of Open Access Journals (Sweden)

    Aniello Russo

    2010-02-01

    Full Text Available RNA interference (RNAi is a post-transcriptional gene-silencing process that occurs in many eukaryotic organisms upon intracellular exposure to double-stranded RNA. Argonaute 2 (Ago2 protein is the catalytic engine of mammalian RNAi. It contains a PIWI domain that is structurally related to RNases H and possibly shares with them a two-metal-ion catalysis mechanism. Here we describe the expression in E. coli of mouse Ago2 and testing of its enzymatic activity in a RISC assay, i.e., for the ability to cleave a target RNA in a single position specified by a complementary small interfering RNA (siRNA. The results show that the enzyme can load the siRNA and cleave the complementary RNA in absence of other cellular factors, as described for human Ago2. It was also found that mutation of Arg669, a residue previously proposed to be involved in substrate and/or B metal ion binding, doesn’t affect the enzymatic activity, suggesting that this residue doesn’t belong to the active site.

  20. A primary sequence analysis of the ARGONAUTE protein family in plants.

    Directory of Open Access Journals (Sweden)

    Daniel Rodriguez-Leal

    2016-08-01

    Full Text Available Small RNA (sRNA-mediated gene silencing represents a conserved regulatory mechanism controlling a wide diversity of developmental processes through interactions of sRNAs with proteins of the ARGONAUTE (AGO family. On the basis of a large phylogenetic analysis that includes 206 AGO genes belonging to 23 plant species, AGO genes group into four clades corresponding to the phylogenetic distribution proposed for the ten family members of Arabidopsis thaliana. A primary analysis of the corresponding protein sequences resulted in 50 sequences of amino acids (blocks conserved across their linear length. Protein members of the AGO4/6/8/9 and AGO1/10 clades are more conserved than members of the AGO5 and AGO2/3/7 clades. In addition to blocks containing components of the PIWI, PAZ, and DUF1785 domains, members of the AGO2/3/7 and AGO4/6/8/9 clades possess other consensus block sequences that are exclusive of members within these clades, suggesting unforeseen functional specialization revealed by their primary sequence. We also show that AGO proteins of animal and plant kingdoms share linear sequences of blocks that include motifs involved in posttranslational modifications such as those regulating AGO2 in humans and the PIWI protein AUBERGINE in Drosophila. Our results open possibilities for exploring new structural and functional aspects related to the evolution of AGO proteins within the plant kingdom, and their convergence with analogous proteins in mammals and invertebrates.

  1. Contribution to the study of the conversion PWR type reactors to the thorium cycle

    International Nuclear Information System (INIS)

    The use of the thorium cycle in PWR reactors is discussed. The fuel has been calculated in the equilibrium condition for a economic comparison with the uranium cycle (in the same condition). First of all, a code named EQUILIBRIO has been developed for the fuel equilibrium calculation. The results gotten by this code, were introduced in the LEOPARD code for the fuel depletion calculation (in the equilibrium cycle). Same important physics details of fuel depletion are studied, for instance: the neutron balance, power sharing, fuel burnup, etc. The calculations have been done taking as reference the Angra-1 PWR reactor. (Author)

  2. Different types of cryogenics Pellet injection systems (PIS) for fusion reactor

    OpenAIRE

    Devarshi Patel; Alkesh Mavani

    2014-01-01

    Fusion reactor is the one of the most capable option for generating the large amount of energy in future. Fusion means joining smaller nuclei (the plural of nucleus) to make a larger nucleus and release energy in the form of neutrons.The sun uses nuclear fusion of hydrogen atoms into helium atoms. This gives off heat and light and other radiation. Hydrogen is used as the fuel in the fusion reactor. We have to inject the solid hydrogen pellet into the tokamak as per the require...

  3. Basic criteria in material selection for equipment of NPPs containing the type WWER reactors

    International Nuclear Information System (INIS)

    General principles of materials selection for NPP equipment are reported with due regard for operating conditions. The emphasis is on the selection of materials for equipment elements of water cooled and moderated reactors. Based on a 25 year operating experience a conclusion is made that on reactor materials selection the following properties should be taken into consideration: corrosion resistance, mechanical strength, low cycle fatigue strength and resistance to brittle fracture. Experimental data on the behaviour of a number of carbon, alloy and heat resistant steels under operating conditions are presented

  4. Computational simulation of the natural circulation occurring in an experimental test section of a pool type research reactor

    International Nuclear Information System (INIS)

    The present work presents a computational simulation of the natural circulation phenomenon developing in an experimental test section of a pool type research reactor. The test section has been designed using a reduced scale in height 1:4.7 in relation to a pool type 30 MW research reactor prototype. It comprises a cylindrical vessel, which is opened to atmosphere, and representing the reactor pool; a natural circulation pipe, a lower plenum, and a heater containing electrical resistors in rectangular plate format, which represents the fuel elements, with a chimney positioned on the top of the resistor assembly. In the computational simulation, it was used a commercial CFD software, without any turbulence model. Besides, in the presence of the natural circulation, a laminar flow has been assumed and the equations of the mass conservation, momentum and energy were solved by the finite element method. In addition, the results of the simulation are presented in terms of velocities and temperatures differences, respectively: at inlet and outlet of the heater and of the natural circulation pipe. (author)

  5. Determination of neutron energy spectrum at a pneumatic rabbit station of a typical swimming pool type material test research reactor

    International Nuclear Information System (INIS)

    The method of multiple foil activation was used to measure the neutron energy spectrum, experimentally, at a rabbit station of Pakistan Research Reactor-1 (PARR-1), which is a typical swimming pool type material test research reactor. The computer codes MSITER and SANDBP were used to adjust the spectrum. The pre-information required by the adjustment codes was obtained by modelling the core and its surroundings in three-dimensions by using the one dimensional transport theory code WIMS-D/4 and the multidimensional finite difference diffusion theory code CITATION. The input spectrum covariance information required by MSITER code was also calculated from the CITATION output. A comparison between calculated and adjusted spectra shows a good agreement

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

    International Nuclear Information System (INIS)

    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.)

  7. Sodium removing facility for core-constitutional elements of FBR type reactor

    International Nuclear Information System (INIS)

    Reactor core-constitutional elements as spent reactor core fuel assemblies are contained in a containing vessel. An inert gas (N2, Ar or He) is filled in the containing vessel through an inert gas supply channel. The temperature of the inert gas is raised by the remaining after heat of the reactor core-constitutional elements. The inert gas is circulated and heated through a preheating circuit by driving a recycling gas blower and returned to the containing vessel. If the inert gas is heated to a predetermined temperature, metal sodium deposited on the surface of the materials of the reactor core-constitutional elements is evaporated. Next, a vacuum pump unit of a vacuum exhaustion channel is driven to suck an inert gas entraining sodium vapor in the containing vessel, and the sodium vapor is cooled, condensed thereby separated in a sodium separator. Then, the inert gas at a low temperature is introduced to a vacuum exhaustion channel to remove and discharge remained sodium vapor by a sodium trap. (I.N.)

  8. Modelling and simulation the radioactive source-term of fission products in PWR type reactors

    International Nuclear Information System (INIS)

    The source-term was defined with the purpose the quantify all radioactive nuclides released the nuclear reactor in the case of accidents. Nowadays the source-term is limited to the coolant of the primary circuit of reactors and may be measured or modelled with computer coders such as the TFP developed in this work. The calculational process is based on the linear chain techniques used in the CINDER-2 code. The TFP code considers forms of fission products release from the fuel pellet: Recoil, Knockout and Migration. The release from the gap to the coolant fluid is determined from the ratio between activity measured in the coolant and calculated activity in the gap. Considered the operational data of SURRY-1 reactor, the TFP code was run to obtain the source=term of this reactor. From the measured activities it was verified the reliability level of the model and the employed computational logic. The accuracy of the calculated quantities were compared to the measured data was considered satisfactory. (author)

  9. Core characteristics on a hybrid type fast reactor system combined with proton accelerator

    Energy Technology Data Exchange (ETDEWEB)

    Kowata, Yasuki; Otsubo, Akira [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1997-06-01

    In our study on a hybrid fast reactor system, we have investigated it from the view point of transmutation ability of trans-uranium (TRU) nuclide making the most effective use of special features (controllability, hard neutron spectrum) of the system. It is proved that a proton beam is superior in generation of neutrons compared with an electron beam. Therefore a proton accelerator using spallation reaction with a target nucleus has an advantage to transmutation of TRU than an electron one. A fast reactor is expected to primarily have a merit that the reactor can be operated for a long term without employment of highly enriched plutonium fuel by using external neutron source such as the proton accelerator. Namely, the system has a desirable characteristic of being possible to self-sustained fissile plutonium. Consequently in the present report, core characteristics of the system were roughly studied by analyses using 2D-BURN code. The possibility of self-sustained fuel was investigated from the burnup and neutronic calculation in a cylindrical core with 300w/cc of power density without considering a target material region for the accelerator. For a reference core of which the height and the radius are both 100 cm, there is a fair prospect that a long term reactor operation is possible with subsequent refueling of natural uranium, if the medium enriched (around 10wt%) uranium or plutonium fuels are fully loaded in the initial core. More precise analyses will be planed in a later fiscal year. (author)

  10. Decay heat removal in pool type fast reactor using passive systems

    Energy Technology Data Exchange (ETDEWEB)

    Parthasarathy, U. [Thermal Hydraulics Section, Nuclear and Safety Engineering Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603102 (India); Sundararajan, T. [Department of Mechanical Engineering, IIT-Madras, Chennai 600 036 (India); Balaji, C., E-mail: balaji@iitm.ac.in [Department of Mechanical Engineering, IIT-Madras, Chennai 600 036 (India); Velusamy, K.; Chellapandi, P.; Chetal, S.C. [Thermal Hydraulics Section, Nuclear and Safety Engineering Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603102 (India)

    2012-09-15

    Highlights: Black-Right-Pointing-Pointer Three dimensional thermal hydraulic analysis of decay heat system in a fast reactor model predictions compared with experimental results from PHENIX. Black-Right-Pointing-Pointer Calculations confirm adequacy of natural convection in decay heat removal. Black-Right-Pointing-Pointer Inter-wrapper flow found to reduce peak temperatures by 50 K in the blanket zone. - Abstract: Post shutdown decay heat removal in a fast reactor is one of the most important safety functions which must be accomplished with a very high reliability. To achieve high reliability, the fast breeder reactor design has emphasized on passive or near passive decay heat removal systems utilizing the natural convection in the heat removal path. A typical passive decay heat removal system used in recent designs of fast breeder reactors consists of a sodium to sodium heat exchanger and sodium to air heat exchanger which together remove heat directly from the hot pool to the final heat sink, which is air. Since these are safety systems, it is necessary to confirm the design with detailed numerical analysis. The numerical studies include pool hydraulics, natural convection phenomena in closed loops, flow through narrow gaps between SA, multi-scale modeling, etc. Toward understanding the evolution of thermal hydraulic parameters during natural convection decay heat removal, a three-dimensional CFD model for the primary system coupled with an appropriate one-dimensional model for the secondary system is proposed. The model has been validated against the results of natural convection test conducted in PHENIX reactor. Adopting the model for the Indian PFBR, six different decay heat removal cases have been studied which bring out the effect of safety grade decay heat removal system (SGDHRS) capacity, secondary sodium inventory and inter-wrapper flow heat transfer on the subassembly outlet temperatures that are important for safety evaluation of the reactor. From the

  11. Decay heat removal in pool type fast reactor using passive systems

    International Nuclear Information System (INIS)

    Highlights: ► Three dimensional thermal hydraulic analysis of decay heat system in a fast reactor model predictions compared with experimental results from PHENIX. ► Calculations confirm adequacy of natural convection in decay heat removal. ► Inter-wrapper flow found to reduce peak temperatures by 50 K in the blanket zone. - Abstract: Post shutdown decay heat removal in a fast reactor is one of the most important safety functions which must be accomplished with a very high reliability. To achieve high reliability, the fast breeder reactor design has emphasized on passive or near passive decay heat removal systems utilizing the natural convection in the heat removal path. A typical passive decay heat removal system used in recent designs of fast breeder reactors consists of a sodium to sodium heat exchanger and sodium to air heat exchanger which together remove heat directly from the hot pool to the final heat sink, which is air. Since these are safety systems, it is necessary to confirm the design with detailed numerical analysis. The numerical studies include pool hydraulics, natural convection phenomena in closed loops, flow through narrow gaps between SA, multi-scale modeling, etc. Toward understanding the evolution of thermal hydraulic parameters during natural convection decay heat removal, a three-dimensional CFD model for the primary system coupled with an appropriate one-dimensional model for the secondary system is proposed. The model has been validated against the results of natural convection test conducted in PHENIX reactor. Adopting the model for the Indian PFBR, six different decay heat removal cases have been studied which bring out the effect of safety grade decay heat removal system (SGDHRS) capacity, secondary sodium inventory and inter-wrapper flow heat transfer on the subassembly outlet temperatures that are important for safety evaluation of the reactor. From the results, it is concluded that the delay in initiation of SGDHRS, replacement

  12. Crystal structure of A. aeolicus argonaute, a site-specific DNA-guided endoribonuclease, provides insights into RISC-mediated mRNA cleavage

    Energy Technology Data Exchange (ETDEWEB)

    Yuan,Y.; Pei, Y.; Ma, J.; Kuryavyi, V.; Zhadina, M.; Meister, G.; Chen, H.; Dauter, Z.; Tuschi, T.; Patel, D.

    2005-01-01

    Argonaute (Ago) proteins constitute a key component of the RNA-induced silencing complex (RISC). We report the crystal structure of Aquifex aeolicus Ago (Aa-Ago) together with binding and cleavage studies, which establish this eubacterial Ago as a bona fide guide DNA strand-mediated site-specific RNA endonuclease. We have generated a stereochemically robust model of the complex, where the guide DNA-mRNA duplex is positioned within a basic channel spanning the bilobal interface, such that the 5' phosphate of the guide strand can be anchored in a basic pocket, and the mRNA can be positioned for site-specific cleavage by RNase H-type divalent cation-coordinated catalytic Asp residues of the PIWI domain. Domain swap experiments involving chimeras of human Ago (hAgo1) and cleavage-competent hAgo2 reinforce the role of the PIWI domain in 'slicer' activity. We propose a four-step Ago-mediated catalytic cleavage cycle model, which provides distinct perspectives into the mechanism of guide strand-mediated mRNA cleavage within the RISC.

  13. Intelligent information data base of flow boiling characteristics in once-through steam generator for integrated type marine water reactor

    International Nuclear Information System (INIS)

    Valuable experimental knowledge with flow boiling characteristics of the helical-coil type once-through steam generator was converted into an intelligent information data base program. The program was created as a windows application using the Visual Basic. Main functions of the program are as follows: (1) steady state flow boiling analysis of any helical-coil type once-through steam generator, (2) analysis and comparison with the experimental data, (3) reference and graph display of the steady state experimental data, (4) reference of the flow instability experimental data and display of the instability threshold correlated by each parameter, (5) summary of the experimental apparatus. (6) menu bar such as a help and print. In the steady state analysis, the region lengths of subcooled boiling, saturated boiling, and super-heating, and the temperature and pressure distributions etc. for secondary water calculated. Steady state analysis results agreed well with the experimental data, with the exception of the pressure drop at high mass velocity. The program will be useful for the design of not only the future integrated type marine water reactor but also the small sized water reactor with helical-coil type steam generator

  14. The reactor core TRIGA Mark-III with fuels type 30/20; El nucleo del reactor TRIGA Mark-III con combustible tipo 30/20

    Energy Technology Data Exchange (ETDEWEB)

    Aguilar H, F., E-mail: fortunato.aguilar@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2012-10-15

    This work describes the calculation series carried out with the program MCNP5 in order to define the configuration of the reactor core with fuels 30/20 (fuels with 30% of uranium content in the Or-Zr-H mixture and a nominal enrichment of 20%). To select the configuration of the reactor core more appropriate to the necessities and future uses of the reactor, the following criterions were taken into account: a) the excess in the reactor reactivity, b) the switch out margin and c) to have new irradiation facilities inside the reactor core. Taking into account these criterions is proceeded to know the characteristics of the components that form the reactor core (dimensions, geometry, materials, densities and positions), was elaborated a base model of the reactor core, for the MCNP5 code, with a configuration composed by 85 fuel elements, 4 control bars and the corresponding structural elements. The high reactivity excess obtained with this model, gave the rule to realize other models of the reactor core in which the reactivity excess and the switch out margin were approximate to the values established in the technical specifications of the reactor operation. Several models were realized until finding the satisfactory model; this is composite for 74 fuels, 4 control bars and 6 additional experimental positions inside the reactor core. (Author)

  15. Life time of nuclear power plants and new types of reactors

    International Nuclear Information System (INIS)

    This report, realized by the Evaluation Parliamentary Office of scientific and technological choices, aims to answer simple but fundamental questions for the french electric power production. What are the phenomena which may limit the exploitation time of nuclear power plants? How can we fight against the aging, at which cost and with which safety? The first chapter presents the management of the nuclear power plants life time, an essential element of the park optimization but not a sufficient element. The second chapter details the EPR and the other reactors for 2015 as a bond between the today and tomorrow parks. The last chapter deals with the necessity of efforts in the research and development to succeed in 2035 and presents other reactors in project. (A.L.B.)

  16. Heat Transfer Calculation on Plate-Type Fuel Assembly of High Flux Research Reactor

    Directory of Open Access Journals (Sweden)

    Daxin Gong

    2015-01-01

    Full Text Available Heat transfer characteristics of fuel assemblies for a high flux research reactor with a neutron trap are numerically investigated in this study. Single-phase turbulence flow is calculated by a commercial code, FLUENT, where the computational objective covers standard and control fuel assemblies. The simulation is carried out with an inlet coolant velocity varying from 4.5 m/s to 7.5 m/s in hot assemblies. The results indicate that the cladding temperature is always lower than the saturation temperature in the calculated ranges. The temperature rise in the control fuel assembly is smaller than that of the standard fuel assembly. Additionally, the assembly with a hot spot is specially studied, and the safety of the research reactor is also approved.

  17. Application of a novel type impinging streams reactor in solid-liquid enzyme reactions and modeling of residence time distribution using GDB model.

    Science.gov (United States)

    Fatourehchi, Niloufar; Sohrabi, Morteza; Dabir, Bahram; Royaee, Sayed Javid; Haji Malayeri, Adel

    2014-02-01

    Solid-liquid enzyme reactions constitute important processes in biochemical industries. The isomerization of d-glucose to d-fructose, using the immobilized glucose isomerase (Sweetzyme T), as a typical example of solid-liquid catalyzed reactions has been carried out in one stage and multi-stage novel type of impinging streams reactors. Response surface methodology was applied to determine the effects of certain pertinent parameters of the process namely axial velocity (A), feed concentration (B), nozzles' flow rates (C) and enzyme loading (D) on the performance of the apparatus. The results obtained from the conversion of glucose in this reactor were much higher than those expected in conventional reactors, while residence time was decreased dramatically. Residence time distribution (RTD) in a one-stage impinging streams reactor was investigated using colored solution as the tracer. The results showed that the flow pattern in the reactor was close to that in a continuous stirred tank reactor (CSTR). Based on the analysis of flow region in the reactor, gamma distribution model with bypass (GDB) was applied to study the RTD of the reactor. The results indicated that RTD in the impinging streams reactor could be described by the latter model.

  18. Heat transfer burnout in tube-type fuel elements of nuclear power reactors

    International Nuclear Information System (INIS)

    The conditions are formulated under which the results of the experimental research of the boilino. water heat transfer burnout carried out on models may be applied to fuel elements of nuclear reactors. Experimental material providing data on the heat transfer burnout was expanded by the results of measurements of the uneven (cosine) longitudinal distribution of heat sources. The results of the effects of helical fins or wires on heat transfer burnout are presented. (F.M.)

  19. A sipping test simulator for identifying defective fuels in MTR type nuclear research reactor

    International Nuclear Information System (INIS)

    Highlights: • This simulator based on windows application of C# programming language. • This simulator could be useful for training of technicians in spent nuclear fuels storage facility. • This simulator is user friendly and easy to learn. - Abstract: Integrity of fuel assemblies is critical to continuous operation of any nuclear reactor. NDT methods and sipping test are practical techniques which are used for this purpose. Assessing the fuel integrity by NDT is a troublesome process which could incur personal overdose due to high radiation, requiring large space, and heavy equipment. Therefore to overcome problems associated with the NDT process, sipping test is widely used. The main purpose of this article is introducing sipping test simulator (STS) which is so important for training. Also, this article describes the procedure and methodology used to perform sipping test on the fuel assemblies either in reactor pool or spent fuel storage pool. A unique ability of this simulator is analyzing direct spectroscopy files from experimental data of a real operating reactor. The sipping test simulator is a full-feature training curriculum in spent nuclear fuels storage technology with a PC-based simulator. This simulator is written in C# programming language for a Windows based computer. The simulator will teach everything needed to know for identifying the fuel defects using sipping test process. As learning the basics of sipping test step wise, a freshman operator will soon be able to accomplish all steps in practice

  20. A novel water perm-selective membrane dual-type reactor concept for Fischer-Tropsch synthesis of GTL (gas to liquid) technology

    International Nuclear Information System (INIS)

    The present study proposes a novel configuration of Fischer-Tropsch synthesis (FTS) reactors in which a fixed-bed water perm-selective membrane reactor is followed by a fluidized-bed hydrogen perm-selective membrane reactor. This novel concept which has been named fixed-bed membrane reactor followed by fluidized-bed membrane reactor (FMFMDR) produces gasoline from synthesis gas. The walls of the tubes of a fixed-bed reactor (water-cooled reactor) of FMFMDR configuration are coated by a high water perm-selective membrane layer. In this new configuration, two membrane reactors instead of one membrane reactor are developed for FTS reactions. In other words, two different membrane layers are used. In order to investigate the performance of FMFMDR, a one-dimensional heterogeneous model is taken into consideration. The simulation results of three schemes named fluidized-bed membrane dual-type reactor (FMDR), FMFMDR and conventional fixed-bed reactor (CR) are presented. They have been compared in terms of temperature, gasoline and CO2 yields, H2 and CO conversions and the water permeation rate through the membrane layer. Results show that the gasoline yield in FMFMDR is higher than the one in FMDR. The FMFMDR configuration not only decreases the undesired product such as CO2 but also produces more gasoline. -- Research highlights: → The application of H-SOD membrane layer in FTS reactors. → Approximate 7.5% and 37% increase in the gasoline yield in terms of [g/g feed x 100] in comparison with FMDR and CR, respectively. → A remarkable decrease in CO2 emission to the environment. → A good configuration mainly due to reduction in catalysts sintering as a result of in situ water removal.

  1. The developmental outcomes of P0-mediated ARGONAUTE destabilization in tomato.

    Science.gov (United States)

    Hendelman, Anat; Kravchik, Michael; Stav, Ran; Zik, Moriyah; Lugassi, Nitsan; Arazi, Tzahi

    2013-01-01

    The plant protein ARGONAUTE1 (AGO1) functions in multiple RNA-silencing pathways, including those of microRNAs, key regulators of growth and development. Genetic analysis of ago1 mutants with informative defects has provided valuable insights into AGO1's biological functions. Tomato encodes two AGO1 homologs (SlAGO1s), but mutants have not been described to date. To analyze SlAGO1s' involvement in development, we confirmed that both undergo decay in the presence of the Polerovirus silencing suppressor P0 and produce a transgenic responder line (OP:P0HA) that, upon transactivation, expresses P0 C-terminally fused to a hemagglutinin (HA) tag (P0HA) and destabilizes SlAGO1s at the site of expression. By crossing OP:P0HA with a battery of driver lines, constitutive as well as organ- and stage-specific SlAGO1 downregulation was induced in the F1 progeny. Activated plants exhibited various developmental phenotypes that partially overlapped with those of Arabidopsis ago1 mutants. Plants that constitutively expressed P0HA had reduced SlAGO1 levels and increased accumulation of miRNA targets, indicating compromised SlAGO1-mediated silencing. Consistent with this, they exhibited pleiotropic morphological defects and their growth was arrested post-germination. Transactivation of P0HA in young leaf and floral organ primordia dramatically modified corresponding organ morphology, including the radialization of leaflets, petals and anthers, suggesting that SlAGO1s' activities are required for normal lateral organ development and polarity. Overall, our results suggest that the OP:P0HA responder line can serve as a valuable tool to suppress SlAGO1 silencing pathways in tomato. The suppression of additional SlAGOs by P0HA and its contribution to the observed phenotypes awaits investigation.

  2. HIV-1 RNAs are Not Part of the Argonaute 2 Associated RNA Interference Pathway in Macrophages.

    Directory of Open Access Journals (Sweden)

    Valentina Vongrad

    Full Text Available MiRNAs and other small noncoding RNAs (sncRNAs are key players in post-transcriptional gene regulation. HIV-1 derived small noncoding RNAs (sncRNAs have been described in HIV-1 infected cells, but their biological functions still remain to be elucidated. Here, we approached the question whether viral sncRNAs may play a role in the RNA interference (RNAi pathway or whether viral mRNAs are targeted by cellular miRNAs in human monocyte derived macrophages (MDM.The incorporation of viral sncRNAs and/or their target RNAs into RNA-induced silencing complex was investigated using photoactivatable ribonucleoside-induced cross-linking and immunoprecipitation (PAR-CLIP as well as high-throughput sequencing of RNA isolated by cross-linking immunoprecipitation (HITS-CLIP, which capture Argonaute2-bound miRNAs and their target RNAs. HIV-1 infected monocyte-derived macrophages (MDM were chosen as target cells, as they have previously been shown to express HIV-1 sncRNAs. In addition, we applied small RNA deep sequencing to study differential cellular miRNA expression in HIV-1 infected versus non-infected MDMs.PAR-CLIP and HITS-CLIP data demonstrated the absence of HIV-1 RNAs in Ago2-RISC, although the presence of a multitude of HIV-1 sncRNAs in HIV-1 infected MDMs was confirmed by small RNA sequencing. Small RNA sequencing revealed that 1.4% of all sncRNAs were of HIV-1 origin. However, neither HIV-1 derived sncRNAs nor putative HIV-1 target sequences incorporated into Ago2-RISC were identified suggesting that HIV-1 sncRNAs are not involved in the canonical RNAi pathway nor is HIV-1 targeted by this pathway in HIV-1 infected macrophages.

  3. Argonaute 2 complexes selectively protect the circulating microRNAs in cell-secreted microvesicles.

    Directory of Open Access Journals (Sweden)

    Limin Li

    Full Text Available Cell-secreted miRNAs are highly stable and can serve as biomarkers for various diseases and signaling molecules in intercellular communication. The mechanism underlying the stability of circulating miRNAs, however, remains incompletely understood. Here we show that Argonaute 2 (Ago2 complexes and microvesicles (MVs provide specific and non-specific protection for miRNA in cell-secreted MVs, respectively. First, the resistance of MV-encapsulated miRNAs to RNaseA was both depended on intact vesicular structure of MVs and protease-sensitive. Second, an immunoprecipitation assay using a probe complementary to human miR-16, a miRNA primarily located in the MVs and showed a strong, protease-sensitive resistance to RNaseA, identified Ago2 as a major miR-16-associated protein. Compared with protein-free miR-16, Ago2-associated miR-16 was resistant to RNaseA in a dose- and time-dependent fashion. Third, when the miR-16/Ago2 complex was disrupted by trypaflavine, the resistance of miR-16 to RNaseA was decreased. In contrast, when the association of miR-16 with the Ago2 complexes was increased during cell apoptosis, although the total amount of miR-16 and Ago2 remained unchanged, the resistance of miR-16 to RNaseA in the MVs was enhanced. A similar correlation between the increase of miR-223/Ago2 association and the resistance of miR-223 against RNaseA was observed during all trans retinoic acid (ATRA-induced cell differentiation of promyelocytic HL60 cells. In conclusion, the association of miRNAs with Ago2 complexes, an event that is linked to cell functional status, plays a critical role in stabilizing the circulating miRNAs in cell-secreted MVs.

  4. Safety Related Investigations of the VVER-1000 Reactor Type by the Coupled Code System TRACE/PARCS

    Science.gov (United States)

    Jaeger, Wadim; Espinoza, Victor Hugo Sánchez; Lischke, Wolfgang

    This study was performed at the Institute of Reactor Safety at the Forschungszentrum Karlsruhe. It is embedded in the ongoing investigations of the international code assessment and maintenance program (CAMP) for qualification and validation of system codes like TRACE(1) and PARCS(2). The chosen reactor type used to validate these two codes was the Russian designed VVER-1000 because the OECD/NEA VVER-1000 Coolant Transient Benchmark Phase 2(3) includes detailed information of the Bulgarian nuclear power plant (NPP) Kozloduy unit 6. The post-test investigations of a coolant mixing experiment have shown that the predicted parameters (coolant temperature, pressure drop, etc.) are in good agreement with the measured data. The coolant mixing pattern, especially in the downcomer, has been also reproduced quiet well by TRACE. The coupled code system TRACE/PARCS which was applied on a postulated main steam line break (MSLB) provided good results compared to reference values and the ones of other participants of the benchmark. The results show that the developed three-dimensional nodalization of the reactor pressure vessel (RPV) is appropriate to describe the coolant mixing phenomena in the downcomer and the lower plenum of a VVER-1000 reactor. This phenomenon is a key issue for investigations of MSLB transient where the thermal hydraulics and the core neutronics are strongly linked. The simulation of the RPV and core behavior for postulated transients using the validated 3D TRACE RPV model, taking into account boundary conditions at vessel in- and outlet, indicates that the results are physically sound and in good agreement to other participant's results.

  5. CARBONACEOUS, NITROGENOUS AND PHOSPHORUS MATTERS REMOVAL FROM DOMESTIC WASTEWATER BY AN ACTIVATED SLUDGE REACTOR OF NITRIFICATION-DENITRIFICATION TYPE

    Directory of Open Access Journals (Sweden)

    MOHAMAD ALI FULAZZAKY

    2009-03-01

    Full Text Available This paper proposes an environmental engineering method based on biotechnology approach as one of the expected solutions that should be considered to implementing the activated sludge for improving the quality of water and living environment, especially to remove the major pollutant elements of domestic wastewater. Elimination of 3 major pollutant elements, i.e., carbon, nitrogen and phosphor containing the domestic wastewater is proposed to carry out biological method of an anoxic-aerobic reactor therein these types of pollutants should be consecutively processed in three steps. Firstly, eliminate the carbonaceous matter in the aerobic reactor. Secondly, to remove the carbonaceous and nitrogenous matters, it is necessary to modify the reactor’s nature from the aerobic condition to an anoxic-aerobic reactor. And finally, when the cycle of nitrification-denitrification is stable to achieve the target’s efficiency of reactor by adding the ferric iron into the activated sludge, it can be continued to remove the carbonaceous, nitrogenous and phosphorous matters simultaneously. The efficiency of carbonaceous and nitrogenous matters removal was confirmed with the effluent standard, COD is less than 100 mgO2/L and the value of global nitrogen is less than 10 mgN/L. The effectiveness of suspended matter removal is higher than 90% and the decantation of activated sludge is very good as identifying the Molhman’s index is below of 120 mL/L. The total phosphorus matter removal is more effective than the soluble phosphorus matter. By maintaining the reactor’s nature at the suitable condition, identifying the range of pH between 6.92 and 7.16 therefore the excellent abatement of phosphor of about 80% is achieving with the molar Fe/P ratio of 1.4.

  6. Thermal performance of fast reactor type uranium-plutonium oxide fuel pins at beginning-life conditions

    International Nuclear Information System (INIS)

    A new thermal conductivity correlation, named as ''PNC'98 equation'', of fast reactor type uranium-plutonium oxide (FR-MOX) fuel was proposed by one of the authors last year. Since maximum operating temperature of fast reactor fuel is generally much higher than that of light water reactor fuel, the predictability of the correlations for FR-MOX fuel at higher temperatures, especially, above 2300K, become more important in case of the fuel pin thermal performance analysis. Although the thermal conductivity measurements at higher temperatures are very scares, power-to-melt (PTM) experiments at beginning-of-life conditions can compensate for the lack of the measurements and be preferable to verify the correlation up to melting point. The PNC'98 equation was applied to the PTM test results irradiated in experimental fast reactor JOYO in order to examine its predictability up to melting point by integral method from a viewpoint of the heat transfer behavior across fuel-to-cladding gap through the fuel pin thermal analysis. Fuel pin irradiation behaviors, such as microstructure evolution and the porosity effect on the heat conduction in fuel pellet, were carefully modeled with some related physical properties based on the latest knowledge in order to minimize the uncertainty of the analysis. The calculated heat transfer behavior across the gap derived from temperature increments in the fuel pin is reasonably consistent to the past studies. This implies that the combinations of the PNC'98 equation with the model are highly predictable for the temperature distribution in the PTM test fuel pins and that the applicability of the PNC'98 equation is verified up to melting point. (author)

  7. Deposition of multi-layer films of hexafluoropropene - ethylene composite polymer with jet-type plasma reactor at atmospheric pressure

    International Nuclear Information System (INIS)

    Multi-layer films of hexafluoropropene - ethylene composite polymer were deposited with a jet-type plasma reactor on poly (ethylene terephthalate) films, which were used as base films, at atmospheric pressure. The multi-layer films were made up by decreasing the flow rate of ethylene gas gradually and increasing that of hexafluoropropene gas simultaneously during the plasma-polymerization. Those films showed low enough peel force, whose value was near that of a Teflon sheet used as a control. Moreover, the bond strength between the multi-layer film and the base film became stronger than that between a plasma-polymerized hexatluoropropene film and the base film. (author)

  8. miRNA-target chimeras reveal miRNA 3'-end pairing as a major determinant of Argonaute target specificity

    DEFF Research Database (Denmark)

    Moore, Michael J; Scheel, Troels K H; Luna, Joseph M;

    2015-01-01

    microRNAs (miRNAs) act as sequence-specific guides for Argonaute (AGO) proteins, which mediate posttranscriptional silencing of target messenger RNAs. Despite their importance in many biological processes, rules governing AGO-miRNA targeting are only partially understood. Here we report a modified...... but divergent 3'-ends. This work provides a means for explicit biochemical identification of miRNA sites in vivo, leading to the discovery that miRNA 3'-end pairing is a general determinant of AGO binding specificity....

  9. Characterization of electrohydrodynamic heat transport components in a space-type nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lipchitz, A.; Harvel, G., E-mail: adam.lipchitz@uoit.ca [Univ. of Ontario Inst. of Tech., Faculty of Energy Systems and Nuclear Sciences, Oshawa, Ontario (Canada)

    2010-07-01

    The paper characterizes and describes the design of a capillary pumped loop with an EHD enhanced evaporator for use in a space reactor. The evaporator uses a wick to transfer the vapour and heat to a vapour section where the vapour is pumped using a EHD gas pump. The vapour is then transferred to a condenser where it is condensed into a liquid and recirculated after the heat has been removed. The design is shown in concept to be an effective method of heat transport in zero and microgravity environments. (author)

  10. N-16 power monitoring system of the RP-10 pool-type reactor

    International Nuclear Information System (INIS)

    The preliminary results of monitoring of power of the RP-10 nuclear reactor by measuring the activity of gamma radiation 16N content in the coolant are presented. A detector NaI(Tl) placed in a window that communicates the decay tank and the pump room of the primary cooling circuit was used. Measurements were performed for different levels of power, from 0,5 to 10 MW. Results show a linear behavior between the power of operation and the activity of 16N. (orig.)

  11. Validation of WIMS-SNAP code systems for calculations in TRIGA-MARK II type reactors

    International Nuclear Information System (INIS)

    The following paper contributes to validate the Nuclear Engineering Department methods to carry out calculations in TRIGA reactors solving a Benchmark. The benchmark is analyzed with the WIMS-D/4-SNAP/3D code system and using the cross section library WIMS-TRIGA. A brief description of the DSN method is presented used in WIMS/d4 code and also the SNAP-3d code is shortly explained. The results are presented and compared with the experimental values. In other hand the possible error sources are analyzed. (author)

  12. A novel electrochemical method for the decontamination of cobalt radionuclides from the primary coolant of PWR-type nuclear reactors

    International Nuclear Information System (INIS)

    In case of intact fuel claddings, the predominant source of radioactivity in the primary circuits of water-cooled nuclear reactors is the activation of corrosion products in the core. The most important corrosion product radionuclides in the primary coolant of pressurized water reactors (PWRs) are 60Co, 58Co, 51Cr, 54Mn, 59Fe (as well as 110mAg in some Soviet-made VVER-type reactor). The present work is focused on the complex studies of the formation and build-up of 60Co-containing species on an austenitic stainless steel type 08X18H10T (GOST 5632-61) often to be used in Soviet-planned VVERs. The kinetics and mechanism of the cobalt accumulation were studied by a combination (coupling) of an in-situ radiotracer method and voltammetry in a model solution of the primary circuit coolant. In addition, the composition and morphology of steel surface as well as the chemical composition of model solution were analyzed by independent techniques like SEM-EDAX and ICP-OES. The experimental results has revealed that (i) The passive behavior of the austenitic stainless steel at open-circuit conditions, the slightly alkaline pH and the reducing water chemistry can be considered to be optimal to minimize the 60Co contamination. (ii) The highly potential dependent deposition of various Co-oxides at E> 1.10 V (vs. RHE) offers a unique possibility to elaborate a novel electrochemical method for the decrease or removal of cobalt traces from borate-containing coolants contaminated with 60Co and/or 58Co radionuclides. (authors)

  13. The differential radiological impact of plutonium recycle in the fuel cycle of LWR type reactors: accidental conditions

    International Nuclear Information System (INIS)

    The radiological impact of the fuel cycle of LWR type reactors using enriched uranium may be changed by plutonium recycle. The differences, which result from accidents which may occur in the different stages of the fuel cycle, are estimated in this study. The differential radiological impact on the population of the European Community is estimated for the recycle of 10t of plutonium metal, taking into consideration some characteristic accidents of each stage of the fuel cycle: fuel fabrication, reactor operation, fuel reprocessing and conversion, and, transport between the different units of the fuel cycle. Each unit is supposed built on an European ''average'' site (mean distributions of the populations and of the agricultural productions, reference meteorological situations). The recycle of plutonium in the fuel cycle involves a few per cent decrease of the radiological impact of the accident choosed for the nuclear power plants. The accidents of transport of plutonium, of new fuels and of plutonium wastes, as also thoses choosed for the fuel fabrication plant involve an increase of the impact for these types of transport and this plant. Finally, the differential radiological impact of the fuel reprocessing plant is positive but low

  14. Significance of coast down time on safety and availability of a pool type fast breeder reactor

    International Nuclear Information System (INIS)

    Highlights: • Plant dynamics studies for quantifying the benefits of flow coast down time. • Establishment of minimum flow coast down time required for safety. • Assessment of influence of flow coast down on enhancing plant availability. • Synthesis of thermo mechanical benefits of flow coast down time on component design. - Abstract: Plant dynamic investigation towards establishing the influence of flow coast down time of primary and secondary sodium systems on safety and availability of plant has been carried out based on one dimensional analysis. From safety considerations, a minimum flow coast down time for primary sodium circuit is essential to be provided to limit the consequences of loss of flow event within allowable limits. Apart from safety benefits, large primary coast down time also improves plant availability by the elimination of reactor SCRAM during short term power failure events. Threshold values of SCRAM parameters also need optimization. By suitably selecting the threshold values for SCRAM parameters, significant reduction in the inertia of pumping systems can be derived to obtain desirable results on plant availability. With the optimization of threshold values and primary flow coast down behaviour equivalent to a halving time of 8 s, there is a possibility to eliminate reactor SCRAM during short term power failure events extending up to 0.75 s duration. Benefits of secondary flow halving on reducing transient thermal loading on components have also been investigated and mixed effects have been observed

  15. The tensile and fatigue properties of type 1.4914 ferritic steel for fusion reactor applications

    International Nuclear Information System (INIS)

    Martensitic steels have received considerable attention as structural materials in fusion reactor applications. In present designs, fusion reactors are expected to operate in a cyclic mode, thus producing cyclic thermal stresses in the first wall. Due to its thermal expansion coefficient and very low swelling rate, 1.4914 martensitic steel is a suitable candidate for the first wall with high neutron loadings. This paper presents the preirradiation results obtained with subsize-specimens designed to be irradiated with a proton beam in the PIREX facility at the Paul Scherrer Institute (PSI) of Wuerenlingen. Both tensile and low cycle fatigue tests were performed in vacuum in the region from 300 K to 870 K (720 K in the case of fatigue tests). Tensile tests on the subsize specimens (0.33 mm thick) compared well to those on bulk specimens, showing a minimum in ductility at around 620 K. The fatigue tests, performed on tubular specimens (3.4 mm external diameter, 0.35 mm wall thickness) showed substantial softening setting in at a low number of cycles. The initial microstructure observed in transmission microscopy consists of fine martensite laths. As cyclic deformation proceeds, dislocation cells form, that gradually replace the martensitic laths. (author) 19 figs., 5 tabs., 16 refs

  16. Single Phase Natural Circulation Behaviors of the Integral Type Marine Reactor Simulator under Rolling Motion Condition

    Directory of Open Access Journals (Sweden)

    Hou-jun Gong

    2015-01-01

    Full Text Available During operation in the sea the reactor natural circulation behaviors are affected by ship rolling motion. The development of an analysis code and the natural circulation behaviors of a reactor simulator under rolling motion are described in this paper. In the case of rolling motion, the primary coolant flow rates in the hot legs and heating channels oscillated periodically, and the amplitude of flow rate oscillation was in direct proportion to rolling amplitude, but in inverse proportion to rolling period. The total mass flow rate also oscillated with half the rolling period, and the average total mass flow rate was less than that in steady state. In the natural circulation under a rolling motion, the flow rate oscillations in the hot legs were controlled by the tangential force; however, the mass flow rate oscillations in the total natural circulation and the heating channels were a result of the combined action of the change of inclination angle, flow resistance, and the extra force arising from the rolling motion. The extra tangential force brought about intense flow rate oscillations in the hot legs, which resulted in increasing total flow resistance; however the extra centrifugal force played a role in increasing thermal driving head.

  17. Sensitivity of k∞ to homogenization, and dimension and composition uncertainties for plate type research reactor fuel

    International Nuclear Information System (INIS)

    As a part of the Dept. of Energy's Innovations in Nuclear Infrastructure and Education (INIE) Program a full facility model is being built of the Ohio State Univ. Research Reactor (OSURR) using the discrete ordinates transport code, PENTRAN. Since the thickness of the fuel plate of OSURR is very small, billions of spatial meshes are required in order to represent the whole reactor core. This, however, is not practical even with the PENTRAN code which is capable of partitioning the memory among processors. Hence, it is essential to consider a certain level of homogenization of fuel, clad, and/or moderator/coolant. Further, since fuel and cladding materials contain impurities and dimensions include tolerances, it also is important to estimate their impacts on the core eigenvalue. The impacts of different homogenization cases as well as the uncertainties in composition and dimensions of the fuel plates on k∞ are examined. To estimate the accuracy of different cases, Monte Carlo reference calculations are performed using the MCNP5 code. The selected combination will be used for full-facility simulation. (authors)

  18. Method for load follow-up operation of PWR type reactor

    International Nuclear Information System (INIS)

    Purpose: To perform load follow-up operation for instant recovery of full power. Method: First, the power of the reactor is lowered by inserting control rods into the reactor core. A subsequent decrease in reactivity by a xenon gas is compensated for by drawing out the control rods, after the control rods have been inserted about 20 to 25 % deep and the axial deviation has reached about 5 to 10 % more on the minus side than the constant value of the full power, the weak absorption control rod banks are drawn out to control while the control rods are held as inserted. Thereafter to raise the power, the control rods and the weak control rod banks are successively drawn out step by step, thereby increasing to the full power. The total reactivity during the load follow-up operation can be changed by the control rod banks and the weak absorption control rod banks, consequently, the change of the critical boron concentration is hadly required to change the total reactivity during load follow-up operation by the control rod banks and the weak absorption control rod banks. (Seki, T.)

  19. Research reactor decommissioning experience - concrete removal and disposal -

    International Nuclear Information System (INIS)

    Removal and disposal of neutron activated concrete from biological shields is the most significant operational task associated with research reactor decommissioning. During the period of 1985 thru 1989 Chem-Nuclear Systems, Inc. was the prime contractor for complete dismantlement and decommissioning of the Northrop TRIGA Mark F, the Virginia Tech Argonaut, and the Michigan State University TRIGA Mark I Reactor Facilities. This paper discusses operational requirements, methods employed, and results of the concrete removal, packaging, transport and disposal operations for these (3) research reactor decommissioning projects. Methods employed for each are compared. Disposal of concrete above and below regulatory release limits for unrestricted use are discussed. This study concludes that activated reactor biological shield concrete can be safely removed and buried under current regulations

  20. Characterization of the biomass of a hybrid anaerobic reactor (HAR) with two types of support material during the treatment of the coffee wastewater

    OpenAIRE

    Vivian Galdino da Silva; Cláudio Milton Montenegro Campos; Erlon Lopes Pereira; Júlia Ferreira da Silva

    2013-01-01

    This study investigated the microbiology of a hybrid anaerobic reactor (HAR) in the removal of pollutant loads. This reactor had the same physical structure of an UASB reactor, however with minifilters inside containing two types of support material: expanded clay and gravel. Two hydraulic retention times (HRT) of 24h and 18h were evaluated at steady-state conditions, resulting in organic loading rates (OLR) of 0.032 and 0.018 kgDBO5m-3d-1 and biological organic loading rates (BOLR) of 0,0015...

  1. Analysis on the `Thermite` reaction consequences in accidents involving research reactors using plate-type fuel; Analisis sobre las concequencias de la reaccion `Termita` en caso de accidentes en reactores de investigacion que utilizan combustible tipo placa

    Energy Technology Data Exchange (ETDEWEB)

    Boero, Norma L.; Bruno, Hernan R.; Camacho, Esteban F.; Cincotta, Daniel O.; Yorio, Daniel [Comision Nacional de Energia Atomica, San Carlos de Bariloche (Argentina). Centro Atomico Constituyentes

    1999-11-01

    The mixture of Al-U{sub 3} O{sub 8} is not in a state of chemical equilibrium, and at temperatures of between 850 deg C and 1000 deg C, it reacts exo thermally. This is known, in corresponding bibliography as a `Thermite reaction. This mixture is used in the manufacturing of the plate-type fuel used in research reactors. It has been pointed out that the release of energy caused by this type of reactions might represent a risk in case of accidents in this type of reactor. Conclusions, in general, tend to indicate that no such risk exists, although no concrete assurance is given that this is the case, and this fact, therefore, leaves room for doubt. The objective of this paper is to provide an in-depth study of what happens to a fuel plate when it is subjected to thermite reaction. We will, furthermore, analyze the consequences of the release of energy generated by this type of reaction within the core of the reactor, clearly defining the problem for this type of fuel and this kind of reactor. (author) 3 refs., 9 figs., 1 tab.

  2. Use of radiotracer for the characterization of anchor-type discontinued reactors

    International Nuclear Information System (INIS)

    The infusion rate of a slug of tracer into an anchor agitated 100 liters batch mixer was characterised by a decay rate constant. This constant was then used to define a dimensionless mixing-rate number which was related to achieve any desired degree of uniformity of the mixture. Tc 99m was used as radiotracer and the mixing process was followed by two scintillation NaI(Ti) sondes situated on the reactor wall at 46 cm from its bottom, one of them near the injection point and the other one in the opposite side. The experimental data was processed by means of a specially for these calculations elaborated program. The programming language was Visual Basic-version 3.0 and Access-version 2.0 was used for data bases making

  3. Irradiations in swimming-pool type reactors from room temperature up to 2000 deg C

    International Nuclear Information System (INIS)

    The irradiations which have been, and are being carried out in the Melusine and Siloe reactors in connection with pure or applied research projects, are effected in widely varying conditions; amongst these, for example, the temperature may vary from -250 deg C to +2000 deg C The eight devices presented are designed for irradiations effected at temperatures of from room temperature up to 2000 deg C. 1. Irradiation device for irradiation at normal temperatures 2. The 'PEF' device 3. The 'CHOUCA' device, 150 to 900 deg C 4. The 'CYRANO' device for EL 4 conditions 5. 'HT' capsules, 800-1000 deg C 6. The 'HEBE' furnace 1400 deg C 7. The 'PEC' device, 1400 deg C 8. The 'HF' furnace 2000 deg C. (authors)

  4. Preliminary steps in partial decommissioning of a swimming pool type reactor Apsara: a health physics experience

    International Nuclear Information System (INIS)

    Full text: Apsara reactor after 50 years of extensive use in production of radio-isotopes, neutron radiography, neutron beam research, shielding experiments etc., is undergoing a partial decommissioning to facilitate refurbishment and up gradation to 2 MWth power using lower enriched Uranium fuel. Partial decommissioning involves defueling and removal of core components as a first step. Radiological safety in defueling is discussed in this paper. Defueling is carried out from top of pile where the radiation level was < 0.10 mR/h, under the strict stipulation that no fuel is to be lifted above water surface during transfer and hence no dose is consumed for this job. Collective dose consumed in the job was only in the SFSB area in Dhruva reactor and was 8.20 person mSv (59% of budgeted dose). This was possible by thorough mock up at Dhruva, SFSB for irradiated fuel handling, satisfying ALARA and refining the procedure. Also radiation mapping of core components, grid plate before and after removal of dry and wet guide tubes were carried out. It was observed that the top portion of grid plate showed a maximum radiation level which was 2-3 times that at bottom portion. The positions around the dry guide tubes G 1 and G 7 showed high radiation levels of 30 R/h. On removing them, the radiation levels reduced to 0.3 -0.5 R/h in all positions. This acted as an input for planning to cut and remove various core components, as also for segregation of high active/low active/inactive components from waste disposal point of view. For example the dry guide tubes G 1 and G 7 was cut and disposed as active waste with upper portion as Cat I and lower portion (was in core region) as Cat II radioactive waste

  5. Stability analysis of a recycling circuit of a BWR type reactor. Theoretical study

    International Nuclear Information System (INIS)

    The Technology, Regulation and Services Management of the National Commission of Nuclear Safety and Safeguards financed and in coordinate form with the I.P.H. Department of the Metropolitan Autonomous-Iztapalapa University developed the present project with the purpose of studying the effect of the recycling system on the linear stability of a BWR reactor whose reference central is the Laguna Verde power station. The present project forms part of a work series focused to the linear stability of the nuclear reactor of the Unit 1 at Laguna Verde power station. The components of the recycling system considered for the study of stability are the recycling external circuit (recycling pumps, valves) and the internal circuit (downcomer, jet pumps, lower full, driers, separators). The mathematical model is obtained applying mass balances and movement quantity in each one of the mentioned circuits. With respect to the nucleus model two regions are considered, the first one is made of a flow in one phase and the second one of a flow in two phases. For modelling the biphasic region it is considered homogenous flow. Generally it is studied the system behavior in the frequency domain starting from the transfer function applied to four operational states which correspond to the lower stability zone in the map power-flow of the Unit 1 of Laguna Verde power station. The Nyquist diagrams corresponding to each state as well as their characteristic frequency were determined. The results show that exists a very clear dependence of the power-flow relation on the stability of the system. It was found that the boiling length is an important parameter for the linear stability of the system. The obtained results show that the characteristic frequencies in unstability zones are similar to the reported data of the Unit 1 of the Laguna Verde power station in the event of power oscillations carried out in January 1995. (Author)

  6. Performance-based improvement of the leakage rate test program for the reactor containment of HTTR. Adoption of revised test programs containing 'Type A, Type B and Type C tests'

    International Nuclear Information System (INIS)

    The reactor containment of HTTR is periodically tested to confirm leak-tight integrity by conducting overall integrated leakage rate tests, so-called 'Type A tests,' in accordance with a standard testing method provided in Japan Electric Association Code (JEAC) 4203. 'Type A test' is identified as a basic one for measuring whole leakage rates for reactor containments, it takes, however, much of cost and time of preparation, implementation and restoration of itself. Therefore, in order to upgrade the maintenance technology of HTTR, the containment leakage rate test program for HTTR was revised by adopting efficient and economical alternatives including Type B and Type C tests' which intend to measure leakage rates for containment penetrations and isolation valves, respectively. In JEAC4203-2004, following requirements are specified for adopting an alternative program: upward trend of the overall integrated leakage rate due to aging affection should not be recognized; performance criterion for combined leakage rate, that is a summation of local leakage rates evaluated by Type B and Type C tests and converted to whole leakage rates, should be established; the criterion of the combined leakage rate should be satisfied as well as of the overall integrated leakage rate; correlation between the overall integrated and combined leakage rates should be recognized. Considering the historical performances, policies of conforming to the forgoing requirements and of carrying out the revised test program were developed, which were accepted by the regulatory agency. This report presents an outline of the leakage rate tests for the reactor containment of HTTR, identifies practical issues of conventional Type A tests, and describes the conforming and implementing policies mentioned above. (author)

  7. High fuel burn-up and nonproliferation in PWR-type reactor on the basis of modified Th-fuel

    International Nuclear Information System (INIS)

    Neutronics-physical characteristics of the fuel lattice of a PWR-type reactor cooled by light water and by a mixture of light and heavy water have been analyzed. Th-fuel containing an essential amount of 231Pa and 232U is used, which allows an increase in fuel burn-up by a factor of 2-5 compared with that of traditional oxide uranium fuel with light water. It is important to underline that this is attained under the negative coolant density reactivity effect using cross sections of 231Pa and 232U from the updated JENDL-3.2 nuclear library. This radical increase of fuel burn-up is accompanied by a small change of reactivity during fuel irradiation (K∞=1.1 / 1.0), that favorably affects safety parameters of the reactor operation. A considerable percentage of 232U in fuel, and consequently in U, is a strong barrier against the proliferation of such weapon nuclide as 233U. (authors)

  8. Analysis of simulation results of damaged nuclear fuel accidents at NPPs with shell-type nuclear reactors

    Directory of Open Access Journals (Sweden)

    Igor L. Kozlov

    2015-03-01

    Full Text Available Lessons from the accident at the Fukushima Daiichi NPP made it necessary to reevaluate and intensificate the work on modeling and analyzing various scenarios of severe accidents with damage to the nuclear fuel in the reactor, containment and spent nuclear fuel storage pool with the expansion of the primary initiating event causes group listing. Further development of computational tools for modeling the explosion prevention criteria as to steam and gas mixtures, considering the specific thermal-hydrodynamic conditions and mechanisms of explosive situations arrival at different stages of a severe accident development, is substantiated. Based on the analysis of the known shell-type nuclear reactors accidents results the explosion safety thermodynamic criteria are presented, the parameters defining the steam and gas explosions conditions are found, the need to perform the further verification and validation of deterministic codes serving to simulate general accident processes behavior as well as phase-to-phase interaction calculated dependencies is established. The main parameters controlling and defining the criteria explosion safety effective regulation areas and their optimization conditions are found.

  9. Dimensional stability of low enriched uranium silicide plate-type fuel for research reactors at transient conditions

    International Nuclear Information System (INIS)

    This paper describes the result of transient experiments using low enriched uranium silicide plate-type fuel for research reactors. The pulse irradiation was carried out at Nuclear Safety Research Reactor (NSRR) in Japan Atomic Energy Research Institute. The results obtained were: (1) At fuel plate temperature of below 400degC, a good dimensional stability of the tested fuel was kept. No fuel failure occurred. (2) At a plate temperature of about 540degC, a local crack was initiated on the Al-3% Mg alloy cladding. Once the cladding temperature exceeded the melting point of 640degC, the fuel plate was degraded much by increased bowing and cracking of the denuded fuel meat occurred after relocation of molten Al cladding. Despite of these degradation, neither fragmentation of the fuel plate nor mechanical energy generation occurred up to the cladding temperature of 971degC. (3) At the temperatures of around 925degC, the reaction of silicide particles with molten Al in the matrix and that of cladding occurred, forming Al riched U (Al, Si) compounds and Si riched (U, Si) compounds at the outermost surface of the silicide particles. (author)

  10. The status of the Bubbler Condenser Containment System for the Reactors of the VVER-440/213 Type

    International Nuclear Information System (INIS)

    VVER-440/213 Pressurized Water Reactors have a pressure-suppression containment structure called a 'Bubbler Condenser' tower which can reduce the design pressure of the entire containment following a design basis accident (DBA), such as a loss-of-coolant accident (LOCA). The bubbler condenser pressure suppression system provides reduction of the LOCA containment pressure by the condensation of released steam in a water pool. World-wide there are 14 nuclear power plants of the VVER-440/213 type in Eastern Europe and Russia. One of the safety concerns for the VVER-440/213 reactors relates to the ability of the bubbler condenser containment system to function satisfactorily and to maintain its integrity following certain postulated accidents and thus limit the release of radioactive material to the environment. The complicated geometry of the bubbler condenser unit, and the dependence on several moving devices and interlocks are the main doubts expressed by different specialists with regard to the design. General description of the bubbler condenser containment system, the physical processes, concerns and design assessment of the bubbler condenser containment system, presentation of the OECD's Unified Bubbler Condenser Research Project (UBCRP) and the European Commission PHARE/TACIS project. Recent utility investigations are also discussed

  11. Effect of gadolinium nitrate concentration on the corrosion compatibility of structural materials in a proposed Indian tube type boiling reactor

    International Nuclear Information System (INIS)

    Gadolinium (Gd3+) is added with nitric acid to moderator heavy water as a neutron poison in nuclear reactors to control reactivity and pH is maintained in the range of 5 to 5.5 to prevent Gd3+ precipitation. Usually ∼15 ppm of Gd3+ is used during actuation of secondary shutdown system and is subsequently removed on ion exchange up to a residual ∼2 ppm before start-up. In the moderator system of a proposed tube type boiling water reactor of Indian origin, a higher concentration (20-400ppm) of Gd(NO3)3 was proposed to be used in the emergency safety shutdown system. With higher concentration of Gd3+, the pH can go down and affect the radiolytic yields and thus affecting the integrity of the structural materials. Considering a long life of 100 years of operation for the proposed reactor, the concentration dependence of Gd3+ on the yields of molecular products like H2 and H2O2 during radiolysis and corrosion compatibility of structural materials like (1) SS 304 LN (proposed structural material for this reactor) and (2) SS 410 (proposed to be used in the valves of the moderator system as an alternative to hard facing alloy, colmonoy) is of interest. The pH and conductivity of the system were observed to be in the range of 5.33-3.76 and 50-870 μS/cm for 20-400 ppm of Gd3+. From the electrochemical studies it was observed that the electrochemical potential increased to more positive potential with increase in Gd3+ concentrations. The yield of H2 and H2O2 was also found to increase with increase in [Gd3+] concentrations. A detailed study on corrosion of the above said alloys at varying [Gd3+] concentrations, temperature, pH and simulated irradiation conditions and its effect on microstructure will be described in the paper. (author)

  12. Heat-transfer characteristics of flowing and stationary particle-bed-type fusion-reactor blankets

    Energy Technology Data Exchange (ETDEWEB)

    Nietert, R.E.

    1983-02-01

    The heat-transfer characteristics of flowing and stationary packed-particle beds have recently become of interest in connection with conceptual designs of fusion reactor blankets. A detailed literature survey has shown that the processes taking place in such beds are not fully understood despite their widespread use in the chemical industry and other engineering disciplines for more than five decades. In this study, two experimental investigations were pursued. In the first, a heat-transfer loop was constructed through which glass microspheres were allowed to flow by rgravity at controlled rates through an electrically heated stainless steel tubular test section. In the second, an annular packed bed was constructed in which heat was applied through the outer wall by electric heating of a stainless steel tube. Cooling occurred at the inner wall of the annular bed by flowing air through the central tube. A second air stream was allowed to flow through the voids of the packed bed. An error-minimization technique was utilized in order to obtain the two-dimensional one-parameter effective conductivity for the bed by comparing the experimental and theoretically predicted temperature profiles. Experiments were conducted for various modified Reynolds numbers less than ten.

  13. Heat-transfer characteristics of flowing and stationary particle-bed-type fusion-reactor blankets

    International Nuclear Information System (INIS)

    The heat-transfer characteristics of flowing and stationary packed-particle beds have recently become of interest in connection with conceptual designs of fusion reactor blankets. A detailed literature survey has shown that the processes taking place in such beds are not fully understood despite their widespread use in the chemical industry and other engineering disciplines for more than five decades. In this study, two experimental investigations were pursued. In the first, a heat-transfer loop was constructed through which glass microspheres were allowed to flow by rgravity at controlled rates through an electrically heated stainless steel tubular test section. In the second, an annular packed bed was constructed in which heat was applied through the outer wall by electric heating of a stainless steel tube. Cooling occurred at the inner wall of the annular bed by flowing air through the central tube. A second air stream was allowed to flow through the voids of the packed bed. An error-minimization technique was utilized in order to obtain the two-dimensional one-parameter effective conductivity for the bed by comparing the experimental and theoretically predicted temperature profiles. Experiments were conducted for various modified Reynolds numbers less than ten

  14. Mounting and joining technology for the dosimetric and control pipelines of WWER-440 type power reactors

    International Nuclear Information System (INIS)

    The development of the mounting and joining technology for the pulse tubes and dosimetric pipelines is described. A GRW type welding machine was chosen for the mechanized welding of austenitic steel pulse tubes. The technology is very effective, high quality of joints can be attained even by workers of low qualification. For the joining of aluminium tubes of the dosimetric pipeline a sticking technology using epoxy adhesive resins has been developed. The AN 134 type resin proved to be very effective. (author)

  15. Probabilistic fracture mechanics. Method and application to the steel containment shell of a PWR-type reactor

    International Nuclear Information System (INIS)

    The thesis is subdivided into two parts. To start off with, the method is shown in general. The question is raised what the purpose of probabilistic fracture mechanics is. In this context some of the problems underlying this method are dealt with. The influence of the fact that curves need to be adapted to rough data on the reliability are of particular interest. Part two of this study describes how a theory developed in the course of part one is applied to a real component, i.e. the steel containment shell of PWR-type reactors. At first, the failure probability of the component is determined. Based on relative statements and a study of parameters, an investigation was carried out to find out what degree the calculated failure probabilities can be relied on. (orig./RW)

  16. Electrochemical incineration of vinasse in filter-press-type FM01-LC reactor using 3D BDD electrode.

    Science.gov (United States)

    Nava, J L; Recéndiz, A; Acosta, J C; González, I

    2008-01-01

    This work shows results obtained in the electrochemical incineration of a synthetic vinasse with initial chemical oxygen demand (COD) of 75.096 g L(-1) in aqueous media (which resembles vinasse industrial wastewater). Electrolyses in a filter-press-type FM01-LC electrochemical reactor equipped with a three-dimensional (3D) boron doped diamond electrode (BDD) were performed at Reynolds values between 22 unity and energy consumption of 168 KW-h m(-3), at Re =109. The mineralization of vinasse indicates that such degradation occurs via hydroxyl radicals formed by the oxidation of water on the BDD surface. Experimental data revealed that hydrodynamic conditions slightly influence the vinasse degradation rate and current efficiency, indicating that the oxidation involves a complex pathway.

  17. Simulation of primary to secondary break in a VVER-type reactor: Results of the IAEA's Third Standard Problem Exercise

    International Nuclear Information System (INIS)

    Seeking to enlarge the experimental data base for code assessment, the International Atomic Energy Agency (IAEA), in collaboration with the Central Research Institute for Physics of the Hungarian Academy of Sciences, has organized the Third Standard Problem Exercise (SPE-3) involving the simulation of a break from primary to secondary in the steam generator of the PMK-NVH experimental facility. The facility is a scaled-down model of a VVER-type reactor, and the experiment addresses the possibility of a break developing in the steam generator collector of the actual plant. This paper presents a brief description of the experimental facility and the experiment. Results of a comparison of pretest and posttest calculations performed by some of the 24 participants in the exercise are also presented. The complete report of the exercise has been published as an IAEA technical document

  18. Failure investigation of type 410 stainless steel retaining block studs of swing check valves in pressurized water reactors

    International Nuclear Information System (INIS)

    Retaining block studs in swing check valves employed in Emergency Core Cooling System of pressurized water reactors have been the subject of cracking during the past few years. The studs are made from ASTM A-193 Gr. 6 type 410 stainless steel material. USNRC issued an information notice on the issue encouraging utilities to take appropriate actions to monitor/replace the studs so as to prevent a potential dislodging of the valve retaining blocks that can cause flow blockage in the valves. The current paper summarizes the results of a metallurgical investigation aimed at establishing the mechanism and cause(s) of the valve stud cracking. Potential candidate replacement materials to remedy the cracking issue are also considered

  19. Optimisation of the flow path in a conceptual pool type reactor under natural circulation with lead coolant

    International Nuclear Information System (INIS)

    This contribution investigates the effects of a bypass flow blocking bottom plate and the influence of the heat transfer between the hot and cold leg in a small pool type reactor cooled through natural convection with lead coolant. The computations are carried out using 3D computational fluid dynamics, where small-detail parts, such as the core and heat exchangers are modeled using a porous media approach. The introduction of full conjugate heat transfer shows that the heat transfer between the hot and cold leg can deteriorate flow in the cold leg and lead to recirculation zones. These zones become even more pronounced with the introduction of a bottom plate, which on the other hand also increases the flow through the core and lowers the maximum temperature in the core by approximately 150 K. Based on the results, redesign suggestions for the bottom plate and the internal wall are made. (author)

  20. Theoretical investigation on the steady-state natural circulation characteristics of a new type of pressurized water reactor

    Institute of Scientific and Technical Information of China (English)

    2006-01-01

    This article presents a theoretical investigation on the steady-state natural circulation characteristics of a new type of pressurized water reactor. Through numerically solving the one-dimensional steady-state single-phase conservative equations for the primary circuit and the steady-state two-phase drift-flux conservative equations for the secondary side of the steam generator, the natural circulation characteristics were studied. On the basis of the preliminary calculation analysis, it was found that natural circulation mass flow rate was proportional to the exponential function of the power and that the value of the exponent is related to the operating conditions of the secondary side of the steam generator. The higher the outlet pressure of the secondary side of the steam generator, the higher the primary natural circulation mass flow rate. The larger height difference between the core center and the steam generator center is favorable for the heat removal capacity of the natural circulation.

  1. MCTP, a code for the thermo-mechanical analysis of a fuel rod of BWR type reactors (Neutron part)

    International Nuclear Information System (INIS)

    In the National Institute of Nuclear Research of Mexico a code for the thermo-mechanical analysis of the fuel rods of the BWR type reactors of the Nucleo electric Central of Laguna Verde is developed. The code solves the diffusion equation in cylindrical coordinates with several energy groups. The code, likewise, calculates the temperature distribution and power distribution in those fuel rods. The code is denominated Multi groups With Temperatures and Power (MCTP). In the code, the energy with which the fission neutrons are emitted it is divided in six groups. They are also considered the produced perturbations by the changes in the temperatures of the materials that constitute the fuel rods, the content of fission products, the uranium consumption and in its case the gadolinium, as well as the plutonium production. In this work there are present preliminary results obtained with the code, using data of operation of the Nucleo electric Central of Laguna Verde. (Author)

  2. Safety analysis for operating the Annular Core Research Reactor with Cintichem-type targets installed in the central region of the core

    Energy Technology Data Exchange (ETDEWEB)

    PARMA JR.,EDWARD J.

    2000-01-01

    Production of the molybdenum-99 isotope at the Annular Core Research Reactor requires highly enriched, uranium oxide loaded targets to be irradiated for several days in the high neutron-flux region of the core. This report presents the safety analysis for the irradiation of up to seven Cintichem-type targets in the central region of the core and compares the results to the Annular Core Research Reactor Safety Analysis Report. A 19 target grid configuration is presented that allows one to seven targets to be irradiated, with the remainder of the grid locations filled with aluminum ''void'' targets. Analyses of reactor, neutronic, thermal hydraulics, and heat transfer calculations are presented. Steady-state operation and accident scenarios are analyzed with the conclusion that the reactor can be operated safely with seven targets in the grid, and no additional risk to the public.

  3. Safety analysis for operating the Annular Core Research Reactor with Cintichem-type targets installed in the central region of the core

    International Nuclear Information System (INIS)

    Production of the molybdenum-99 isotope at the Annular Core Research Reactor requires highly enriched, uranium oxide loaded targets to be irradiated for several days in the high neutron-flux region of the core. This report presents the safety analysis for the irradiation of up to seven Cintichem-type targets in the central region of the core and compares the results to the Annular Core Research Reactor Safety Analysis Report. A 19 target grid configuration is presented that allows one to seven targets to be irradiated, with the remainder of the grid locations filled with aluminum ''void'' targets. Analyses of reactor, neutronic, thermal hydraulics, and heat transfer calculations are presented. Steady-state operation and accident scenarios are analyzed with the conclusion that the reactor can be operated safely with seven targets in the grid, and no additional risk to the public

  4. Characterization of the biomass of a hybrid anaerobic reactor (HAR with two types of support material during the treatment of the coffee wastewater

    Directory of Open Access Journals (Sweden)

    Vivian Galdino da Silva

    2013-06-01

    Full Text Available This study investigated the microbiology of a hybrid anaerobic reactor (HAR in the removal of pollutant loads. This reactor had the same physical structure of an UASB reactor, however with minifilters inside containing two types of support material: expanded clay and gravel. Two hydraulic retention times (HRT of 24h and 18h were evaluated at steady-state conditions, resulting in organic loading rates (OLR of 0.032 and 0.018 kgDBO5m-3d-1 and biological organic loading rates (BOLR of 0,0015 and 0.001 kgDBO5kgSVT- 1d¹, respectively. The decrease in concentration of organic matter in the influent resulted an endogenous state of the biomass in the reactor. The expanded clay was the best support material for biofilm attachment.

  5. Importance and necessity of a joint thermal-physics and thermal-hydraulics data library foundation based on experiments with WWER type reactors

    International Nuclear Information System (INIS)

    At a meeting of CMEA countries experts on WWER type reactor thermal physics in 1980 it was decided to foundate in GDR a joint thermal-physics and thermal-hydraulics data library based upon experiments with CMEA countries NPPs. The stored documents will contain large amount of data on the performed experiments including all the parameters necessary for further calculations, curves of parameter variations with noted measurement accuracy and oth (a detailed parameter list is given). The library structure will be designed in accordance with the following experiment types: natural circulation experiments; power load cut-off experiments; reactor shut-down experiments and oth

  6. Evaluation of the cracking by stress corrosion in nuclear reactor environments type BWR

    International Nuclear Information System (INIS)

    The stress corrosion cracking susceptibility was studied in sensitized, solution annealed 304 steel, and in 304-L welded with a heat treatment that simulated the radiation induced segregation, by the slow strain rate test technique, in a similar environment of a boiling water reactor (BWR), 288 C, 8 MPa, low conductivity and a electrochemical corrosion potential near 200 mV. vs. standard hydrogen electrode (She). The electrochemical noise technique was used for the detection of the initiation and propagation of the cracking. The steels were characterized by metallographic studies with optical and scanning electronic microscopy and by the electrochemical potentiodynamic reactivation of single loop and double loop. In all the cases, the steels present delta ferrite. The slow strain rate tests showed that the 304 steel in the solution annealed condition is susceptible to transgranular stress corrosion cracking (TGSCC), such as in a normalized condition showed granulated. In the sensitized condition the steel showed intergranular stress corrosion cracking, followed by a transition to TGSCC. The electrochemical noise time series showed that is possible associated different time sequences to different modes of cracking and that is possible detect sequentially cracking events, it is means, one after other, supported by the fractographic studies by scanning electron microscopy. The parameter that can distinguish between the different modes of cracking is the re passivation rate, obtained by the current decay rate -n- in the current transients. This is due that the re passivation rate is a function of the microstructure and the sensitization. Other statistic parameters like the localized index, Kurtosis, Skew, produce results that are related with mixed corrosion. (Author)

  7. Expression levels of the microRNA maturing microprocessor complex component DGCR8 and the RNA-induced silencing complex (RISC) components argonaute-1, argonaute-2, PACT, TARBP1, and TARBP2 in epithelial skin cancer.

    Science.gov (United States)

    Sand, Michael; Skrygan, Marina; Georgas, Dimitrios; Arenz, Christoph; Gambichler, Thilo; Sand, Daniel; Altmeyer, Peter; Bechara, Falk G

    2012-11-01

    The microprocessor complex mediates intranuclear biogenesis of precursor microRNAs from the primary microRNA transcript. Extranuclear, mature microRNAs are incorporated into the RNA-induced silencing complex (RISC) before interaction with complementary target mRNA leads to transcriptional repression or cleavage. In this study, we investigated the expression profiles of the microprocessor complex subunit DiGeorge syndrome critical region gene 8 (DGCR8) and the RISC components argonaute-1 (AGO1), argonaute-2 (AGO2), as well as double-stranded RNA-binding proteins PACT, TARBP1, and TARBP2 in epithelial skin cancer and its premalignant stage. Patients with premalignant actinic keratoses (AK, n = 6), basal cell carcinomas (BCC, n = 15), and squamous cell carcinomas (SCC, n = 7) were included in the study. Punch biopsies were harvested from the center of the tumors (lesional), from healthy skin sites (intraindividual controls), and from healthy skin sites in a healthy control group (n = 16; interindividual control). The DGCR8, AGO1, AGO2, PACT, TARBP1, and TARBP2 mRNA expression levels were detected by quantitative real-time reverse transcriptase polymerase chain reaction. The DGCR8, AGO1, AGO2, PACT, and TARBP1 expression levels were significantly higher in the AK, BCC, and SCC groups than the healthy controls (P  0.05). This study indicates that major components of the miRNA pathway, such as the microprocessor complex and RISC, are dysregulated in epithelial skin cancer.

  8. Implications of reactor type and conditions on first-order hydrolysis rate assessment of maize

    NARCIS (Netherlands)

    Pabon Pereira, C.P.; Zeeman, G.; Zhao, R.; Ekmekci, B.; Lier, van J.B.

    2009-01-01

    The biodegradability and first-order hydrolysis coefficient of maize silage have been assessed from batch experiments using different types of inoculum and substrate to inocula (S/I) ratios, and from CSTRs working at different hydraulic retention times (HRTs). In the batch experiments, the assessed

  9. Design analysis of the molten core confinement within the reactor vessel in the case of severe accidents at nuclear power plants equipped with a reactor of the VVER type

    Science.gov (United States)

    Zvonaryov, Yu. A.; Budaev, M. A.; Volchek, A. M.; Gorbaev, V. A.; Zagryazkin, V. N.; Kiselyov, N. P.; Kobzar', V. L.; Konobeev, A. V.; Tsurikov, D. F.

    2013-12-01

    The present paper reports the results of the preliminary design estimate of the behavior of the core melt in vessels of reactors of the VVER-600 and VVER-1300 types (a standard optimized and informative nuclear power unit based on VVER technology—VVER TOI) in the case of beyond-design-basis severe accidents. The basic processes determining the state of the core melt in the reactor vessel are analyzed. The concept of molten core confinement within the vessel based on the idea of outside cooling is discussed. Basic assumptions and models, as well as the results of calculation of the interaction between molten materials of the core and the wall of the reactor vessel performed by means of the SOCRAT severe accident code, are presented and discussed. On the basis of the data obtained, the requirements on the operation of the safety systems are determined, upon the fulfillment of which there will appear potential prerequisites for implementing the concept of the confinement of the core melt within the reactor in cases of severe accidents at nuclear power plants equipped with VVER reactors.

  10. Development of the radiation models of a BWR type reactor and it facility in the SUN-RAH; Desarrollo de modelos de radiacion de un reactor tipo BWR y su instalacion en el SUN-RAH

    Energy Technology Data Exchange (ETDEWEB)

    Barron A, I. [Facultad de Ingenieria, UNAM, 04510 Mexico D.F. (Mexico)]. e-mail: isbarron@yahoo.com.mx

    2005-07-01

    This work about generation models, transport in processes and radioactive contamination of areas of a BWR central, is an amplification to the project developed in the UNAM to have a support tool in subjects or electric generation courses. It is planned about the implementation of models of radiation generation in a BWR type reactor for complement the functions developed in the University Simulator of Nucleo electric- Boiling water reactor (SUN-RAH) which it has been implemented in Simulink of MatLab and it has a model for the dynamics of one nucleo electric central that presents the main characteristics of the reactor vessel, the recirculation system, steam lines, turbines, generator, condensers and feeding water, defined by the main processes that intervene in the generation of energy of these plants. By this way the radiation monitoring systems for area and process, operate simultaneously with the processes of energy generation, with that is possible to observe the changes that present with respect to the operation conditions of the plant, and likewise to appreciate the radiation transport process through the components of the reactor, steam lines and turbines, for different operation conditions and possible faults that they could be presented during the reactor operation. (Author)

  11. Dose calculation for accident situations at TRIGA research reactor using LEU fuel type

    International Nuclear Information System (INIS)

    The 14 MW TRIGA R.R. is a unique design of TRIGA conception. The core was fully converted in May 2006 to use LEU fuel instead of the HEU fuel type. The core contains 29 fuel assemblies, 8 control rods and beryllium reflector, associated instrumentation and controls. The U-235 enrichment for TRIGA - HEU fuel is 93.15 wt % and for TRIGA - LEU is 40.00 wt %. The differences between the two fuel types, as shown by the calculations, will results in a higher core inventory especially for heavy elements (i.e. actinides and transuranium elements), but modifications for noble gases, halogens and other volatile fission products are not so important. Dose calculations for an hypothetical accident scenario was considered and dose and radiological consequence calculations were performed. The results of the calculations and a discussion related on the differences between the consequences in the two cases are also presented. (authors)

  12. A CFD based approach for thermal hydraulic design of main vessel cooling system of pool type fast reactors

    International Nuclear Information System (INIS)

    Highlights: ► We study thermal hydraulic design of main vessel cooling system of fast reactors. ► A CFD based approach is proposed for determination of coolant flow rate. ► Effect of cooling system ovality on temperature asymmetry is quantified. ► Suitable flow distribution device is identified to achieve acceptable flow field. ► To compare efficacy of various devices, a flow mal-distribution index is defined. - Abstract: A computational fluid dynamics (CFDs) based approach is proposed for the thermal hydraulic design of the main vessel cooling system for pool type sodium cooled reactors. Usage of the proposed method is demonstrated by applying it to a future Indian commercial fast breeder reactor. Towards quantifying the amount of sodium flow rate for the main vessel cooling system, two-dimensional CFD investigations have been performed. The conjugate conduction–convection models adopted for this purpose are validated against sodium experiments available in literature. The required flow fraction has been determined to be 2.6% of core flow, which is 175.6 kg/s at full power conditions. The heat loss from the hot pool to the cold pool through the main vessel cooling system is estimated to be 10.6 MW at full power and 3.7 MW at 20% power conditions. By detailed three-dimensional CFD studies, the effect of ovality in the main vessel cooling annuli due to manufacturing tolerances has been assessed and the associated circumferential temperature difference in the main vessel is determined to be 14 °C, which is less than the permissible upper limit of 30 °C. The uniformity of sodium flow in the cooling annulus has been investigated by a three-dimensional hydraulic analysis with a view to identify a suitable passive device that can render a uniform velocity distribution. To compare the effectiveness of various devices, a flow mal-distribution index is defined. Detailed parametric studies have been carried out to identify an appropriate porous jet breaker

  13. Assessment of susceptibility of Type 304 stainless steel to intergranular stress corrosion cracking in simulated Savannah River Reactor environments

    International Nuclear Information System (INIS)

    Intergranular stress corrosion cracking (IGSCC) of Type 304 stainless steel rate tests (CERT) of specimens machined was evaluated by constant extension from Savannah River Plant (SRP) decontaminated process water piping. Results from 12 preliminary CERT tests verified that IGSCC occurred over a wide range of simulated SRP envirorments. 73 specimens were tested in two statistical experimental designs of the central composite class. In one design, testing was done in environments containing hydrogen peroxide; in the other design, hydrogen peroxide was omitted but oxygen was added to the environment. Prediction equations relating IGSCC to temperature and environmental variables were formulated. Temperature was the most important independent variable. IGSCC was severe at 100 to 120C and a threshold temperature between 40C and 55C was identified below which IGSCC did not occur. In environments containing hydrogen peroxide, as in SRP operation, a reduction in chloride concentration from 30 to 2 ppB also significantly reduced IGSCC. Reduction in sulfate concentration from 50 to 7 ppB was effective in reducing IGSCC provided the chloride concentration was 30 ppB or less and temperature was 95C or higher. Presence of hydrogen peroxide in the environment increased IGSCC except when chloride concentration was 11 ppB or less. Actual concentrations of hydrogen peroxide, oxygen and carbon dioxide did not affect IGSCC. Large positive ECP values (+450 to +750 mV Standard Hydrogen Electrode (SHE)) in simulated SRP environments containing hydrogen peroxide and were good agreement with ECP measurements made in SRP reactors, indicating that the simulated environments are representative of SRP reactor environments. Overall CERT results suggest that the most effective method to reduce IGSCC is to reduce chloride and sulfate concentrations

  14. Evaluation of plate type fuel options for small power reactors; Avaliacao de alternativas de combustivel tipo placa para reatores de pequeno porte

    Energy Technology Data Exchange (ETDEWEB)

    Andrzejewski, Claudio de Sa

    2005-07-01

    Plate type fuels are generally used in research reactor. The utilization of this kind of configuration improves significantly the overall performance fuel. The conception of new fuels for small power reactors based in plate-type configuration needs a complete review of the safety criteria originally used to conduce power and research reactor projects. In this work, a group of safety criteria is established for the utilization of plate-type fuels in small power reactors taking into consideration the characteristics of power and research reactors. The performance characteristics of fuel elements are strongly supported by its materials properties and the adopted configuration for its fissile particles. The present work makes an orientated bibliographic investigation searching the best material properties (structural materials and fuel compounds) related to the performance fuel. Looking for good parafermionic characteristics and manufacturing exequibility associated to existing facilities in national research centres, this work proposes several alternatives of plate type fuels, considering its utilization in small power reactors: dispersions of UO{sub 2} in stainless steel, of UO{sub 2} in zircaloy, and of U-Mo alloy in zircaloy, and monolithic plates of U-Mo cladded with zircaloy. Given the strong dependency of radiation damage with temperature increase, the safety criteria related to heat transfer were verified for all the alternatives, namely the DNBR; coolant temperature lower than saturation temperature; peak meat temperature to avoid swelling; peak fuel temperature to avoid meat-matrix reaction. It was found that all alternatives meet the safety criteria including the 0.5 mm monolithic U-Mo plate cladded with zircaloy. (author)

  15. NUCLEAR REACTOR

    Science.gov (United States)

    Miller, H.I.; Smith, R.C.

    1958-01-21

    This patent relates to nuclear reactors of the type which use a liquid fuel, such as a solution of uranyl sulfate in ordinary water which acts as the moderator. The reactor is comprised of a spherical vessel having a diameter of about 12 inches substantially surrounded by a reflector of beryllium oxide. Conventionnl control rods and safety rods are operated in slots in the reflector outside the vessel to control the operation of the reactor. An additional means for increasing the safety factor of the reactor by raising the ratio of delayed neutrons to prompt neutrons, is provided and consists of a soluble sulfate salt of beryllium dissolved in the liquid fuel in the proper proportion to obtain the result desired.

  16. A comparative study of charcoal gasification in two types of spouted bed reactors

    Energy Technology Data Exchange (ETDEWEB)

    Abdul Salam, P. [Energy Field of Study, School of Environment, Resources and Development, Asian Institute of Technology, P.O. Box 4, Klongluang, Pathumthani 12120 (Thailand); Bhattacharya, S.C. [Energy Field of Study, School of Environment, Resources and Development, Asian Institute of Technology, P.O. Box 4, Klongluang, Pathumthani 12120 (Thailand)] e-mail: bhatta@ait.ac.th

    2006-03-01

    Gasification is considered to be a favourable method for converting a solid fuel into a more versatile gaseous fuel. Performance of a gasifier depends on the design of the gasifier, type of fuel used and air flow rate, etc. The applications of spouted bed for a variety of processes such as drying, coating, pyrolysis, gasification and combustion have been reported. Gasification of solid fuels in a spouted bed, which has certain potential advantages over other fluid bed configurations, appears to be an under-exploited technique so far. Central jet distributors are the most commonly used in the experimental studies that has been reported in the literature. Circular slit distributor is a new concept. This paper presents results of a comparative experimental study on air gasification of charcoal in central jet and circular slit inert sand spouted beds. The experiments were carried for an equivalence ratio of 0.25. The effect of spouting velocity and type of the distributor on the gasification performance were discussed. The steady state dense bed temperature varied between 979 and 1183 deg C for central jet spouted bed and between 964 and 1235 deg C for circular slit spouted bed. At higher spouting velocities, the gasification efficiency of the circular slit spouted bed was slightly more compared with that of central jet spouted bed.

  17. A comparative study of charcoal gasification in two types of spouted bed reactors

    Energy Technology Data Exchange (ETDEWEB)

    Salam, P.A.; Bhattacharya, S.C. [Asian Institute of Technology, Pathumthani (Thailand). School of Environmental, Resources and Development

    2006-03-01

    Gasification is considered to be a favourable method for converting a solid fuel into a more versatile gaseous fuel. Performance of a gasifier depends on the design of the gasifier, type of fuel used and airflow rate, etc. The applications of spouted bed for a variety of processes such as drying, coating, pyrolysis, gasification and combustion have been reported. Gasification of solid fuels in a spouted bed, which has certain potential advantages over other fluid bed configurations, appears to be an under-exploited technique so far. Central jet distributors are the most commonly used in the experimental studies that has been reported in the literature. Circular slit distributor is a new concept. This paper presents results of a comparative experimental study on air gasification of charcoal in central jet and circular slit inert sand spouted beds. The experiments were carried for an equivalence ratio of 0.25. The effect of spouting velocity and type of the distributor on the gasification performance were discussed. The steady state dense bed temperature varied between 979 and 1183{sup o}C for central jet spouted bed and between 964 and 1235{sup o}C for circular slit spouted bed. At higher spouting velocities, the gasification efficiency of the circular slit spouted bed was slightly more compared with that of central jet spouted bed. (author)

  18. Development of Pool-type Sodium-cooled Fast Reactor System Analysis Code%池式钠冷快堆系统分析程序开发

    Institute of Scientific and Technical Information of China (English)

    王晋; 张东辉; 胡文军

    2016-01-01

    针对池式钠冷快堆的特点,在对快堆系统的水力模型、热工模型和中子动力学模型进行详细分类和建模的基础上,利用 FORTRAN95语言开发了可用于池式钠冷快堆事故分析的系统分析程序(FASYS程序)。以中国实验快堆为计算对象对FASYS程序模型进行了初步验证,所获得的结果和试验值与其他系统程序计算值符合良好,证明了所开发的系统分析程序的正确性。%According to the characteristics of pool‐type sodium‐cooled fast reactor ,and with the fast reactor hydraulic model , thermal model and neutron kinetics model thoroughly classified and developed ,a fast reactor system analysis code (FASYS code) was developed by FORTRAN95 language for pool‐type sodium‐cooled fast reactor acci‐dent analysis .Transient conditions in CEFR were calculated with FASYS code and the results were used for code validation .The calculation results are consistent with the test data and other fast reactor system analysis code results , and the correctness of the FASYS code is proved .

  19. Compact Reactor

    International Nuclear Information System (INIS)

    Weyl's Gauge Principle of 1929 has been used to establish Weyl's Quantum Principle (WQP) that requires that the Weyl scale factor should be unity. It has been shown that the WQP requires the following: quantum mechanics must be used to determine system states; the electrostatic potential must be non-singular and quantified; interactions between particles with different electric charges (i.e. electron and proton) do not obey Newton's Third Law at sub-nuclear separations, and nuclear particles may be much different than expected using the standard model. The above WQP requirements lead to a potential fusion reactor wherein deuterium nuclei are preferentially fused into helium nuclei. Because the deuterium nuclei are preferentially fused into helium nuclei at temperatures and energies lower than specified by the standard model there is no harmful radiation as a byproduct of this fusion process. Therefore, a reactor using this reaction does not need any shielding to contain such radiation. The energy released from each reaction and the absence of shielding makes the deuterium-plus-deuterium-to-helium (DDH) reactor very compact when compared to other reactors, both fission and fusion types. Moreover, the potential energy output per reactor weight and the absence of harmful radiation makes the DDH reactor an ideal candidate for space power. The logic is summarized by which the WQP requires the above conditions that make the prediction of DDH possible. The details of the DDH reaction will be presented along with the specifics of why the DDH reactor may be made to cause two deuterium nuclei to preferentially fuse to a helium nucleus. The presentation will also indicate the calculations needed to predict the reactor temperature as a function of fuel loading, reactor size, and desired output and will include the progress achieved to date

  20. Applied model of through-wall crack of coolant vessels of WWER-type reactors

    International Nuclear Information System (INIS)

    We propose an applied-model of Through-Wall Crack (TWC) for WWER-type units primary vessels. The model allows to simulate the main morphological parameters of real TWC, i.e. length, area of inlet and outlet openings, channel depth and small and large size unevenness of the crack surface. The model can be used for developing and improving the coolant-leak detectors for the primary circuit vessels of WWER-units. Also, it can be used for research of the coolant two-phase leakage phenomenon through narrow cracks/channels and thermo-physical processes in heat-insulation layer of the Main Coolant Piping (MCP) during the leak

  1. Are fast explorers slow reactors? Linking personality type and anti-predator behaviour.

    Science.gov (United States)

    Jones, Katherine A; Godin, Jean-Guy J

    2010-02-22

    Response delays to predator attack may be adaptive, suggesting that latency to respond does not always reflect predator detection time, but can be a decision based on starvation-predation risk trade-offs. In birds, some anti-predator behaviours have been shown to be correlated with personality traits such as activity level and exploration. Here, we tested for a correlation between exploration behaviour and response latency time to a simulated fish predator attack in a fish species, juvenile convict cichlids (Amatitlania nigrofasciata). Individual focal fish were subjected to a standardized attack by a robotic fish predator while foraging, and separately given two repeated trials of exploration of a novel environment. We found a strong positive correlation between exploration and time taken to respond to the predator model. Fish that were fast to explore the novel environment were slower to respond to the predator. Our study therefore provides some of the first experimental evidence for a link between exploration behaviour and predator-escape behaviour. We suggest that different behavioural types may differ in how they partition their attention between foraging and anti-predator vigilance.

  2. Optimization design study of an innovative divertor concept for future experimental tokamak-type fusion reactors

    International Nuclear Information System (INIS)

    The design optimization study of an innovative divertor concept for future experimental tokamak-type fusion devices is both an answer to the actual problems encountered in the multilayer divertor proposals and an illustration of a rational modelling philosophy and optimization strategy for the development of a new divertor structure. Instead of using mechanical attachment or metallurgical bonding of the protective material to the heat sink as in most actual divertor concepts, the so-called brush divertor in this study uses an array of unidirectional fibers penetrating in both the protective armor and the underling composite heat sink. Although the approach is fully concentrated on the divertor performance, including both a description of its function from the theoretical point of view and an overview of the problems related to the materials choice and evaluation, both the approach followed in the numerical modelling and the judgment of the results are thought to be valid also for other applications. Therefore the spin-off of the study must be situated in both the technological progress towards a feasible divertor solution, which introduces no additional physical uncertainties, and in the general area of the thermo-mechanical finite-element modelling on both macro-and microscale. The brush divertor itself embodies the use, and thus the modelling, of advanced materials such as tailor-made metal matrix composites and dispersion strengthened metals, and is shown to offer large potential advantages, demanding however and experimental validation under working conditions. It is clearly indicated where the need originates for an integrated experimental program which must allow to verify the basic modelling assumptions in order to arrive at the use of numerical computation as a powerful and realistic tool of structural testing and life-time prediction

  3. Nondestructive testing of PWR type fuel rods by eddy currents and metrology in the OSIRIS reactor pool

    International Nuclear Information System (INIS)

    The Saclay Reactor Department has developed a nondestructive test bench, now installed above channel 1 of the OSIRIS reactor. As part of investigations into the dynamics of PWR fuel degradation, a number of fuel rods underwent metrological and eddy current inspection, after irradiation

  4. Commercial-Scale Performance Predictions for High-Temperature Electrolysis Plants Coupled to Three Advanced Reactor Types

    Energy Technology Data Exchange (ETDEWEB)

    M. G. McKellar; J. E. O' Brien; J. S. Herring

    2007-09-01

    This report presents results of system analyses that have been developed to assess the hydrogen production performance of commercial-scale high-temperature electrolysis (HTE) plants driven by three different advanced reactor – power-cycle combinations: a high-temperature helium cooled reactor coupled to a direct Brayton power cycle, a supercritical CO2-cooled reactor coupled to a direct recompression cycle, and a sodium-cooled fast reactor coupled to a Rankine cycle. The system analyses were performed using UniSim software. The work described in this report represents a refinement of previous analyses in that the process flow diagrams include realistic representations of the three advanced reactors directly coupled to the power cycles and integrated with the high-temperature electrolysis process loops. In addition, this report includes parametric studies in which the performance of each HTE concept is determined over a wide range of operating conditions. Results of the study indicate that overall thermal-to- hydrogen production efficiencies (based on the low heating value of the produced hydrogen) in the 45 - 50% range can be achieved at reasonable production rates with the high-temperature helium cooled reactor concept, 42 - 44% with the supercritical CO2-cooled reactor and about 33 - 34% with the sodium-cooled reactor.

  5. Analysis on tritium management in FLiBe blanket for LHD-type helical reactor FFHR2

    International Nuclear Information System (INIS)

    In FFHR2 (LHD-type helical reactor) design, FLiBe has been selected as a self-cooling tritium breeder for low reactivity with oxygen and water and lower conductivity. Considering the fugacity of the tritium, particular care and adequate mitigation measures should be applied for the effectively extracting tritium from breeder and controlling the tritium release to the environment. In this paper, a tritium analysis model of the FLiBe blanket system was developed and the preliminary analysis on tritium permeation and extraction for FLiBe blanket system were done. The results of the analysis showed that it was reasonable to select W alloy as heat exchanger (HX) material, the proportion of FLiBe flow in tritium recover system (TRS) was 0.2, the efficiency of TRS was 0.85 and tritium permeation reduction factor (TPRF) was 20 in blanket etc.. In addition, further R and D efforts were required for FFHR2 tritium system to guarantee the tritium self-sufficient and safety, for example reasonable quality of tritium permeation barriers on blanket, requirement for the TRS and fabrication technology of the heat exchanger etc.. (author)

  6. In-situ combustion of heavy crude oil in a consolidate porous media using a disc type reactor

    International Nuclear Information System (INIS)

    Thermal recovery is the most mature technology available for (EOR) Enhanced Oil Recovery). Thermal processes are either heat-injection processes or in-situ heat generation processes, and are particularly applicable to the recovery of heavy oils and tar sands. In-situ heat generation processes require combustion within the reservoir. These processes are categorized as either forward combustion, reverse combustion, or wet combustion. A high pressure disc type reactor has been used to study the effect of the main variables thought to affect the kinetics of heavy oil in a consolidated porous media on combustion. The processes observed were very dependent on the conditions used. The rate of oxygen consumption and carbon burning rate increased when the air flux is increased. This behavior is attributed to the overall increase in the rate of oxidation reactions. When the heating rate was reduced from 5 degree C min/sup -1/ to 1.5 degree C min/sup-1/ with constant inlet flux, LTO (Low Temperature Oxidation) reactions were found more significant. In addition, the rate of fuel burning reaction HTO (High Temperature Oxidation) decreased with decreased heating rate. The change in the rate of the fuel burning reaction is attributed to the change in the nature of the fuel burnt in a consolidated formation. (author)

  7. Improvement of hydrogen production via ethanol-type fermentation in an anaerobic down-flow structured bed reactor.

    Science.gov (United States)

    Anzola-Rojas, Mélida del Pilar; Zaiat, Marcelo; De Wever, Heleen

    2016-02-01

    Although a novel anaerobic down-flow structured bed reactor has shown feasibility and stable performance for a long-term compared to other anaerobic fixed bed systems for continuous hydrogen production, the volumetric rates and yields have so far been too low. In order to improve the performance, an operation strategy was applied by organic loading rate (OLR) variation (12-96 g COD L(-1) d(-1)). Different volumetric hydrogen rates, and yields at the same OLR indicated that the system was mainly driven by the specific organic load (SOL). When SOL was kept between 3.8 and 6.2 g sucrose g(-1) VSS d(-1), the volumetric rates raised from 0.1 to 8.9 L H2 L(-1) d(-1), and the yields were stable around 2.0 mol H2 mol(-1) converted sucrose. Furthermore, hydrogen was produced mainly via ethanol-type fermentation, reaching a total energy conversion rate of 23.40 kJ h(-1) L(-1) based on both hydrogen and ethanol production.

  8. The Argonaute CSR-1 and its 22G-RNA co-factors target germline genes and are required for holocentric chromosome segregation

    OpenAIRE

    Claycomb, Julie M.; Batista, Pedro J; Pang, Ka Ming; Gu, Weifeng; Vasale, Jessica J.; van Wolfswinkel, Josien C.; Chaves, Daniel A.; Shirayama, Masaki; Mitani, Shohei; Ketting, René F.; Conte, Darryl; Mello, Craig C

    2009-01-01

    RNAi-related pathways regulate diverse processes, from developmental timing to transposon silencing. Here, we show that in C. elegans the Argonaute CSR-1, the RNA-dependent RNA polymerase EGO-1, the Dicer-related helicase DRH-3, and the Tudor-domain protein EKL-1 localize to chromosomes and are required for proper chromosome segregation. In the absence of these factors chromosomes fail to align at the metaphase plate and kinetochores do not orient to opposing spindle poles. Surprisingly, the ...

  9. Field of the application and general project decisions of atomic thermal power plant ATETS-80 based on integral WWER type reactor

    International Nuclear Information System (INIS)

    Design Bureau of Machine Building is gradually being developed the direction of improvement WWER type reactors of the small and medium power for the production of heat and electricity. It is presented the integral WWER energetic reactor ATETS-80 in composition of thermal power plant, destined for the combined processing of electric power, steam, hot water, desalination of marine and pickling water. Electric power of the bloc ATETS-80 equal to 85 MWt, the heat power equal to 250 MWt. The project ATETS-80 of the promoted safety worked out on base of common project decisions, established and worked for reactor installations WWER, atomic icebreakers and general atomic plants of heat deliveries. General project decisions: integral accomplishment of the reactor; ramjet steam generator with superheating; two loops of exchange heat with channels of ECCS; autonomous channel ECCS on reactor; insurance vessel, discharged to full emergency pressure in case of the rupture of first contour; the containment shell, provided the protection from external effects. 2 figs

  10. Research Nuclear Reactors

    International Nuclear Information System (INIS)

    Published in English and in French, this large report first proposes an overview of the use and history of research nuclear reactors. It discusses their definition, and presents the various types of research reactors which can be either related to nuclear power (critical mock-ups, material test reactors, safety test reactors, training reactors, prototypes), or to research (basic research, industry, health), or to specific particle physics phenomena (neutron diffraction, isotope production, neutron activation, neutron radiography, semiconductor doping). It reports the history of the French research reactors by distinguishing the first atomic pile (ZOE), and the activities and achievements during the fifties, the sixties and the seventies. It also addresses the development of instrumentation for research reactors (neutron, thermal, mechanical and fission gas release measurements). The other parts of the report concern the validation of neutronics calculations for different reactors (the EOLE water critical mock-up, the MASURCA air critical mock-up dedicated to fast neutron reactor study, the MINERVE water critical mock-up, the CALIBAN pulsed research reactor), the testing of materials under irradiation (OSIRIS reactor, laboratories associated with research reactors, the Jules Horowitz reactor and its experimental programs and related devices, irradiation of materials with ion beams), the investigation of accident situations (on the CABRI, Phebus, Silene and Jules Horowitz reactors). The last part proposes a worldwide overview of research reactors

  11. Contribution to the study of the evolution of nuclear fuel composition in PWR type reactors. Reactor cores in three and four regions of enrichment

    International Nuclear Information System (INIS)

    The main methods for calculations of fuel depletion are studied and some approaches to do it are mentioned; the LEOPARD Code is described and full details are given for each subroutine, flow charts are included; the method given by the code for calculation of fuel depletion is described; some imperfections from the IPR's version are listed, and corrected, for instance: the method for burn-up calculation of heavy isotopes; the results of calculations for a reference reactor based on data of the Preliminary Safety Analysis Report (PSAR) for Angra I Nuclear Plant are presented and discussed. (author)

  12. Design of an equilibrium nucleus of a BWR type reactor based in a Thorium-Uranium fuel; Diseno de un nucleo de equilibrio de un reactor tipo BWR basado en un combustible de Torio-Uranio

    Energy Technology Data Exchange (ETDEWEB)

    Francois, J.L.; Nunez C, A. [Laboratorio de Analisis en Ingenieria de Reactores Nucleares, Facultad de Ingenieria-UNAM, Paseo Cuauhnahuac 8532, Jiutepec, Morelos (Mexico)

    2003-07-01

    In this work the design of the reactor nucleus of boiling water using fuel of thorium-uranium is presented. Starting from an integral concept based in a type cover-seed assemble is carried out the design of an equilibrium reload for the nucleus of a reactor like that of the Laguna Verde Central and its are analyzed some of the main design variables like the cycle length, the reload fraction, the burnt fuel, the vacuum distribution, the generation of lineal heat, the margin of shutdown, as well as a first estimation of the fuel cost. The results show that it is feasible to obtain an equilibrium reload, comparable to those that are carried out in the Laguna Verde reactors, with a good behavior of those analyzed variables. The cost of the equilibrium reload designed with the thorium-uranium fuel is approximately 2% high that the uranium reload producing the same energy. It is concluded that it is convenient to include burnable poisons, type gadolinium, in the fuel with the end of improving the reload design, the fuel costs and the margin of shutdown. (Author)

  13. An Innovative Passive Residual Heat Removal System of an Open-Pool Type Research Reactor with Pump Flywheel and Gravity Core Cooling Tank

    Directory of Open Access Journals (Sweden)

    Kwon-Yeong Lee

    2015-01-01

    Full Text Available In an open-pool type research reactor, the primary cooling system can be designed to have a downward flow inside the core during normal operation because of the plate type fuel geometry. There is a flow inversion inside the core from the downward flow by the inertia force of the primary coolant to the upward flow by the natural circulation when the pump is turned off. To delay the flow inversion time, an innovative passive system with pump flywheel and GCCT is developed to remove the residual heat. Before the primary cooling pump starts up, the water level of the GCCT is the same as that of the reactor pool. During the primary cooling pump operation, the water in the GCCT is moved into the reactor pool because of the pump suction head. After the pump stops, the potential head generates a downward flow inside the core by moving the water from the reactor pool to the GCCT and removes the residual heat. When the water levels of the two pools are the same again, the core flow has an inversion of the flow direction, and natural circulation is developed through the flap valves.

  14. Study the effects of different reflector types on the neutronic parameters of the 10 MW MTR reactor using the MCNP4C code

    International Nuclear Information System (INIS)

    Highlights: • A 3-D neutronic model for the 10 MW MTR has been conducted using the MCNP4C code. • Studying the effect of different reflectors on the neutronics parameters of the reactor. • Beryllium reflector was found to be the most efficient reflector among the studied reflectors. • The graphite reflector gave the highest maximum thermal neutron flux in the water trap. - Abstract: A 3-D neutronic model for the 10 MW MTR research reactor has been conducted for the HEU (93%), MEU (45%) and LEU (20%) fuels using the MCNP4C code. This model has been used to study the effect of different types of reflector materials on the reactor multiplication factor and neutron flux distribution in the reactor. It was found that the beryllium reflector was the most efficient reflector among the studied reflector groups (beryllium, heavy water, graphite and light water) since it gave the highest reactor multiplication factor, 1.21441. It followed by heavy water, graphite and light water with the following reactor multiplication factors: 1.19458, 1.19287 and 1.16867 respectively. The graphite reflector gave the highest maximum thermal neutron flux in the water trap, 2.576E14 n cm−2 s−1. It followed by heavy water, light water, and beryllium with the following results: 2.533E14, 2.526E14 and 2.525E14 n cm−2 s−1 respectively. Considerable gains in reactivity were not appreciably influenced by changing the fuel enrichment

  15. Safeguarding research reactors

    International Nuclear Information System (INIS)

    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

  16. Design and tuning of a Decentralized Fuzzy Logic Controller for a MIMO type Pressurized Heavy Water Reactor

    International Nuclear Information System (INIS)

    Highlights: • A 14 inference modules based DFLC is designed for 70th order MIMO PHWR system. • Auto tuning of DFLC for PHWR is performed using NMA. • A novel approach is presented to overcome the shortcomings of NMA in tuning the DFLC. • The optimally tuned DFLC is evaluated for robustness and reference tracking capabilities. - Abstract: A Pressurized Heavy Water Reactor (PHWR) is a highly complex and unstable system. Designing a safe, reliable and robust controller with good performance for such a large and complex system is an important control engineering problem. In this work, a Decentralized Fuzzy Logic Controller (DFLC) with 140 input and 70 output membership functions, is designed for a 70th order Multi-Input Multi-Output (MIMO) type PHWR. In order to obtain high performance of the controller, it needs to be tuned optimally, however, it is very challenging task to optimally tune the DFLC with such a large membership functions. Moreover, PHWR is a coupled system which imposes additional limitation in tuning the controller since the output of one PHWR’s zone affects the outputs of other zones. In this work, an application of Nelder–Mead Algorithm (NMA) is presented for auto tuning the DFLC. The NMA performance depends upon objective function and initial points given to the NMA at the start of the tuning process. A novel method for selecting the optimal objective function and initial points for the NMA is also proposed since their selection is another complicated process. Although several objective functions have been proposed by the researchers for use with NMA, this work focuses five common indices (IAE, ISE, ITAE, ITSE and ISTE) as objective functions, which are simple and system independent. Finally, the optimally tuned high-performance DFLC is applied to the PHWR and evaluated by simulating different scenarios. The simulation results show that the controller is efficient, fast and robust and ensures the safety and reliability of the PHWR

  17. Successful scaling-up of self-sustained pyrolysis of oil palm biomass under pool-type reactor.

    Science.gov (United States)

    Idris, Juferi; Shirai, Yoshihito; Andou, Yoshito; Mohd Ali, Ahmad Amiruddin; Othman, Mohd Ridzuan; Ibrahim, Izzudin; Yamamoto, Akio; Yasuda, Nobuhiko; Hassan, Mohd Ali

    2016-02-01

    An appropriate technology for waste utilisation, especially for a large amount of abundant pressed-shredded oil palm empty fruit bunch (OFEFB), is important for the oil palm industry. Self-sustained pyrolysis, whereby oil palm biomass was combusted by itself to provide the heat for pyrolysis without an electrical heater, is more preferable owing to its simplicity, ease of operation and low energy requirement. In this study, biochar production under self-sustained pyrolysis of oil palm biomass in the form of oil palm empty fruit bunch was tested in a 3-t large-scale pool-type reactor. During the pyrolysis process, the biomass was loaded layer by layer when the smoke appeared on the top, to minimise the entrance of oxygen. This method had significantly increased the yield of biochar. In our previous report, we have tested on a 30-kg pilot-scale capacity under self-sustained pyrolysis and found that the higher heating value (HHV) obtained was 22.6-24.7 MJ kg(-1) with a 23.5%-25.0% yield. In this scaled-up study, a 3-t large-scale procedure produced HHV of 22.0-24.3 MJ kg(-1) with a 30%-34% yield based on a wet-weight basis. The maximum self-sustained pyrolysis temperature for the large-scale procedure can reach between 600 °C and 700 °C. We concluded that large-scale biochar production under self-sustained pyrolysis was successfully conducted owing to the comparable biochar produced, compared with medium-scale and other studies with an electrical heating element, making it an appropriate technology for waste utilisation, particularly for the oil palm industry.

  18. Successful scaling-up of self-sustained pyrolysis of oil palm biomass under pool-type reactor.

    Science.gov (United States)

    Idris, Juferi; Shirai, Yoshihito; Andou, Yoshito; Mohd Ali, Ahmad Amiruddin; Othman, Mohd Ridzuan; Ibrahim, Izzudin; Yamamoto, Akio; Yasuda, Nobuhiko; Hassan, Mohd Ali

    2016-02-01

    An appropriate technology for waste utilisation, especially for a large amount of abundant pressed-shredded oil palm empty fruit bunch (OFEFB), is important for the oil palm industry. Self-sustained pyrolysis, whereby oil palm biomass was combusted by itself to provide the heat for pyrolysis without an electrical heater, is more preferable owing to its simplicity, ease of operation and low energy requirement. In this study, biochar production under self-sustained pyrolysis of oil palm biomass in the form of oil palm empty fruit bunch was tested in a 3-t large-scale pool-type reactor. During the pyrolysis process, the biomass was loaded layer by layer when the smoke appeared on the top, to minimise the entrance of oxygen. This method had significantly increased the yield of biochar. In our previous report, we have tested on a 30-kg pilot-scale capacity under self-sustained pyrolysis and found that the higher heating value (HHV) obtained was 22.6-24.7 MJ kg(-1) with a 23.5%-25.0% yield. In this scaled-up study, a 3-t large-scale procedure produced HHV of 22.0-24.3 MJ kg(-1) with a 30%-34% yield based on a wet-weight basis. The maximum self-sustained pyrolysis temperature for the large-scale procedure can reach between 600 °C and 700 °C. We concluded that large-scale biochar production under self-sustained pyrolysis was successfully conducted owing to the comparable biochar produced, compared with medium-scale and other studies with an electrical heating element, making it an appropriate technology for waste utilisation, particularly for the oil palm industry. PMID:26612557

  19. A neural networks based ``trip`` analysis system for PWR-type reactors; Um sistema de analise de ``trip`` em reatores PWR usando redes neuronais

    Energy Technology Data Exchange (ETDEWEB)

    Alves, Antonio Carlos Pinto Dias

    1993-12-31

    The analysis short after automatic shutdown (trip) of a PWR-type nuclear reactor takes a considerable amount of time, not only because of the great number of variables involved in transients, but also the various equipment that compose a reactor of this kind. On the other hand, the transients`inter-relationship, intended to the detection of the type of the accident is an arduous task, since some of these accidents (like loss of FEEDWATER and station BLACKOUT, for example), generate transients similar in behavior (as cold leg temperature and steam generators mixture levels, for example). Also, the sequence-of-events analysis is not always sufficient for correctly pin point the causes of the trip. (author) 11 refs., 39 figs.

  20. Discrimination of source reactor type by multivariate statistical analysis of uranium and plutonium isotopic concentrations in unknown irradiated nuclear fuel material.

    Science.gov (United States)

    Robel, Martin; Kristo, Michael J

    2008-11-01

    The problem of identifying the provenance of unknown nuclear material in the environment by multivariate statistical analysis of its uranium and/or plutonium isotopic composition is considered. Such material can be introduced into the environment as a result of nuclear accidents, inadvertent processing losses, illegal dumping of waste, or deliberate trafficking in nuclear materials. Various combinations of reactor type and fuel composition were analyzed using Principal Components Analysis (PCA) and Partial Least Squares Discriminant Analysis (PLSDA) of the concentrations of nine U and Pu isotopes in fuel as a function of burnup. Real-world variation in the concentrations of (234)U and (236)U in the fresh (unirradiated) fuel was incorporated. The U and Pu were also analyzed separately, with results that suggest that, even after reprocessing or environmental fractionation, Pu isotopes can be used to determine both the source reactor type and the initial fuel composition with good discrimination.

  1. Argonaute 2 in cell-secreted microvesicles guides the function of secreted miRNAs in recipient cells.

    Directory of Open Access Journals (Sweden)

    Zhiyuan Lv

    Full Text Available MicroRNAs (miRNAs secreted by cells into microvesicles (MVs form a novel class of signal molecules that mediate intercellular communication. However, several fundamental aspects of secreted miRNAs remain unknown, particularly the mechanism that governs the function or fate of exogenous miRNAs in recipient cells. In the present study, we provide evidence indicating that Argonaute 2 (Ago2 plays a role in stabilizing miRNAs and facilitating the packaging of secreted miRNAs into MVs. More importantly, Ago2 in origin cell-secreted MVs (but not in recipient cells directs the function of secreted miRNAs. First, Ago2 overexpression clearly increased the level of miR-16 in cells transfected with a miR-16 mimic by protecting the miRNAs from degradation in lysosomes. Second, Ago2 overexpression increased the level of miR-16 in cell-secreted MVs, suggesting that Ago2 may facilitate the packaging of secreted miRNAs into MVs. Third, exogenous miR-16 delivered by MVs within the origin cells significantly reduced the Bcl2 protein level in recipient cells, and miR-16 and Bcl2 mRNA were physically associated with exogenous HA-tagged Ago2 (HA-Ago2. Finally, the effect of MV-delivered miR-16 on the production of the Bcl2 protein in recipient cells was not abolished by knocking down Ago2 in the recipient cells.

  2. Collapse of germline piRNAs in the absence of Argonaute3 reveals somatic piRNAs in flies.

    Science.gov (United States)

    Li, Chengjian; Vagin, Vasily V; Lee, Soohyun; Xu, Jia; Ma, Shengmei; Xi, Hualin; Seitz, Hervé; Horwich, Michael D; Syrzycka, Monika; Honda, Barry M; Kittler, Ellen L W; Zapp, Maria L; Klattenhoff, Carla; Schulz, Nadine; Theurkauf, William E; Weng, Zhiping; Zamore, Phillip D

    2009-05-01

    Piwi-interacting RNAs (piRNAs) silence transposons in animal germ cells. piRNAs are thought to derive from long transcripts spanning transposon-rich genomic loci and to direct an autoamplification loop in which an antisense piRNA, bound to Aubergine or Piwi protein, triggers production of a sense piRNA bound to the PIWI protein Argonaute3 (Ago3). In turn, the new piRNA is envisioned to produce a second antisense piRNA. Here, we describe strong loss-of-function mutations in ago3, allowing a direct genetic test of this model. We find that Ago3 acts to amplify piRNA pools and to enforce on them an antisense bias, increasing the number of piRNAs that can act to silence transposons. We also detect a second, Ago3-independent piRNA pathway centered on Piwi. Transposons targeted by this second pathway often reside in the flamenco locus, which is expressed in somatic ovarian follicle cells, suggesting a role for piRNAs beyond the germline. PMID:19395009

  3. Independent chromatin binding of ARGONAUTE4 and SPT5L/KTF1 mediates transcriptional gene silencing.

    Directory of Open Access Journals (Sweden)

    M Jordan Rowley

    2011-06-01

    Full Text Available Eukaryotic genomes contain significant amounts of transposons and repetitive DNA elements, which, if transcribed, can be detrimental to the organism. Expression of these elements is suppressed by establishment of repressive chromatin modifications. In Arabidopsis thaliana, they are silenced by the siRNA-mediated transcriptional gene silencing pathway where long non-coding RNAs (lncRNAs produced by RNA Polymerase V (Pol V guide ARGONAUTE4 (AGO4 to chromatin and attract enzymes that establish repressive chromatin modifications. It is unknown how chromatin modifying enzymes are recruited to chromatin. We show through chromatin immunoprecipitation (ChIP that SPT5L/KTF1, a silencing factor and a homolog of SPT5 elongation factors, binds chromatin at loci subject to transcriptional silencing. Chromatin binding of SPT5L/KTF1 occurs downstream of RNA Polymerase V, but independently from the presence of 24-nt siRNA. We also show that SPT5L/KTF1 and AGO4 are recruited to chromatin in parallel and independently of each other. As shown using methylation-sensitive restriction enzymes, binding of both AGO4 and SPT5L/KTF1 is required for DNA methylation and repressive histone modifications of several loci. We propose that the coordinate binding of SPT5L and AGO4 creates a platform for direct or indirect recruitment of chromatin modifying enzymes.

  4. Expression of the Argonaute protein PiwiL2 and piRNAs in adult mouse mesenchymal stem cells

    Energy Technology Data Exchange (ETDEWEB)

    Wu, Qiuling; Ma, Qi; Shehadeh, Lina A.; Wilson, Amber; Xia, Linghui; Yu, Hong [Department of Molecular and Cellular Pharmacology, Vascular Biology Institute, University of Miami Miller School of Medicine, Miami, FL 33136 (United States); Webster, Keith A., E-mail: kwebster@med.miami.edu [Department of Molecular and Cellular Pharmacology, Vascular Biology Institute, University of Miami Miller School of Medicine, Miami, FL 33136 (United States)

    2010-06-11

    Piwi (P-element-induced wimpy testis) first discovered in Drosophila is a member of the Argonaute family of micro-RNA binding proteins with essential roles in germ-cell development. The murine homologue of PiwiL2, also known as Mili is selectively expressed in the testes, and mice bearing targeted mutations of the PiwiL2 gene are male-sterile. PiwiL2 proteins are thought to protect the germ line genome by suppressing retrotransposons, stabilizing heterochromatin structure, and regulating target genes during meiosis and mitosis. Here, we report that PiwiL2 and associated piRNAs (piRs) may play similar roles in adult mouse mesenchymal stem cells. We found that PiwiL2 is expressed in the cytoplasm of metaphase mesenchymal stem cells from the bone marrow of adult and aged mice. Knockdown of PiwiL2 with a specific siRNA enhanced cell proliferation, significantly increased the number of cells in G1/S and G2/M cell cycle phases and was associated with increased expression of cell cycle genes CCND1, CDK8, microtubule regulation genes, and decreased expression of tumor suppressors Cables 1, LATS, and Cxxc4. The results suggest broader roles for Piwi in genome surveillance beyond the germ line and a possible role in regulating the cell cycle of mesenchymal stem cells.

  5. The effects of fuel microstructure evolution on thermal performance in fast reactor type uranium-plutonium oxide fuel pins up to high burnup

    International Nuclear Information System (INIS)

    Higher power density and larger thermal gradients in fast reactor type uranium-plutonium oxide fuel pins than in water reactor type fuel rods induces the drastic fuel microstructure evolutions, which are useful for estimating the thermal condition at power in the fuel pellets. Recently, fuel-to-cladding gap reopen phenomena is paid attention to due to the influences on the heat transfer behaviors. The objectives of this work are to clarify the dominant factors for fuel microstructure evolution and investigate the gap reopen phenomena on the fuel pin thermal performance up to high burnup. The results of irradiation tests accumulated in JOYO Mk-II core, FFTF, Phenix were exhaustively collected and compiled to the database in this work. Then, the influences of irradiation conditions on the extent of microstructure evolution, such as central void, columnar grain, gas bubble region, densified region, dark ring, and residual gap (including gap reopen phenomena), were reviewed and investigated. Also the thermal behaviors in fuel pins were modeled to apply for the thermal analysis based on the recent knowledge referred to water reactor fuel experiences. It was revealed that the gap reopen phenomena related directly with the fission gas release behaviors in JOYO Mk-II J2 type driver fuel pins whose fabrication, irradiation, and post-irradiation conditions were well characterized. The results of one dimensional thermal analysis at the transverse sections were proved that gap conductance were improved by the condensed material which fills the gap even after the gap reopen phenomena. (author)

  6. The effects of neutron irradiation on the type 316 stainless steel for homemade fast reactor fuel element cladding material

    International Nuclear Information System (INIS)

    The irradiation experiments on the homemade 316 stainless steel of six kinds of chemical composition with different treatment technology used for fast reactor fuel element cladding material are introduced. The materials have been irradiated in the High Flux Engineering Test Reactor (HFETR) to a fluence of 3.1 x 1021 neutron/cm2 (>0.1 MeV) at 650 degree C and subsequently tension has been tested at the same irradiation temperature and room temperature. Microstructure of the some tensile specimens were examined. The experiment results are analyzed and assessed. (authors)

  7. Dependence of form factors for power and temperature on time and power level at FDR-type reactors

    International Nuclear Information System (INIS)

    In order to test neutron power and fuel temperature form factor variations in a reactor core during operational transients several calculations were performed by means of the dynamics program KINE. This code includes one-dimensional time-dependent feedbacks in order to check-point reactor model calculations. It could be shown that the variation of the form factor was positive or negative dependent on the transient, but not more than 3%. At a hypothetical accident with an unrealistic increase of power by 70% an increase of the form factors by about 10% resulted, which is small compared to the safety margins in quasistationary accident analyses with a point model. (orig.)

  8. A new type of Neutrino Detector for Sterile Neutrino Search at Nuclear Reactors and Nuclear Nonproliferation Applications

    OpenAIRE

    Lane, C.; Usman, S. M.; Blackmon, J.; Rasco, C.; Mumm, H. P.; Markoff, D.; Jocher, G. R.; Dorrill, R.; Duvall, M.; J. G. Learned; Li, V; Maricic, J.; Matsuno, S.; Milincic, R.; Negrashov, S.

    2015-01-01

    We describe a new detector, called NuLat, to study electron anti-neutrinos a few meters from a nuclear reactor, and search for anomalous neutrino oscillations. Such oscillations could be caused by sterile neutrinos, and might explain the "Reactor Antineutrino Anomaly". NuLat, is made possible by a natural synergy between the miniTimeCube and mini-LENS programs described in this paper. It features a "Raghavan Optical Lattice" (ROL) consisting of 3375 boron or $^6$Li loaded plastic scintillator...

  9. Decay profiles of β and γ for a radionuclide inventory in equilibrium cycle of a BWR type reactor

    International Nuclear Information System (INIS)

    Presently work the β and γ radiation decay profiles for a radionuclides inventory in equilibrium cycle of a BWR type reactor is presented. The profiles are presented in terms of decay in the activity of the total inventory as well as of the chemical groups that conform the inventory. In the obtaining of the radionuclides inventory in equilibrium cycle the ORIGEN2 code, version 1 was used, which simulates fuel burnup cycles and it calculates the evolution of the isotopic composition as a result of the burnt one, irradiation and decay of the nuclear fuel. It can be observed starting from the results that the decrease in the activity for the initial inventory and the different chemical groups that conform it is approximately proportional to the base 10 logarithm of the time for the first 24 hours of having concluded the burnt one. It can also be observed that the chemical groups that contribute in more proportion to the total activity of the inventory are the lanthanides-actinides and the transition metals, with 39% and 28%, respectively. The groups of alkaline earth metals, halogens, metalloids, noble gases and alkaline metals, contribute with percentages that go from the 8 to 5%. The groups that less they contribute to the total activity of the inventory they are the non metals and semi-metals with smaller proportions that 1%. The chemical groups that more contribute to the energy of β and γ radiation its are the transition metals and the lanthanides-actinides with a change in the order of importance at the end of the 24 hours period. The case of the halogens is of relevance for the case of the γ radiation energy due that occupying the very near third site to the dimensions of the two previous groups. Additionally, the decay in the activity for the total inventory and the groups that conform it can be simulated by means of order 6 polynomials or smaller than describe its behavior appropriately. The results presented in this work, coupled to a distribution model

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-09-15

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

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

    International Nuclear Information System (INIS)

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

  12. Conceptual design of a passive moderator cooling system for a pressure tube type natural circulation boiling water cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, Mukesh [Reactor Engineering Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India); Pal, Eshita, E-mail: eshi.pal@gmail.com [Homi Bhabha National Institute, Anushaktinagar, Mumbai 400 094 (India); Nayak, Arun K.; Vijayan, Pallipattu K. [Reactor Engineering Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India)

    2015-09-15

    Highlights: • Passive moderator cooling system is designed to cool moderator passively during SBO. • PMCS is a system of two natural circulation loops, coupled via a heat exchanger. • RELAP5 analyses show that PMCS maintains moderator within safe limits for 7 days. - Abstract: The recent Fukushima accident has raised strong concern and apprehensions about the safety of reactors in case of a prolonged Station Black Out (SBO) continuing for several days. In view of this, a detailed study was performed simulating this condition in Advanced Heavy Water Reactor. In this study, a novel concept of moderator cooling by passive means has been introduced in the reactor design. The Passive Moderator Cooling System (PMCS) consists of a shell and tube heat exchanger designed to remove 2 MW heat from the moderator inside Calandria. The heat exchanger is located at a suitable elevation from the Calandria of the reactor, such that the hot moderator rises due to buoyancy into the heat exchanger and upon cooling from shell side water returns to Calandria forming a natural circulation loop. The shell side of the heat exchanger is also a natural circulation loop connected to an overhead large water reservoir, namely the GDWP. The objective of the PMCS is to remove the heat from the moderator in case of an SBO and maintaining its temperature below the permissible safe limit (100 °C) for at least 7 days. The paper first describes the concept of the PMCS. The concept has been assessed considering a prolonged SBO for at least 7 days, through an integrated analysis performed using the code RELAP5/MOD3.2 considering all the major components of the reactor. The analysis shows that the PMCS is able to maintain the moderator temperature below boiling conditions for 7 days.

  13. Sonochemical Reactors.

    Science.gov (United States)

    Gogate, Parag R; Patil, Pankaj N

    2016-10-01

    Sonochemical reactors are based on the generation of cavitational events using ultrasound and offer immense potential for the intensification of physical and chemical processing applications. The present work presents a critical analysis of the underlying mechanisms for intensification, available reactor configurations and overview of the different applications exploited successfully, though mostly at laboratory scales. Guidelines have also been presented for optimum selection of the important operating parameters (frequency and intensity of irradiation, temperature and liquid physicochemical properties) as well as the geometric parameters (type of reactor configuration and the number/position of the transducers) so as to maximize the process intensification benefits. The key areas for future work so as to transform the successful technique at laboratory/pilot scale into commercial technology have also been discussed. Overall, it has been established that there is immense potential for sonochemical reactors for process intensification leading to greener processing and economic benefits. Combined efforts from a wide range of disciplines such as material science, physics, chemistry and chemical engineers are required to harness the benefits at commercial scale operation.

  14. Life time of nuclear power plants and new types of reactors; La duree de vie des centrales nucleaires et les nouveaux types de reacteurs

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2003-05-01

    This report, realized by the Evaluation Parliamentary Office of scientific and technological choices, aims to answer simple but fundamental questions for the french electric power production. What are the phenomena which may limit the exploitation time of nuclear power plants? How can we fight against the aging, at which cost and with which safety? The first chapter presents the management of the nuclear power plants life time, an essential element of the park optimization but not a sufficient element. The second chapter details the EPR and the other reactors for 2015 as a bond between the today and tomorrow parks. The last chapter deals with the necessity of efforts in the research and development to succeed in 2035 and presents other reactors in project. (A.L.B.)

  15. The power control system of the Siemens-KWU nuclear power station of the PWR [pressurized water reactors] type

    International Nuclear Information System (INIS)

    Starting with the first nuclear power plant constructed by Siemens AG of the pressurized light water reactor line (PWR), the Obrigheim Nuclear Power Plant (340 MWe net), until the recently constructed plants of 1300 MWe (named 'Konvoi'), the design of the power control system of the plant was continuously improved and optimized using the experience gained in the operation of the earlier generations of plants. The reactor power control system of the Siemens - KWU nuclear power plants is described. The features of this design and of the Siemens designed heavy water power plants (PHWR) Atucha I and Atucha II are mentioned. Curves showing the behaviour of the controlled variables during load changes obtained from plant tests are also shown. (Author)

  16. 3. Interindustry conference on reactor materials science

    International Nuclear Information System (INIS)

    This document contains abstracts on papers presented at the Third Interindustry Conference on Reactor Materials Science (Dimitrovgrad, 27-30 October 1992). The subject scope of the papers is a follows: fuel and fuel elements of power reactors; structural materials of fast breeder reactors and thermonuclear reactors; structural materials of WWER and RBMK type reactors; absorbers and moderators

  17. Analysis of simulation results of damaged nuclear fuel accidents at NPPs with shell-type nuclear reactors

    OpenAIRE

    Igor L. Kozlov

    2015-01-01

    Lessons from the accident at the Fukushima Daiichi NPP made it necessary to reevaluate and intensificate the work on modeling and analyzing various scenarios of severe accidents with damage to the nuclear fuel in the reactor, containment and spent nuclear fuel storage pool with the expansion of the primary initiating event causes group listing. Further development of computational tools for modeling the explosion prevention criteria as to steam and gas mixtures, considering the specific therm...

  18. Type B investigation of the iridium contamination event at the High Flux Isotope Reactor on September 7, 1993

    International Nuclear Information System (INIS)

    On the title date, at ORNL, area radiation alarms sounded during a routine transfer of a shielding cask (containing 60 Ci192Ir) from the HFIR pool side to a transport truck. Small amounts of Ir were released from the cask onto the reactor bay floor. The floor was cleaned, and the cask was shipped to a hot cell at Building 3047 on Oct. 3, 1993. The event was caused by rupture of one of the Ir target rods after it was loaded into the cask for normal transport operations; the rupture was the result of steam generation in the target rod soon after it was placed in the cask (water had entered the target rod through a tiny defect in a weld while it was in the reactor under pressure). While the target rods were in the reactor and reactor pool, there was sufficient cooling to prevent steam generation; when the target rod was loaded into the dry transport cask, the temperature increased enough to result in boiling of the trapped water and produced high enough pressure to result in rupture. The escaping steam ejected some of the Ir pellets. The event was reported as Occurrence Report Number ORO--MMES-X10HFIR-1993-0030, dated Sept. 8, 1993. Analysis indicated that the following conditions were probable causes: less than adequate welding procedures, practices, or techniques, material controls, or inspection methods, or combination thereof, could have led to weld defects, affecting the integrity of target rod IR-75; less than adequate secondary containment in the cask allowed Ir pellets to escape

  19. Results of R+D of nuclear power plants with WWER-1000 type light-water reactors. Part 2

    International Nuclear Information System (INIS)

    Volume II of the proceedings contains full texts of 32 contributions, all of which fall within the INIS subject scope. The conference was organized to facilitate exchange of knowledge and experience between Czechoslovak organizations engaged in the State Research Project ''Nuclear Power Facilities with 1000 MW Light-Water Reactors'' and domestic suppliers of the requisite equipment. (Z.M.). 45 figs., 9 tabs., 201 refs

  20. Results of R+D of nuclear power plants with WWER-1000 type light-water reactors. Part 1

    International Nuclear Information System (INIS)

    Volume I of the proceedings contains full texts of 44 invited lectures and contributions, all of which fall within the INIS subject scope. The conference was organized to facilitate exchange of knowledge and experience between Czechoslovak organizations engaged in the State Research Project ''Nuclear Power Facilities with 1000 MW Light-Water Reactors'' and domestic suppliers of the requisite equipment. (Z.M.). 111 figs., 24 tabs., 15 refs

  1. 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)

  2. A report on the transport of MTR-type spent fuel assemblies of the Philippine Research Reactor (PRR-1)

    International Nuclear Information System (INIS)

    Fifty one (51) fuel assemblies of mixed enrichment from the Philippine Research Reactor (PRR-1), consisting of 50 spent and 1 fresh, were shipped to the United States last 14 March 1999 under the U.S. Return of Foreign Research Reactor (FRR) fuel policy. The shipment was in line with the U.S. initiative to implement its Record of Decision (ROD) which took effect on 13 May 1996 to accept and manage all FRR uranium fuel of U.S. origin and enriched in the United States. The shipment program would last10 years, ending midnight of 13 May 2006. The ROD provided a 3 year extension period within which to accept FRR spent nuclear fuel (SNF) withdrawn from reactors after 2006. The U.S. policy gave priority to the NPT significance of high enriched U, as the prime target of the return of FRR policy. Classified as a developing country, the Philippines, through the PNRI, signed a contract with the U.S. Department of Energy for the cost-free shipment of PRR-1 spent fuel to the United States. Spent fuel loading and transport operations to the port area lasted seven (7) days, from 8 to 14 March 1999. (Author)

  3. Possible pressurized thermal shock events during large primary to secondary leakage. The Hungarian AGNES project and PRISE accident scenarios in VVER-440/V213 type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Perneczky, L. [KFKI Atomic Energy Research Inst., Budabest (Hungary)

    1997-12-31

    Nuclear power plants of WWER-440/213-type have several special features. Consequently, the transient behaviour of such a reactor system should be different from the behaviour of the PWRs of western design. The opening of the steam generator (SG) collector cover, as a specific primary to secondary circuit leakage (PRISE) occurring in WWER-type reactors happened first time in Rovno NPP Unit I on January 22, 1982. Similar accident was studied in the framework of IAEA project RER/9/004 in 1987-88 using the RELAP4/mod6 code. The Hungarian AGNES (Advanced General and New Evaluation of Safety) project was performed in the period 1991-94 with the aim to reassess the safety of the Paks NPP using state-of-the-art techniques. The project comprised three type of analyses for the primary to secondary circuit leakages: Design Basis Accident (DBA) analyses, Pressurized Thermal Shock (PTS) study and deterministic analyses for Probabilistic Safety Analysis (PSA). Major part of the thermohydraulic analyses has been performed by the RELAP5/mod2.5/V251 code version with two input models. 32 refs.

  4. Fuel gases generation in the primary contention during a coolant loss accident in a nuclear power plant with reactor type BWR

    International Nuclear Information System (INIS)

    During an accident design base of coolant loos, the hydrogen gas can accumulate inside the primary contention as a result of several generation mechanisms among those that are: 1) the reaction metal-water involving the zirconium of the fuel cladding and the reactor coolant, 2) the metals corrosion for the solutions used in the emergency cooling and dew of the contention, and 3) the radio-decomposition of the cooling solutions of post-accident emergency. In this work the contribution of each generation mechanism to the hydrogen total in the primary contention is analyzed, considering typical inventories of zirconium, zinc, aluminum and fission products in balance cycle of a reactor type BWR. In the analysis the distribution model of fission products and hydrogen production proposed in the regulator guide 1.7, Rev. 2 of the US NRC was used. The results indicate that the mechanism that more contributes to the hydrogen generation at the end of a period of 24 hours of initiate the accident is the radio-decomposition of the cooling solutions of post-accident emergency continued by the reaction metal-water involving the zirconium of the fuel cladding with the reactor coolant, and lastly the aluminum and zinc oxidation present in the primary contention. However, the reaction metal-water involving the zirconium of the fuel cladding and the reactor coolant is the mechanism that more contributes to the hydrogen generation in the first moments after the accident. This study constitutes the first part of the general analysis of the generation, transport and control of fuel gases in the primary contention during a coolant loss accident in BWRs. (Author)

  5. Experience of complex use of radioactive methods of FFD on operating and shut down reactor for diagnostics of types of untight fuel rods

    International Nuclear Information System (INIS)

    The report contains a review of principles and methods, used by the experts of the Russian Research Center 'Kurchatov Institute' at diagnostics of fuel rods tightness during operation period of WWER type NPP, and also interpretation practices of obtained assessments. The agreement between assessments of untight rods parameters, which were determined by FFD radioactive methods applied for operating/shut down reactor, and the scale and character of rod leakage having a place in a reality is illustrated with concrete examples, taken from experience of WWER operation. (authors)

  6. A Potential Protein-RNA Recognition Event Along the RISC-Loading Pathway from the Structure of A. aeolicus Argonaute with Externally Bound siRNA

    Energy Technology Data Exchange (ETDEWEB)

    Yuan,Y.; Pei, Y.; Chen, H.; Tuschl, T.; Patel, D.

    2006-01-01

    Argonaute proteins are key components of the RNA-induced silencing complex (RISC). They provide both architectural and catalytic functionalities associated with small interfering RNA (siRNA) guide strand recognition and subsequent guide strand-mediated cleavage of complementary mRNAs. We report on the 3.0 {angstrom} crystal structures of 22-mer and 26-mer siRNAs bound to Aquifex aeolicus Argonaute (Aa-Ago), where one 2 nt 3' overhang of the siRNA inserts into a cavity positioned on the outer surface of the PAZ-containing lobe of the bilobal Aa-Ago architecture. The first overhang nucleotide stacks over a tyrosine ring, while the second overhang nucleotide, together with the intervening sugar-phosphate backbone, inserts into a preformed surface cavity. Photochemical crosslinking studies on Aa-Ago with 5-iodoU-labeled single-stranded siRNA and siRNA duplex provide support for this externally bound siRNA-Aa-Ago complex. The structure and biochemical data together provide insights into a protein-RNA recognition event potentially associated with the RISC-loading pathway.

  7. Natural variation of the amino-terminal glutamine-rich domain in Drosophila argonaute2 is not associated with developmental defects.

    Directory of Open Access Journals (Sweden)

    Daniel Hain

    Full Text Available The Drosophila argonaute2 (ago2 gene plays a major role in siRNA mediated RNA silencing pathways. Unlike mammalian Argonaute proteins, the Drosophila protein has an unusual amino-terminal domain made up largely of multiple copies of glutamine-rich repeats (GRRs. We report here that the ago2 locus produces an alternative transcript that encodes a putative short isoform without this amino-terminal domain. Several ago2 mutations previously reported to be null alleles only abolish expression of the long, GRR-containing isoform. Analysis of drop out (dop mutations had previously suggested that variations in GRR copy number result in defects in RNAi and embryonic development. However, we find that dop mutations genetically complement transcript-null alleles of ago2 and that ago2 alleles with variant GRR copy numbers support normal development. In addition, we show that the assembly of the central RNAi machinery, the RISC (RNA induced silencing complex, is unimpaired in embryos when GRR copy number is altered. In fact, we find that GRR copy number is highly variable in natural D. melanogaster populations as well as in laboratory strains. Finally, while many other insects share an extensive, glutamine-rich Ago2 amino-terminal domain, its primary sequence varies drastically between species. Our data indicate that GRR variation does not modulate an essential function of Ago2 and that the amino-terminal domain of Ago2 is subject to rapid evolution.

  8. miRNA–target chimeras reveal miRNA 3′-end pairing as a major determinant of Argonaute target specificity

    Science.gov (United States)

    Moore, Michael J.; Scheel, Troels K. H.; Luna, Joseph M.; Park, Christopher Y.; Fak, John J.; Nishiuchi, Eiko; Rice, Charles M.; Darnell, Robert B.

    2015-01-01

    microRNAs (miRNAs) act as sequence-specific guides for Argonaute (AGO) proteins, which mediate posttranscriptional silencing of target messenger RNAs. Despite their importance in many biological processes, rules governing AGO–miRNA targeting are only partially understood. Here we report a modified AGO HITS-CLIP strategy termed CLEAR (covalent ligation of endogenous Argonaute-bound RNAs)-CLIP, which enriches miRNAs ligated to their endogenous mRNA targets. CLEAR-CLIP mapped ∼130,000 endogenous miRNA–target interactions in mouse brain and ∼40,000 in human hepatoma cells. Motif and structural analysis define expanded pairing rules for over 200 mammalian miRNAs. Most interactions combine seed-based pairing with distinct, miRNA-specific patterns of auxiliary pairing. At some regulatory sites, this specificity confers distinct silencing functions to miRNA family members with shared seed sequences but divergent 3′-ends. This work provides a means for explicit biochemical identification of miRNA sites in vivo, leading to the discovery that miRNA 3′-end pairing is a general determinant of AGO binding specificity. PMID:26602609

  9. State of art report for critical flow model to analyze a break flow in pressurizer of integral type reactor

    International Nuclear Information System (INIS)

    At a critical flow condition, the flow rate can't exceed a maximum value for given upstream conditions and the limited flow rate is called as a critical flow rate. The phenomena of critical flow occur at the discharge of a single phase gas or subcooled water through nozzles and pipes. Among the previous researches on critical flow, many accurate correlations on pressure, temperature and flow rate are represented for the single phase gas. However, for the two phase critical flow, the results of previous work showed that there was a large discrepancy between the analytical and experimental data and the data were in agreement for the limited thermodynamic conditions. Thus, further studies are required to enhance the two phase critical flow model. In the integral reactor, the critical flows of nitrogen gas and subcooled water are expected for the break of gas cylinder pipeline connected to the pressurizer. It requires that the inlet shape of the pipe and the nitrogen gas effect should be considered for the critical flow of integral reactor. The nitrogen gas exist in the pressurizer may affect the flow rate of primary coolant, which has been considered only for a few previous researches. Thus, the evaluation of the effect of the nitrogen on the critical flow gas should be preceded for the proper analysis of the critical flow in the integral reactor. In this report, not only the essences of previous work on critical flow were investigated and summarized but also the effect of nitrogen gas and the inlet shape of the pipe on the critical flow were also investigated. (author)

  10. State of art report for critical flow model to analyze a break flow in pressurizer of integral type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Yeon Moon; Lee, D. J.; Yoon, J. H.; Kim, J. P.; Kim, H. Y

    1999-03-01

    At a critical flow condition, the flow rate can't exceed a maximum value for given upstream conditions and the limited flow rate is called as a critical flow rate. The phenomena of critical flow occur at the discharge of a single phase gas or subcooled water through nozzles and pipes. Among the previous researches on critical flow, many accurate correlations on pressure, temperature and flow rate are represented for the single phase gas. However, for the two phase critical flow, the results of previous work showed that there was a large discrepancy between the analytical and experimental data and the data were in agreement for the limited thermodynamic conditions. Thus, further studies are required to enhance the two phase critical flow model. In the integral reactor, the critical flows of nitrogen gas and subcooled water are expected for the break of gas cylinder pipeline connected to the pressurizer. It requires that the inlet shape of the pipe and the nitrogen gas effect should be considered for the critical flow of integral reactor. The nitrogen gas exist in the pressurizer may affect the flow rate of primary coolant, which has been considered only for a few previous researches. Thus, the evaluation of the effect of the nitrogen on the critical flow gas should be preceded for the proper analysis of the critical flow in the integral reactor. In this report, not only the essences of previous work on critical flow were investigated and summarized but also the effect of nitrogen gas and the inlet shape of the pipe on the critical flow were also investigated. (author)

  11. MCTP, a code for the thermo-mechanical analysis of a fuel rod of BWR type reactors (Neutron part); MCTP, un codigo para el analisis termo-mecanico de una barra combustible de reactores tipo BWR (Parte Neutronica)

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez L, H.; Ortiz V, J. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: hhl@nuclear.inin.mx

    2003-07-01

    In the National Institute of Nuclear Research of Mexico a code for the thermo-mechanical analysis of the fuel rods of the BWR type reactors of the Nucleo electric Central of Laguna Verde is developed. The code solves the diffusion equation in cylindrical coordinates with several energy groups. The code, likewise, calculates the temperature distribution and power distribution in those fuel rods. The code is denominated Multi groups With Temperatures and Power (MCTP). In the code, the energy with which the fission neutrons are emitted it is divided in six groups. They are also considered the produced perturbations by the changes in the temperatures of the materials that constitute the fuel rods, the content of fission products, the uranium consumption and in its case the gadolinium, as well as the plutonium production. In this work there are present preliminary results obtained with the code, using data of operation of the Nucleo electric Central of Laguna Verde. (Author)

  12. Kinetic parameters determination through power spectral densities measurements using pulse-type detectors in the IPEN/MB-01 research reactor

    International Nuclear Information System (INIS)

    Nowadays, a collaborative effort to improve the prediction accuracy of some kinetic parameters has been recommended. In special, a target accuracy of ±3% (1 s.d.) was requested for βeff calculations, in way that βeff values must be measured with an experimental error of less than 3%. In such a way, the Reactor Physics Group at IPEN/MB-01 Research Reactor has been compiled an experimental data bank of kinetic parameters, including βeff, Λ, βeff /Λ and others, based on noise analyses techniques. The implemented techniques are: Power Spectral Densities (PSD) using current-type detectors, Rossi-α and Feynman-α techniques. All these techniques provided βeff values with uncertainties within the target accuracy. In order to conclude this data bank, in this work we have been performed PSD measurements using pulse-type detectors. The main advantage of this technique is that it is possible to eliminate some electronic modules, needed in the current-mode experiments, which are sources of parasitic noises. Furthermore, this technique can explain an anomalous behavior reported in current-mode measurements, which is a non-observation of a theoretical predicted plateau above 200 Hz, approximately. Once completed, the kinetic parameters data bank should provide valuable information to determine whether or not the currently data libraries are sufficiently accurate to predict these measured parameters. Further, we intend to submit a proposal for the first international benchmark related to βeff measurements, to the International Reactor Physics Experiment Evaluation Project (IRPhEP, NEA data bank). (author)

  13. Kinetic parameters determination through power spectral densities measurements using pulse-type detectors in the IPEN/MB-01 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kuramoto, Renato Yoichi Ribeiro; Santos, Adimir dos; Jerez, Rogerio; Diniz, Ricardo [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)]. E-mails: ryrkuram@ipen.br; asantos@ipen.br

    2007-07-01

    Nowadays, a collaborative effort to improve the prediction accuracy of some kinetic parameters has been recommended. In special, a target accuracy of {+-}3% (1 s.d.) was requested for {beta}{sub eff} calculations, in way that {beta}{sub eff} values must be measured with an experimental error of less than 3%. In such a way, the Reactor Physics Group at IPEN/MB-01 Research Reactor has been compiled an experimental data bank of kinetic parameters, including {beta}{sub eff}, {lambda}, {beta}{sub eff} /{lambda} and others, based on noise analyses techniques. The implemented techniques are: Power Spectral Densities (PSD) using current-type detectors, Rossi-{alpha} and Feynman-{alpha} techniques. All these techniques provided {beta}{sub eff} values with uncertainties within the target accuracy. In order to conclude this data bank, in this work we have been performed PSD measurements using pulse-type detectors. The main advantage of this technique is that it is possible to eliminate some electronic modules, needed in the current-mode experiments, which are sources of parasitic noises. Furthermore, this technique can explain an anomalous behavior reported in current-mode measurements, which is a non-observation of a theoretical predicted plateau above 200 Hz, approximately. Once completed, the kinetic parameters data bank should provide valuable information to determine whether or not the currently data libraries are sufficiently accurate to predict these measured parameters. Further, we intend to submit a proposal for the first international benchmark related to {beta}{sub eff} measurements, to the International Reactor Physics Experiment Evaluation Project (IRPhEP, NEA data bank). (author)

  14. N Reactor

    Data.gov (United States)

    Federal Laboratory Consortium — The last of Hanfordqaodmasdkwaspemas7ajkqlsmdqpakldnzsdflss nine plutonium production reactors to be built was the N Reactor.This reactor was called a dual purpose...

  15. The use of U/sub 3/Si/sub 2/ dispersed in aluminum in plate-type fuel elements for research and test reactors

    Energy Technology Data Exchange (ETDEWEB)

    Snelgrove, J.L.; Domagala, R.F.; Hofman, G.L.; Wiencek, T.C.; Copeland, G.L.; Hobbs, R.W.; Senn, R.L.

    1987-10-01

    A high-density fuel based on U/sub 3/Si/sub 2/ dispersed in aluminum has been developed and tested for use in converting plate-type research and test reactors from the use of highly enriched uranium to the use of low-enriched uranium. Results of preirradiation testing and the irradiation and postirradiation examination of miniature fuel plates and full-sized fuel elements are summarized. Swelling of the U/sub 3/Si/sub 2/ fuel particles is a linear function of the fission density in the particle to well beyond the fission density achievable in low-enriched fuels. U/sub 3/Si/sub 2/ particle swelling rate is approximately the same as that of the commonly used UAl/sub x/ fuel particle. The presence of minor amounts of U/sub 3/Si or uranium solid solution in the fuel result in greater, but still acceptable, fuel swelling. Blister threshold temperatures are at least as high as those of currently used fuels. An exothermic reaction occurs near the aluminum melting temperature, but the measured energy releases were low enough not to substantially worsen the consequences of an accident. U/sub 3/Si/sub 2/-aluminum dispersion fuel with uranium densities up to at least 4.8 Mg/m/sup 3/ is a suitable LEU fuel for typical plate-type research and test reactors. 42 refs., 28 figs., 7 tabs.

  16. Application of perturbation theory to sensitivity calculations of PWR type reactor cores using the two-channel model

    International Nuclear Information System (INIS)

    Sensitivity calculations are very important in design and safety of nuclear reactor cores. Large codes with a great number of physical considerations have been used to perform sensitivity studies. However, these codes need long computation time involving high costs. The perturbation theory has constituted an efficient and economical method to perform sensitivity analysis. The present work is an application of the perturbation theory (matricial formalism) to a simplified model of DNB (Departure from Nucleate Boiling) analysis to perform sensitivity calculations in PWR cores. Expressions to calculate the sensitivity coefficients of enthalpy and coolant velocity with respect to coolant density and hot channel area were developed from the proposed model. The CASNUR.FOR code to evaluate these sensitivity coefficients was written in Fortran. The comparison between results obtained from the matricial formalism of perturbation theory with those obtained directly from the proposed model makes evident the efficiency and potentiality of this perturbation method for nuclear reactor cores sensitivity calculations (author). 23 refs, 4 figs, 7 tabs

  17. A new type of Neutrino Detector for Sterile Neutrino Search at Nuclear Reactors and Nuclear Nonproliferation Applications

    CERN Document Server

    Lane, C; Blackmon, J; Rasco, C; Mumm, H P; Markoff, D; Jocher, G R; Dorrill, R; Duvall, M; Learned, J G; Li, V; Maricic, J; Matsuno, S; Milincic, R; Negrashov, S; Sakai, M; Rosen, M; Varner, G; Huber, P; Pitt, M L; Rountree, S D; Vogelaar, R B; Wright, T; Yokley, Z

    2015-01-01

    We describe a new detector, called NuLat, to study electron anti-neutrinos a few meters from a nuclear reactor, and search for anomalous neutrino oscillations. Such oscillations could be caused by sterile neutrinos, and might explain the "Reactor Antineutrino Anomaly". NuLat, is made possible by a natural synergy between the miniTimeCube and mini-LENS programs described in this paper. It features a "Raghavan Optical Lattice" (ROL) consisting of 3375 boron or $^6$Li loaded plastic scintillator cubical cells 6.3\\,cm (2.500") on a side. Cell boundaries have a 0.127\\,mm (0.005") air gap, resulting in total internal reflection guiding most of the light down the 3 cardinal directions. The ROL detector technology for NuLat gives excellent spatial and energy resolution and allows for in-depth event topology studies. These features allow us to discern inverse beta decay (IBD) signals and the putative oscillation pattern, even in the presence of other backgrounds. We discuss here test venues, efficiency, sensitivity an...

  18. Defuelling of the UTR-300 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Scott, R.D.; Banford, H.M.; East, B.W. [Scottish Universities Research and Reactor Centre, Glasgow (United Kingdom)

    1997-07-01

    The UTR-300 reactor at the Scottish Universities Research and Reactor Centre was based on the original Argonaut design with two aluminium core tanks set in a graphite reflector each containing six fuel elements cooled and moderated by water flowing up through the tanks in a closed primary circuit. The fuel plates in the original 13-plate elements were uranium oxide-aluminium with a 22g loading of 90% {sup 235}U. After 7 years of operation at 100 kW (10 kW average), the maximum power was increased to 300 kW (30 kW average) and, in order to maintain the operational excess reactivity, it was necessary to add another plate to each element progressively over the years until they all contained 14 plates. These extra plates were uranium metal-aluminium with 24.5 g of 90% {sup 235}U. No further modification of the elements was possible and so, with reactivity steadily decreasing, and for a variety of other reasons, a decision was taken to cease operation in September 1995. This paper describes the procedures whereby the fuel was unloaded from the core into a UNIFETCH flask equipped with a specially designed rotating gamma ray shield and then transported on two separate loads to Dounreay for reprocessing. (author)

  19. Defuelling of the UTR-300 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Scott, R.D.; Banford, H.M.; East, B.W. [Scottish Universities Research and Reactor Centre, Glasgow (United Kingdom)

    1997-07-01

    The UTR-300 reactor at the Scottish Universities Research and Reactor Centre was based on the original Argonaut design with two aluminium core tanks set in a graphite reflector each containing six fuel elements cooled and moderated by water flowing up through the tanks in a closed primary circuit. The fuel plates in the original 13-plate elements were uranium oxide-aluminium with a 22g loading of 90% {sup 235}U. After 7 years of operation at 100 kW (10 kW average), the maximum power was increased to 300 kW (30 kW average) and, in order to maintain the operational excess reactivity, it was necessary to add another plate to each element progressively over the years until they all contained 14 plates. These extra plates were uranium metal-aluminium with 24.5 g of 90% {sup 235}U. No further modification of the elements was possible and so, with reactivity steadily decreasing, and for a variety of other reasons, a decision was taken to cease operation in September 1995. This paper describes the procedures whereby the fuel was unloaded from the core into a UNIFETCH flask equipped with a specially designed rotating gamma ray shield and then transported on two separate loads to Dounreay for reprocessing. (author) 2 figs., 2 tabs., refs.

  20. Análisis para la modelación y optimización geométrica de un reactor tipo tornillo sin-fin empleando el método de grafos dicromáticos//Analysis for geometric modeling and optimization of a worm type reactor using the method of dichromatic graph

    Directory of Open Access Journals (Sweden)

    Armando Díaz-Concepción

    2015-09-01

    Full Text Available En el presente trabajo se realiza la modelación, simulación y optimización de un reactor utilizado en las plantas para la obtención de un alimento animal, sobre la base de la predigestión del bagacillo de caña y el hidróxido de calcio en presencia de vapor denominado PREDICAL utilizando grafos dicromáticos. Se obtuvo el modelo matemático para el diseño del reactor, donde se vinculan las variables geométricas y tecnológicas. El modelo formulado permitió la optimización de la variable costo a partir de minimizar la variable geométrica diámetro exterior del reactor. Palabras claves: modelación reactor tipo tornillo sinfin, grafos dicromáticos, modelo matemático________________________________________________________________________________AbstractThe present work performs modeling, simulation and optimization of a reactor used in plants for the obtencion of animal feed. It's made on the basis of pre-digestion of cane bagasse and calcium hydroxide in the presence of steam called PREDICAL and using dichromatic graphs. It was achieved the mathematical model for the design of the reactor, where are linked geometric and technological variables. The model developed allowed cost optimization based on minimize the geometric variable outside diameter of the reactor. Key words: worm type reactor modeling, dichromatic graphs, mathematical model.

  1. Reactor Vessel Surveillance Program for Advanced Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Kyeong-Hoon; Kim, Tae-Wan; Lee, Gyu-Mahn; Kim, Jong-Wook; Park, Keun-Bae; Kim, Keung-Koo

    2008-10-15

    This report provides the design requirements of an integral type reactor vessel surveillance program for an integral type reactor in accordance with the requirements of Korean MEST (Ministry of Education, Science and Technology Development) Notice 2008-18. This report covers the requirements for the design of surveillance capsule assemblies including their test specimens, test block materials, handling tools, and monitors of the surveillance capsule neutron fluence and temperature. In addition, this report provides design requirements for the program for irradiation surveillance of reactor vessel materials, a layout of specimens and monitors in the surveillance capsule, procedures of installation and retrieval of the surveillance capsule assemblies, and the layout of the surveillance capsule assemblies in the reactor.

  2. Nuclear reactor noise investigations on boiling effects in a simulated MTR-type fuel assembly. Doctoral thesis

    Energy Technology Data Exchange (ETDEWEB)

    Kozma, R.

    1992-05-04

    Boiling experiments at HOR have a relatively long history that began more than a decade ago. Following some introductory out-of-pile experiments, a boiling setup was operating at HOR between 1984 and 1986. Based on the experience of that setup, a new boiling experiment, NIOBE (Noise Investigations On Boiling Effects), has been designed. The NIOBE setup has been in operation since the end of 1986. The present work contains results of experiments performed between 1988 and 1991, with the exception of Chapter 5, in which experiments with the predecessor of NIOBE are treated. The thesis is based on results of experiments performed at the NIOBE loop located in a research reactor (HOR).

  3. The use of waveguide acoustic probes for void fraction measurement in the evaporator of BN-350-Type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Melnikov, V.I.; Nigmatulin, B.I.

    1995-09-01

    The present paper deals with some results of the experimental studies which have been carried out to investigate the steam generation dynamics in the Field tubes of sodium-water evaporators used in the BN-350 reactors. The void fraction measurements have been taken with the aid of waveguide acoustic transducers manufactured in accordance with a specially designed technology (waveguide acoustic transducers-WAT technology). Presented in this paper also the transducer design and calibration methods, as well as the diagram showing transducers arrengment in the evaporator. The transducers under test featured a waveguide of about 4 m in length and a 200-mm long sensitive element (probe). Besides, this paper specifies the void fraction data obtained through measurements in diverse points of the evaporator. The studies revealed that the period of observed fluctuations in the void fraction amounted to few seconds and was largely dependent on the level of water in the evaporator.

  4. Effect of nitrogen on high temperature mechanical properties of type 316LN SS for fast reactor applications

    International Nuclear Information System (INIS)

    Long term creep, low cycle fatigue and creep-fatigue interaction properties as well as compatibility with liquid sodium coolant, govern the choice of materials for out-of-core structural components of sodium cooled fast reactors (SFRs). 316 L(N) SS containing 0.07 Wt.% nitrogen is the current preferred material for all the high temperature structural components of SFRs. For the prototype Fast Breeder Reactor also, which is nearing commissioning at Kalpakkam, 316 LN (SS) has been used for all the major high temperature structural components. It is proposed to design future SFRs for a life of at least 60 years. Therefore materials with higher creep and low cycle fatigue strength are required. Increasing the nitrogen content in 316L(N) SS is one of the approaches being pursued at Indira Gandhi Centre for Atomic Research (IGCAR) to develop a SFR structural material suitable for 60 years design life. Towards this, a strong R and D programme is currently underway at IGCAR. The effect of nitrogen in the range of 0.07 to 0.22 Wt.% on the tensile creep and low cycle fatigue behavior of 316 LN SS has been extensively investigated. Both tensile and creep strength were found to increase with increase in nitrogen content. High temperature low cycle fatigue life was found to peak in 316LN SS containing 0.14 Wt.%. The paper discusses the monotonic and cyclic deformation and fracture mechanisms by which nitrogen improves the creep and low cycle fatigue properties of 316 LN SS. (author)

  5. Small, long-life high temperature gas-cooled reactor free from prompt supercritical accidents by particle-type burnable poisons

    International Nuclear Information System (INIS)

    A design concept for a high temperature gas-cooled reactor without the possibility of a prompt supercritical accident has been proposed by coupling the use of particle-type burnable poison (BP) and criticality control by the core temperature. The combinations of two different BPs, B4C and Gd2O3 particles and B4C and CdO particles, with the proper particle sizes and the appropriate volume ratio, showed excellent performance in controlling excess reactivity and flattening the reactivity swing. To maintain reactivity at a lower level than the prompt critical state, the reactor was designed to operate in a subcritical mode for a burnup period or for the whole operation cycle. Under subcritical operation during the partial burnup period, the core temperature had to be lowered by at least 164 K for the loading of B4C + Gd2O3 particles and by at least 178 K for the B4C + CdO particles, which in turn dropped the thermal efficiency from 48% to 42.26% and 41.77%, respectively. On the other hand, under full subcritical operation, a greater decrease of core temperature was required. Remarkable decreases in the core temperatures, approximately 347 K for the B4C + Gd2O3 case and approximately 280 K for the B4C + CdO case, resulted in the drop of thermal efficiency to only 35.9% and 38.2%, respectively. Therefore, the relative importance of the increase in passive safety and the decrease in thermal efficiency must be considered with regard to their importance in nuclear reactor design. (author)

  6. The importance of carry out studies about the use of passive autocatalytic recombiners for hydrogen control in reactors type ESBWR; La importancia de realizar estudios sobre el uso de recombinadores autocataliticos pasivos para control de hidrogeno en reactores tipo ESBWR

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez J, J.; Morales S, J. B. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Mexico D. F. (Mexico)], e-mail: jersonsanchez@gmail.com

    2009-10-15

    A way to satisfy and to guarantee the energy necessities in the future is increasing in a gradual way the creation of nuclear power plants, introducing advanced designs in its systems that contribute in way substantial in the security of the same nuclear plants. The tendency of new designs of these nuclear plants is the incorporation of systems more reliable and sure, and that the operation does not depend on external factors as the electric power, motors diesel or the action of the operator of nuclear plant, what is known as security passive systems. In this sense, the passive autocatalytic recombiners are a contribution toward the use of this type of systems. At the present time it is had studies of the incorporation of passive autocatalytic recombiners in nuclear plants in operation and that they have contributed to minimize the danger associated to hydrogen. The present work contains a first approach to the study of hydrogen recombiners incorporation in advanced nuclear plants, for this case in a nuclear power plant of ESBWR type. To achieve our objective it seeks to use specialized codes as RELAP/SCDAP to obtain simulations of passive autocatalytic recombiners behaviour and we can to estimate their operation inside the reactor contention, contemplating the possibility to use other codes like SCILAB and/or MATLAB for the simulation of a passive autocatalytic recombiner. (Author)

  7. TRIGA-LEU fuel immediately available for substitution in plate-type research reactors up to 15MW

    International Nuclear Information System (INIS)

    New 20% enriched TRIGA type fuel has been developed to replace the TRIGA 70% and 90% enriched fuel and to replace highly enriched uranium plate-type fuel. At present, production elements and undergoing in-pile demonstration testing. The new fuel contains the characteristics of the U-ZrH fuel and can be normally used with the existing core structure. (author)

  8. On the shape of stress corrosion cracks in sensitized Type 304 SS in Boiling Water Reactor primary coolant piping at 288 °C

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sang-Kwon; Kramer, Daniel [Center for Electrochemical Science and Technology, Department of Materials Science and Engineering, Pennsylvania State University, University Park, PA 16802 (United States); Macdonald, Digby D., E-mail: macdonald@berkeley.edu [Department of Nuclear Engineering, University of California at Berkeley, Berkeley, CA 94720 (United States)

    2014-11-15

    Evolution of the shape of surface cracks in sensitized Type 304 SS in Boiling Water Reactor primary coolant circuit piping at the reactor operating temperature of 288 °C is explored as a function of various environmental variables, such as electrochemical potential (ECP), solution conductivity, flow velocity, and multiplier for the oxygen reduction reaction (ORR) standard exchange current density (SECD), using the coupled environment fracture model (CEFM). For this work, the CEFM was upgraded by incorporating Shoji’s model for calculating the crack tip strain rate and more advanced expressions were used for estimating the stress intensity factor for semi-elliptical surface cracks. This revised CEFM accurately predicts the dependence of the crack growth rate on stress intensity factor and offers an alternative explanation for the development of semi-elliptical cracks than that provided by fracture mechanics alone. The evolution of surface crack semi-elliptical shape depends strongly upon various environmental variables identified above, and the CEFM predicts that the minor axis of the ellipse should be oriented perpendicular to the surface, in agreement with observation. The development of the observed semi-elliptical cracks with the minor axis perpendicular to the surface is therefore attributed to the dependence of the crack growth rate on the electrochemical crack length.

  9. Reactor safety in Eastern Europe

    International Nuclear Information System (INIS)

    The papers presented to the GRS colloquium refer to the cooperative activities for reactor accident analysis and modification of the GRS computer codes for their application to reactors of the Russian design types of WWER or RBMK. Another topic is the safety of RBMK reactors in particular, and the current status of investigations and studies addressing the containment of unit 4 of the Chernobyl reactor station. All papers are indexed separately in report GRS--117. (HP)

  10. Susceptibility to stress corrosion in stainless steels type AISI 321 and 12X18H10T used in PWR type reactors (WWER)

    International Nuclear Information System (INIS)

    Titanium stabilized stainless steels have been utilized in sovietic pressurized water reactors (VVER) for avoid the susceptibility to Intergranular Corrosion (IGC) present in other austenitic stainless steels. However the Intergranular Corrosion resistance of this kind of materials has been questioned because of Intergranular Stress Corrosion Cracking failures (IGSCC) have been reported. This paper study the electrochemical behavior of the AISI 321 stainless steel in a H3BO3 Solution contaminated with chlorides and its susceptibility to Intergranular Corrosion.Electrochemical prediction diagrams of the stainless steels AISI 321 and 12X18H10T (sovietic) sensitized (600 Centigrade, 3 h.) were compared. Cylindrical and conical samples were used in Slow Strain Rate Tests (SSRT), to determine the susceptibility to Stress Corrosion Cracking (SCC) in AISI 321 and 12X18H10T stainless steels. The results obtained showed that the temperature of the solution is a very important factor to detect this susceptibility. Fractography studies on the fracture surfaces of the samples obtained in the SSRT at high temperature were realized. Corrosion velocities of both AISI 321 and 12X18H10T stainless steels were determined using conical samples in the CERT system at high temperature. E.D.A.X. analysis was employed in both AISI 321 and 12X18H10T stainless steels in order to explain the degree of sensitization. (Author)

  11. Application of the MOVE algorithm for the identification of reduced order models of a core of a BWR type reactor

    International Nuclear Information System (INIS)

    Presently work is applied the modified algorithm of the ellipsoid of optimal volume (MOVE) to a reduced order model of 5 differential equations of the core of a boiling water reactor (BWR) with the purpose of estimating the parameters that model the dynamics. The viability is analyzed of carrying out an analysis that calculates the global dynamic parameters that determine the stability of the system and the uncertainty of the estimate. The modified algorithm of the ellipsoid of optimal volume (MOVE), is a method applied to the parametric identification of systems, in particular to the estimate of groups of parameters (PSE for their initials in English). It is looked for to obtain the ellipsoid of smaller volume that guarantees to contain the real value of the parameters of the model. The PSE MOVE is a recursive identification method that can manage the sign of noise and to ponder it, the ellipsoid represents an advantage due to its easy mathematical handling in the computer, the results that surrender are very useful for the design of Robust Control since to smaller volume of the ellipsoid, better is in general the performance of the system to control. The comparison with other methods presented in the literature to estimate the reason of decline (DR) of a BWR is presented. (Author)

  12. Fuel composition optimization in a 78-element fuel bundle for use in a pressure tube type supercritical water-cooled reactor

    International Nuclear Information System (INIS)

    A 78-element fuel bundle containing a plutonium-thorium fuel mixture has been proposed for a Generation IV pressure tube type supercritical water-cooled reactor. In this work, using a lattice cell model created with the code DRAGON,the lattice pitch, fuel composition (fraction of PuO2 in ThO2) and radial enrichment profile of the 78-element bundle is optimized using a merit function and a metaheuristic search algorithm.The merit function is designed such that the optimal fuel maximizes fuel utilization while minimizing peak element ratings and coolant void reactivity. A radial enrichment profile of 10 wt%, 11 wt% and 20 wt% PuO2 (inner to outer ring) with a lattice pitch of 25.0 cm was found to provide the optimal merit score based on the aforementioned criteria. (author)

  13. Comparison of counterpart test results with the VISTA-ITL and FESTA on SBLOCA scenarios for an integral type reactor SMART

    International Nuclear Information System (INIS)

    Counterpart test results were compared for small break loss-of-coolant accident (SBLOCA) scenarios of shutdown cooling system (SCS) and pressurizer safety valve (PSV) line breaks for an integral type reactor SMART. As counter-part tests for an SBLOCA for the SMART design, two integral effect test facilities of VISTA-ITL and FESTA were used, and their results were compared to better understand the phenomena expected to occur in the SMART design. The initial and boundary conditions were appropriately provided for the tests, and the overall trend of the major thermal-hydraulic parameters showed reasonable results. Although there are minor differences between the tests results from VISTA-ITL and FESTA due to their different scales, it is considered that they provide reasonable thermal-hydraulic behaviors against SMART during the SBLOCA scenarios. Therefore, these two IET facilities, VISTA-ITL and FESTA, can be used together to simulate the thermal-hydraulic behavior of the SMART design. (author)

  14. Reduced-scale water test of natural circulation for decay heat removal in loop-type sodium-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Murakami, T., E-mail: murakami@criepi.denken.or.jp [Central Research Institute of Electric Power Industry, 1646 Abiko, Chiba (Japan); Eguchi, Y., E-mail: eguchi@criepi.denken.or.jp [Central Research Institute of Electric Power Industry, 1646 Abiko, Chiba (Japan); Oyama, K., E-mail: kazuhiro_oyama@mfbr.mhi.co.jp [Mitsubishi FBR Systems, Inc., 2-34-17 Jinguumae, Shibuya, Tokyo (Japan); Watanabe, O., E-mail: osamu4_watanabe@mfbr.mhi.co.jp [Mitsubishi FBR Systems, Inc., 2-34-17 Jinguumae, Shibuya, Tokyo (Japan)

    2015-07-15

    Highlights: • The natural circulation characteristics of a loop-type SFR are examined by a water test. • The performance of decay heat removal system is evaluated using a similarity law. • The effects of flow deviation in the parallel piping of a primary loop are clarified. • The reproducibility of the natural circulation test is confirmed. - Abstract: Water tests of a loop-type sodium-cooled fast reactor have been conducted to physically evaluate the natural circulation characteristics. The water test apparatus was manufactured as a 1/10-scale mock-up of the Japan Sodium-Cooled Fast Reactor, which adopts a decay heat removal system (DHRS) utilizing natural circulation. Tests simulating a variety of events and operation conditions clarified the thermal hydraulic characteristics and core-cooling performance of the natural circulation in the primary loop. Operation conditions such as the duration of the pump flow coast-down and the activation time of the DHRS affect the natural circulation characteristics. A long pump flow coast-down cools the upper plenum of the reactor vessel (RV). This causes the loss of the buoyant force in the RV. The test result indicates that a long pump flow coast-down tends to result in a rapid increase in the core temperature because of the loss of the buoyant force. The delayed activation of the DHRS causes a decrease in the natural circulation flow rate and a temperature rise in the RV. Flow rate deviation and a reverse flow appear in the parallel cold-leg piping in some events, which cause thermal stratification in the cold-leg piping. The DHRS prevents the core temperature from fatally rise even for the most severe design-basis event, in which sodium leakage in a secondary loop of the DHRS and the opening failure of a single damper of the air cooler occur simultaneously. In the water test for the case of siphon break in the primary loop, which is one of the design extension conditions, a circulation flow consisting of ascendant

  15. Prokaryotic homologs of Argonaute proteins are predicted to function as key components of a novel system of defense against mobile genetic elements

    Directory of Open Access Journals (Sweden)

    van der Oost John

    2009-08-01

    Full Text Available Abstract Background In eukaryotes, RNA interference (RNAi is a major mechanism of defense against viruses and transposable elements as well of regulating translation of endogenous mRNAs. The RNAi systems recognize the target RNA molecules via small guide RNAs that are completely or partially complementary to a region of the target. Key components of the RNAi systems are proteins of the Argonaute-PIWI family some of which function as slicers, the nucleases that cleave the target RNA that is base-paired to a guide RNA. Numerous prokaryotes possess the CRISPR-associated system (CASS of defense against phages and plasmids that is, in part, mechanistically analogous but not homologous to eukaryotic RNAi systems. Many prokaryotes also encode homologs of Argonaute-PIWI proteins but their functions remain unknown. Results We present a detailed analysis of Argonaute-PIWI protein sequences and the genomic neighborhoods of the respective genes in prokaryotes. Whereas eukaryotic Ago/PIWI proteins always contain PAZ (oligonucleotide binding and PIWI (active or inactivated nuclease domains, the prokaryotic Argonaute homologs (pAgos fall into two major groups in which the PAZ domain is either present or absent. The monophyly of each group is supported by a phylogenetic analysis of the conserved PIWI-domains. Almost all pAgos that lack a PAZ domain appear to be inactivated, and the respective genes are associated with a variety of predicted nucleases in putative operons. An additional, uncharacterized domain that is fused to various nucleases appears to be a unique signature of operons encoding the short (lacking PAZ pAgo form. By contrast, almost all PAZ-domain containing pAgos are predicted to be active nucleases. Some proteins of this group (e.g., that from Aquifex aeolicus have been experimentally shown to possess nuclease activity, and are not typically associated with genes for other (putative nucleases. Given these observations, the apparent extensive

  16. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2001-04-01

    The Reactor Physics and MYRRHA 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 of reactor fuel. This expertise is applied in the 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 by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised.

  17. Reactor safeguards

    CERN Document Server

    Russell, Charles R

    2013-01-01

    Reactor Safeguards provides information for all who are interested in the subject of reactor safeguards. Much of the material is descriptive although some sections are written for the engineer or physicist directly concerned with hazards analysis or site selection problems. The book opens with an introductory chapter on radiation hazards, the construction of nuclear reactors, safety issues, and the operation of nuclear reactors. This is followed by separate chapters that discuss radioactive materials, reactor kinetics, control and safety systems, containment, safety features for water reactor

  18. Reactor operation

    CERN Document Server

    Shaw, J

    2013-01-01

    Reactor Operation covers the theoretical aspects and design information of nuclear reactors. This book is composed of nine chapters that also consider their control, calibration, and experimentation.The opening chapters present the general problems of reactor operation and the principles of reactor control and operation. The succeeding chapters deal with the instrumentation, start-up, pre-commissioning, and physical experiments of nuclear reactors. The remaining chapters are devoted to the control rod calibrations and temperature coefficient measurements in the reactor. These chapters also exp

  19. The improvement of control rod in experimental fast reactor JOYO. The development of a sodium bonded type control rod

    Energy Technology Data Exchange (ETDEWEB)

    Soga, T.; Miyakawa, S.; Mitsugi, T. [Japan Nuclear Cycle Development Inst., Oarai Engineering Center, Irradiation Center, Irradiation and Administration Section, Oarai, Ibaraki (Japan)

    1999-06-01

    Currently, the lifetime of control rods in JOYO is limited by Absorber-Cladding Mechanical Interaction (ACMI) due to swelling of B{sub 4}C(boron carbide) pellets accelerated by relocation of pellet fragments. A sodium bonded type control rod was developed which improves the thermal conductivity by means of charging sodium into the gap between B{sub 4}C and cladding and by utilizing a shroud which wraps the pellet fragments in a thin tube. This new design will be able to enlarge the gap between B{sub 4}C and cladding, without heating B{sub 4}C or fragment relocation, thus extending the life of the control rod. The sodium bonded type will be fabricated as the ninth reload control rods in JOYO. (1) The specification of a sodium bonded type control rod was determined with the wide gap between B{sub 4}C and cladding. In the design simulation, main component temperature were below the maximum limit. And the local heating by helium bubble generated from B{sub 4}C in the sodium gap, was not a serious problem in the analysis which was considered. (2) A structural design for the sodium entrance into the pin was determined. A formula was developed which the limit for sodium charging given physical dimension of the structure and sodium property. Result from sodium out-pile experiments validated the theoretical formula. (3) The analysis of ACMI indicated a lifetime extension of the sodium bonded type by 4.6% in comparison with lifetime of the helium bonded type of 1.6%. This is due to the boron10 burn-up rate being three times higher in the sodium bonded type than in the helium bonded type. To achieve a target burn-up 10% in the future, it will be necessary to modify design based on irradiation data which will be obtained by practical use of the sodium bonded control rods in JOYO. (4) The effects due to Absorber-Cladding Chemical Interaction (ACCI) were reduced by controlling the cladding temperature and chromium coating to the cladding's inner surface. It was confirmed

  20. Honeycomb reactor washcoated with mordenite type zeolite catalysts for the reduction of NO{sub x} by NH{sub 3}

    Energy Technology Data Exchange (ETDEWEB)

    Choi, H.; Ham, S.W.; Nam, I.S.; Kim, Y.G. [Pohang Univ. of Science and Technology (Korea, Republic of)]|[Research Inst. of Industrial Science and Technology, Pohang (Korea, Republic of)

    1996-01-01

    A low pressure drop reactor was prepared by washcoating Cu ion-exchange mordenite on a honeycomb. The reactor configuration including reaction conditions was experimentally optimized both for low-pressure drop and for high catalytic activity of the honeycomb reactor. Over 90% of NO conversion was achieved by both Cu ion-exchanged synthetic zeolite (CuHM) and natural zeolite (CuNZA). The pressure drops due to honeycomb reactors were low enough to meet the constraint of the pressure drop (5 in. of H{sub 2}O) for its application to a utility boiler. A mathematical model based on fluid flow, reaction kinetics, and pressure drop was derived for the design of the reactor, and then the reactor was simulated to examine the effects of operating conditions on NO conversion. Finally, the sulfur tolerance of the honeycomb reactor developed in this work has also been investigated.

  1. Study on shielding design method of radiation streaming in a tokamak-type DT fusion reactor based on Monte Carlo calculation

    International Nuclear Information System (INIS)

    In tokamak-type DT nuclear fusion reactor, there are various type slits and ducts in the blanket and the vacuum vessel. The helium production in the rewelding location of the blanket and the vacuum vessel, the nuclear properties in the super-conductive TF coil, e.g. the nuclear heating rate in the coil winding pack, are enhanced by the radiation streaming through the slits and ducts, and they are critical concern in the shielding design. The decay gamma ray dose rate around the duct penetrating the blanket and the vacuum vessel is also enhanced by the radiation streaming through the duct, and they are also critical concern from the view point of the human access to the cryostat during maintenance. In order to evaluate these nuclear properties with good accuracy, three dimensional Monte Carlo calculation is required but requires long calculation time. Therefore, the development of the effective simple design evaluation method for radiation streaming is substantially important. This study aims to establish the systematic evaluation method for the nuclear properties of the blanket, the vacuum vessel and the Toroidal Field (TF) coil taking into account the radiation streaming through various types of slits and ducts, based on three dimensional Monte Carlo calculation using the MNCP code, and for the decay gamma ray dose rates penetrated around the ducts. The present thesis describes three topics in five chapters as follows; 1) In Chapter 2, the results calculated by the Monte Carlo code, MCNP, are compared with those by the Sn code, DOT3.5, for the radiation streaming in the tokamak-type nuclear fusion reactor, for validating the results of the Sn calculation. From this comparison, the uncertainties of the Sn calculation results coming from the ray-effect and the effect due to approximation of the geometry are investigated whether the two dimensional Sn calculation can be applied instead of the Monte Carlo calculation. Through the study, it can be concluded that the

  2. Transmutation of actinides in power reactors.

    Science.gov (United States)

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

    2005-01-01

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

  3. Post-processor for simulations of the ORIGEN program and calculation of the composition of the activity of a burnt fuel core by a BWR type reactor

    International Nuclear Information System (INIS)

    The composition calculation and the activity of nuclear materials subject to processes of burnt, irradiation and decay periods are of utility for diverse activities inside the nuclear industry, as they are it: the processes design and operations that manage radioactive material, the calculation of the inventory and activity of a core of burnt nuclear fuel, for studies of type Probabilistic Safety Analysis (APS), as well as for regulation processes and licensing of nuclear facilities. ORIGEN is a program for computer that calculates the composition and the activity of nuclear materials subject to periods of burnt, irradiation and decay. ORIGEN generates a great quantity of information whose processing and analysis are laborious, and it requires thoroughness to avoid errors. The automation of the extraction, conditioning and classification of that information is of great utility for the analyst. By means of the use of the post-processor presented in this work it is facilitated, it speeds up and wide the capacity of analysis of results, since diverse consultations with several classification options and filtrate of results can be made. As illustration of the utility of the post-processor, and as an analysis of interest for itself, it is also presented in this work the composition of the activity of a burned core in a BWR type reactor according to the following classification criteria: by type of radioisotope (fission products, activation products and actinides), by specie type (gassy, volatile, semi-volatile and not volatile), by element and by chemical group. The results show that the total activity of the studied core is dominated by the fission products and for the actinides, in proportion four to one, and that the gassy and volatile species conform a fifth part of the total activity of the core. (Author)

  4. Target-induced nano-enzyme reactor mediated hole-trapping for high-throughput immunoassay based on a split-type photoelectrochemical detection strategy.

    Science.gov (United States)

    Zhuang, Junyang; Tang, Dianyong; Lai, Wenqiang; Xu, Mingdi; Tang, Dianping

    2015-09-15

    Photoelectrochemical (PEC) detection is an emerging and promising analytical tool. However, its actual application still faces some challenges like potential damage of biomolecules (caused by itself system) and intrinsic low-throughput detection. To solve the problems, herein we design a novel split-type photoelectrochemical immunoassay (STPIA) for ultrasensitive detection of prostate specific antigen (PSA). Initially, the immunoreaction was performed on a microplate using a secondary antibody/primer-circular DNA-labeled gold nanoparticle as the detection tag. Then, numerously repeated oligonucleotide sequences with many biotin moieties were in situ synthesized on the nanogold tag via RCA reaction. The formed biotin concatamers acted as a powerful scaffold to bind with avidin-alkaline phosphatase (ALP) conjugates and construct a nanoenzyme reactor. By this means, enzymatic hydrolysate (ascorbic acid) was generated to capture the photogenerated holes in the CdS quantum dot-sensitized TiO2 nanotube arrays, resulting in amplification of the photocurrent signal. To elaborate, the microplate-based immunoassay and the high-throughput detection system, a semiautomatic detection cell (installed with a three-electrode system), was employed. Under optimal conditions, the photocurrent increased with the increasing PSA concentration in a dynamic working range from 0.001 to 3 ng mL(-1), with a low detection limit (LOD) of 0.32 pg mL(-1). Meanwhile, the developed split-type photoelectrochemical immunoassay exhibited high specificity and acceptable accuracy for analysis of human serum specimens in comparison with referenced electrochemiluminescence immunoassay method. Importantly, the system was not only suitable for the sandwich-type immunoassay mode, but also utilized for the detection of small molecules (e.g., aflatoxin B1) with a competitive-type assay format. PMID:26291091

  5. Target-induced nano-enzyme reactor mediated hole-trapping for high-throughput immunoassay based on a split-type photoelectrochemical detection strategy.

    Science.gov (United States)

    Zhuang, Junyang; Tang, Dianyong; Lai, Wenqiang; Xu, Mingdi; Tang, Dianping

    2015-09-15

    Photoelectrochemical (PEC) detection is an emerging and promising analytical tool. However, its actual application still faces some challenges like potential damage of biomolecules (caused by itself system) and intrinsic low-throughput detection. To solve the problems, herein we design a novel split-type photoelectrochemical immunoassay (STPIA) for ultrasensitive detection of prostate specific antigen (PSA). Initially, the immunoreaction was performed on a microplate using a secondary antibody/primer-circular DNA-labeled gold nanoparticle as the detection tag. Then, numerously repeated oligonucleotide sequences with many biotin moieties were in situ synthesized on the nanogold tag via RCA reaction. The formed biotin concatamers acted as a powerful scaffold to bind with avidin-alkaline phosphatase (ALP) conjugates and construct a nanoenzyme reactor. By this means, enzymatic hydrolysate (ascorbic acid) was generated to capture the photogenerated holes in the CdS quantum dot-sensitized TiO2 nanotube arrays, resulting in amplification of the photocurrent signal. To elaborate, the microplate-based immunoassay and the high-throughput detection system, a semiautomatic detection cell (installed with a three-electrode system), was employed. Under optimal conditions, the photocurrent increased with the increasing PSA concentration in a dynamic working range from 0.001 to 3 ng mL(-1), with a low detection limit (LOD) of 0.32 pg mL(-1). Meanwhile, the developed split-type photoelectrochemical immunoassay exhibited high specificity and acceptable accuracy for analysis of human serum specimens in comparison with referenced electrochemiluminescence immunoassay method. Importantly, the system was not only suitable for the sandwich-type immunoassay mode, but also utilized for the detection of small molecules (e.g., aflatoxin B1) with a competitive-type assay format.

  6. Reactor container

    International Nuclear Information System (INIS)

    Purpose: To prevent shocks exerted on a vent head due to pool-swell caused within a pressure suppression chamber (disposed in a torus configuration around the dry well) upon loss of coolant accident in BWR type reactors. Constitution: The following relationship is established between the volume V (m3) of a dry well and the ruptured opening area A (m2) at the boundary expected upon loss of coolant accident: V >= 30340 (m) x A Then, the volume of the dry well is made larger than the ruptured open area, that is, the steam flow rate of leaking coolants upon loss of coolant accident to decrease the pressure rise in the dry well at the initial state where loss of coolant accident is resulted. Accordingly, the pressure of non-compressive gases jetted out from the lower end of the downcomer to the pool water is decreased to suppress the pool-swell. (Ikeda, J.)

  7. Mathematical model analysis on the enhancement of aeration efficiency using ladder-type flat membrane module forms in the Submerged Membrane Bio-reactor (SMBR)

    Institute of Scientific and Technical Information of China (English)

    LI Bo; YE MaoSheng; YANG FengLin; MA Hui

    2009-01-01

    The cross-flow shearing action produced from the inferior aeration in the Submerged Membrane Bio-reactor (SMBR) Is an effective way to further improve anti-fouling effects of membrane modules.Based on the widely-applied vertical structure of flat membrane modules, improvements are made that ladder-type flat membrane structure is designed with a certain inclined angle θ so that the cross-flow velocity of bubble near the membrane surface can be held, and the intensity and times of elastic colli-sion between bubbles and membrane surface can be increased. This can improve scouring action ofmembrane surface on aeration and reduce energy consumption of strong aeration in SMBR. By de-ducing and improving the mathematics model of collision between bubble and vertical flat put forward by Vries, the relatively suitable Incline angle θ under certain aeration place and in certain size rang ofbubble can be obtained with the computer iterative calculation technology. Finally, for many groups of ladder-type flat membrane in parallel placement in the practical application of SMBR, some sugges-tions are offered: the interval distance of membrane modules is 8--15 mm, and aeration should be op-erated at 5--7 mm among membrane modules, and the optimal design angle of trapeziform membrane is 1.7°--2.5°.

  8. Mathematical model analysis on the enhancement of aeration efficiency using ladder-type flat membrane module forms in the Submerged Membrane Bio-reactor(SMBR)

    Institute of Scientific and Technical Information of China (English)

    2009-01-01

    The cross-flow shearing action produced from the inferior aeration in the Submerged Membrane Bio-reactor(SMBR) is an effective way to further improve anti-fouling effects of membrane modules.Based on the widely-applied vertical structure of flat membrane modules,improvements are made that ladder-type flat membrane structure is designed with a certain inclined angle θ so that the cross-flow velocity of bubble near the membrane surface can be held,and the intensity and times of elastic colli-sion between bubbles and membrane surface can be increased.This can improve scouring action of membrane surface on aeration and reduce energy consumption of strong aeration in SMBR.By de-ducing and improving the mathematics model of collision between bubble and vertical flat put forward by Vries,the relatively suitable incline angle θ under certain aeration place and in certain size rang of bubble can be obtained with the computer iterative calculation technology.Finally,for many groups of ladder-type flat membrane in parallel placement in the practical application of SMBR,some sugges-tions are offered:the interval distance of membrane modules is 8―15 mm,and aeration should be op-erated at 5―7 mm among membrane modules,and the optimal design angle of trapeziform membrane is 1.7°―2.5°.

  9. Modular simulation of the dynamics of a 925 MWe PWR electronuclear type reactor and design of a multivariable control algorithm

    International Nuclear Information System (INIS)

    This work has been consecrated to the modular simulation of a PWR 925 MWe power plant's dynamic and to the design of a multivariable algorithm control: a mathematical model of a plant type was developed. The programs were written on a structured manner in order to maximize flexibility. A multivariable control algorithm based on pole placement with output feedback was elaborated together with its correspondent program. The simulation results for different normal transients were shown and the capabilities of the new method of multivariable control are illustrated through many examples

  10. Development of the Approach by States method and thermodynamical study of a 1300 MWe PWR type reactor following a complete water loss of vapor generator alimentation with the Cathare 2 code

    International Nuclear Information System (INIS)

    The objective of this report is to study the thermohydraulic behavior of a 1300 MWe PWR type reactor for a complete loss accident in water supplying of vapor generators. The Cathare computer code has been used in this aim. (N.C.)

  11. Research reactors

    International Nuclear Information System (INIS)

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

  12. Jet-Stirred Reactors

    OpenAIRE

    Herbinet, Olivier; Guillaume, Dayma

    2013-01-01

    The jet-stirred reactor is a type of ideal continuously stirred-tank reactor which is well suited for gas phase kinetic studies. It is mainly used to study the oxidation and the pyrolysis of hydrocarbon and oxygenated fuels. These studies consist in recording the evolution of the conversion of the reactants and of the mole fractions of reaction products as a function of different parameters such as reaction temperature, residence time, pressure and composition of the inlet gas. Gas chromatogr...

  13. Reactor physics and reactor computations

    International Nuclear Information System (INIS)

    Mathematical methods and computer calculations for nuclear and thermonuclear reactor kinetics, reactor physics, neutron transport theory, core lattice parameters, waste treatment by transmutation, breeding, nuclear and thermonuclear fuels are the main interests of the conference

  14. The safety of light water reactors

    International Nuclear Information System (INIS)

    The book describes the principles and practices of reactor safety as applied to the design, regulation and operation of both pressurized water reactors and boiling water reactors. The central part of the book is devoted to methods and results of safety analysis. Some significant events are described, notably the Three Mile Island accident. The book concludes with a chapter on the PIUS principle of inherent reactor safety as applied to the SECURE type of reactor developed in Sweden. (G.B.)

  15. Analysis of higher power research reactors' parameters

    International Nuclear Information System (INIS)

    The objective of this monograph was to analyze and compare parameters of different types of research reactors having higher power. This analysis could be used for decision making and choice of a reactor which could possibly replace the existing ageing RA reactor in Vinca. Present experimental and irradiation needs are taken into account together with the existing reactors operated in our country, RB and TRIGA reactor

  16. Qualification and interpretation of MR test reactor irradiation data on WWER-440 type fuel rods for fuel thermal model validation

    International Nuclear Information System (INIS)

    Conducted jointly by the Kurchatov Institute of Atomic Energy of the USSR and the Finnish state owned power utility Imatran Voima Oy, an irradiation program, SOFIT, including a series of characterised and instrumented test rod bundles is in progress to examine the thermal and mechanical behaviour of WWER type fuel rods. The main purposes of the work are to promote the flexibility and the economy of the fuel utilisation, and to verify, for the WWER annular fuel and Zr1%Nb cladding, the fuel behaviour computer programs used in the Soviet Union and in Finland. The more specific objectives are to obtain information on fuel temperature, internal gas pressure and cladding elongation during irradiation for different fabrication parameters. Qualification and interpretation of in-pile results from the first irradiation, completed at maximum local exposure of about 16 MWd/kgU, are presented. Quantitative evaluation of affecting phenomena could be deduced from the restored thermocouple readings. The measured temperatures show dependency on the as-fabricated gap size and weakness on the helium fill pressure. The results of the calculations, obtained by the existing steady state fuel behaviour computer codes are compared with the experimental temperatures. The importance of careful qualification of the data is emphasised. The related uncertainties are discussed. Preliminary fission gas data of some of the rods have become available, which show low or moderate gas release. (author). 5 refs, 7 figs, 1 tab

  17. Assessment of the thorium fuel cycle in power reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    1977-01-01

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

  18. Evolution of weld metals nanostructure and properties under irradiation and recovery annealing of VVER-type reactors

    Science.gov (United States)

    Gurovich, B.; Kuleshova, E.; Shtrombakh, Ya.; Fedotova, S.; Zabusov, O.; Prikhodko, K.; Zhurko, D.

    2013-03-01

    The results of VVER-440 steel Sv-10KhMFT and VVER-1000 steel SV-10KhGNMAA investigations by transmission electron microscopy, scanning electron microscopy, Auger-electron spectroscopy and mechanical tests are presented in this paper. The both types of weld metals with different content of impurities and alloying elements were studied after irradiations to fast neutron (E > 0.5 MeV) fluences in the wide range below and beyond the design values, after recovery annealing procedures and after re-irradiation following the annealing. The distinctive features of embrittlement kinetics of VVER-440 and VVER-1000 RPV weld metals conditioned by their chemical composition differences were investigated. It is shown that the main contribution into radiation strengthening within the design fluence can be attributed to radiation-induced precipitates, on reaching the design or beyond design values of fast neutron fluencies the main contribution into VVER-440 welds strengthening is made by radiation-induced dislocation loops, and in case of VVER-1000 welds - radiation-induced precipitates and grain-boundary phosphorous segregations. Recovery annealing of VVER-440 welds at 475 °C during 100 h causes irradiation-induced defects disappearance, transformation of copper enriched precipitates into bigger copper-rich precipitates with lower number density and leads to almost full recovery of mechanical properties followed by comparatively slow re-embrittlement rate. The recovery annealing temperature of VVER-1000 welds was higher - 565 °C during 100 h - to avoid temper brittleness. The annealing of VVER-1000 welds leads to almost full recovery of mechanical properties due to irradiation-induced defects disappearance and decrease in precipitates number density and grain-boundary segregation of phosphorus. The re-embrittlement rate of VVER-1000 weld during subsequent re-irradiation is at least not higher than the initial rate.

  19. Evolution of weld metals nanostructure and properties under irradiation and recovery annealing of VVER-type reactors

    International Nuclear Information System (INIS)

    Highlights: ► At operating fluences radiation strengthening of VVER is specified by precipitates. ► Grain-boundary segregation level in VVER-1000 weld increases during the operation. ► Recovery annealing decreases grain-boundary segregation level in VVER-1000 weld. ► Re-embrittlement of VVER-1000 weld is at least not higher than primary embrittlement. ► Re-embrittlement of VVER-440 weld is lower than primary embrittlement. -- Abstract: The results of VVER-440 steel Sv-10KhMFT and VVER-1000 steel SV-10KhGNMAA investigations by transmission electron microscopy, scanning electron microscopy, Auger-electron spectroscopy and mechanical tests are presented in this paper. The both types of weld metals with different content of impurities and alloying elements were studied after irradiations to fast neutron (E > 0.5 MeV) fluences in the wide range below and beyond the design values, after recovery annealing procedures and after re-irradiation following the annealing. The distinctive features of embrittlement kinetics of VVER-440 and VVER-1000 RPV weld metals conditioned by their chemical composition differences were investigated. It is shown that the main contribution into radiation strengthening within the design fluence can be attributed to radiation-induced precipitates, on reaching the design or beyond design values of fast neutron fluencies the main contribution into VVER-440 welds strengthening is made by radiation-induced dislocation loops, and in case of VVER-1000 welds – radiation-induced precipitates and grain-boundary phosphorous segregations. Recovery annealing of VVER-440 welds at 475 °C during 100 h causes irradiation-induced defects disappearance, transformation of copper enriched precipitates into bigger copper-rich precipitates with lower number density and leads to almost full recovery of mechanical properties followed by comparatively slow re-embrittlement rate. The recovery annealing temperature of VVER-1000 welds was higher – 565

  20. Design of Liner Type Double Heated Biomass Pyrolysis Reactor%新型内胆式双热型生物质热解反应器的设计

    Institute of Scientific and Technical Information of China (English)

    李三平; 王述洋; 孙雪; 曹有为

    2014-01-01

    Nowadays the cost of the device for preparing bio-oil is very high with low durability. On the other hand, the reactor usually needs exogenous heating. This results in high energy consumption in preparation of bio-oil clean fuel. On the basis of the existing types of fluidized bed reactor design experience, this paper presented a calculation model for designing fluidized biomass pyrolysis reactor. By using the model, a kind of liner type double heated biomass pyrolysis reactor is designed in this paper. Then the main parameters of the reactor are optimized using VB. The optimal reactor is bed diameter 0. 221 m with the height 1.445 m.%为了解决目前生物质制生物燃油清洁燃料装置造价高、需要外源燃料供热、寿命低及生物质制生物燃油清洁燃料能耗高等问题,结合各类流化床反应器的设计经验公式,提出一种流化式生物质热解主反应器设计的计算模型;利用该模型研究并设计了一种内胆式双热型生物质热解主反应器;利用VB对主反应器的主要参数进行了优化设计,得到其主要设计参数为床径为0.221 m,床高为1.445 m,并通过冷态实验进行了验证。

  1. Scientific-technical cooperation with Russia. Transient analyses for alternative types of water-cooled reactors. Final report; WTZ mit Russland. Transientenanalysen fuer wassergekuehlte Kernreaktoren. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Rohde, Ulrich [Forschungszentrum Dresden-Rossendorf (Germany). Inst. fuer Sicherheitsforschung; Kozmenkov, Yaroslav [Forschungszentrum Dresden-Rossendorf (Germany). Inst. fuer Sicherheitsforschung; Institute of Physics and Power Engineering, Obninsk (Russian Federation); Pivovarov, Valeri; Matveev, Yurij [Institute of Physics and Power Engineering, Obninsk (Russian Federation)

    2010-12-15

    The recently developed multi-group version DYN3D-MG of the reactor dynamics code DYN3D has been qualified for applications to water-cooled reactor concepts different from industrial PWR and BWR. An extended DYN3D version was applied to the graphite-moderated pressure tube reactor EGP-6 (NPP Bilibino) and conceptual design studies of an advanced Boiling Water Reactor with reduced moderation (RMWR) as well as the RUTA-70 reactor for low temperature heat supply. Concerning the RUTA reactor, safe heat removal by natural circulation of the coolant at low pressure has to be shown. For the corresponding validation of thermo-hydraulic system codes like ATHLET and RELAP5, experiments on flashing-induced natural circulation instabilities performed at the CIRCUS test facility at the TU Delft were simulated using the RELAP5 code. For the application to alternative water-cooled reactors, DYN3D model extensions and modifications were implemented, in particular adaptations of heat conduction and heat transfer models. Performing code-to-code comparisons with the Russian fine-mesh neutron diffusion code ACADEM contributed to the verification of DYN3D-MG. Validation has been performed by calculating reactor dynamics experiments at the NPP Bilibino. For the reactors EGP-6, RMWR and RUTA, analyses of various protected and unprotected control rod withdrawal and ejection transients were performed. The beyond design basis accident (BDBA) scenario ''Coast-down of all main coolant pumps at nominal power without scram'' for the RUTA reactor was analyzed using the code complexes DYN3D/ATHLET and DYN3D/RELAP5. It was shown, that the reactor passes over to a save asymptotic state at reduced power with coolant natural circulation. Analyzing the BDBA ''Unprotected withdrawal of a control rod group'' for the RMWR, the safety against Departure from Nucleate Boiling (DNB) could not be shown with the necessary confidence. Finally, conclusions have been drawn

  2. Experience with Kamini reactor

    International Nuclear Information System (INIS)

    Kamini is a 233U fuelled, 30 kW(th) research reactor. It is one of the best neutron source facility with a core average flux of 1012 n/cm2/s in IGCAR used for neutron radiography of active and nonradioactive objects, activation analysis and radiation physics research. The core consists of nine plate type fuel elements with a total fuel inventory of 590 g of 233U. Two safety control plates made of cadmium are used for start up and shutdown of the reactor. Three beam tubes, two-thimble irradiation site outside reflector and one irradiation site nearer to the core constitute the testing facilities of Kamini. Kamini attained first criticality on 29th October 96 and nominal power of 30 kW in September 1997. This paper covers the design features of the reactor, irradiation facilities and their utilities and operating experience of the reactor. (author)

  3. Hydrogen Production in Fusion Reactors

    OpenAIRE

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

    1993-01-01

    As one of methods of innovative energy production in fusion reactors without having a conventional turbine-type generator, an efficient use of radiation produced in a fusion reactor with utilizing semiconductor and supplying clean fuel in a form of hydrogen gas are studied. Taking the candidates of reactors such as a toroidal system and an open system for application of the new concepts, the expected efficiency and a concept of plant system are investigated.

  4. Reactor building

    International Nuclear Information System (INIS)

    The whole reactor building is accommodated in a shaft and is sealed level with the earth's surface by a building ceiling, which provides protection against penetration due to external effects. The building ceiling is supported on walls of the reactor building, which line the shaft and transfer the vertical components of forces to the foundations. The thickness of the walls is designed to withstand horizontal pressure waves in the floor. The building ceiling has an opening above the reactor, which must be closed by cover plates. Operating equipment for the reactor can be situated above the building ceiling. (orig./HP)

  5. Decay profiles of {beta} and {gamma} for a radionuclide inventory in equilibrium cycle of a BWR type reactor; Perfiles de decaimiento de radiacion {beta} y {gamma} para un inventario de radionuclidos en ciclo de equilibrio de un reactor tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Salaices, M.; Sandoval, S.; Ovando, R. [Instituto de Investigaciones Electricas. Gerencia de Energia Nuclear, Av. Reforma 113 Col. Palmira. 62490 Cuernavaca, Morelos (Mexico)]. e-mail: sal@iie.org.mx

    2007-07-01

    Presently work the {beta} and {gamma} radiation decay profiles for a radionuclides inventory in equilibrium cycle of a BWR type reactor is presented. The profiles are presented in terms of decay in the activity of the total inventory as well as of the chemical groups that conform the inventory. In the obtaining of the radionuclides inventory in equilibrium cycle the ORIGEN2 code, version 1 was used, which simulates fuel burnup cycles and it calculates the evolution of the isotopic composition as a result of the burnt one, irradiation and decay of the nuclear fuel. It can be observed starting from the results that the decrease in the activity for the initial inventory and the different chemical groups that conform it is approximately proportional to the base 10 logarithm of the time for the first 24 hours of having concluded the burnt one. It can also be observed that the chemical groups that contribute in more proportion to the total activity of the inventory are the lanthanides-actinides and the transition metals, with 39% and 28%, respectively. The groups of alkaline earth metals, halogens, metalloids, noble gases and alkaline metals, contribute with percentages that go from the 8 to 5%. The groups that less they contribute to the total activity of the inventory they are the non metals and semi-metals with smaller proportions that 1%. The chemical groups that more contribute to the energy of {beta} and {gamma} radiation its are the transition metals and the lanthanides-actinides with a change in the order of importance at the end of the 24 hours period. The case of the halogens is of relevance for the case of the {gamma} radiation energy due that occupying the very near third site to the dimensions of the two previous groups. Additionally, the decay in the activity for the total inventory and the groups that conform it can be simulated by means of order 6 polynomials or smaller than describe its behavior appropriately. The results presented in this work, coupled

  6. Estimation of fast neutron fluence in steel specimens type Laguna Verde in TRIGA Mark III reactor; Estimacion de la fluencia de neutrones rapidos en probetas de acero tipo Laguna Verde en el reactor Triga Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Galicia A, J.; Francois L, J. L. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico); Aguilar H, F., E-mail: blink19871@hotmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2015-09-15

    The main purpose of this work is to obtain the fluence of fast neutrons recorded within four specimens of carbon steel, similar to the material having the vessels of the BWR reactors of the nuclear power plant of Laguna Verde when subjected to neutron flux in a experimental facility of the TRIGA Mark III reactor, calculating an irradiation time to age the material so accelerated. For the calculation of the neutron flux in the specimens was used the Monte Carlo code MCNP5. In an initial stage, three sheets of natural molybdenum and molybdenum trioxide (MoO{sub 3}) were incorporated into a model developed of the TRIGA reactor operating at 1 M Wth, to calculate the resulting activity by setting a certain time of irradiation. The results obtained were compared with experimentally measured activities in these same materials to validate the calculated neutron flux in the model used. Subsequently, the fast neutron flux received by the steel specimens to incorporate them in the experimental facility E-16 of the reactor core model operating at nominal maximum power in steady-state was calculated, already from these calculations the irradiation time required was obtained for values of the neutron flux in the range of 10{sup 18} n/cm{sup 2}, which is estimated for the case of Laguna Verde after 32 years of effective operation at maximum power. (Author)

  7. Research reactors: design, safety requirements and applications

    International Nuclear Information System (INIS)

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

  8. A new loss-of-function allele 28y reveals a role of ARGONAUTE1 in limiting asymmetric division of stomatal lineage ground cell

    Institute of Scientific and Technical Information of China (English)

    Kezhen Yangy; Min Jiangy; Jie Le

    2014-01-01

    In Arabidopsis thaliana L., stomata are produced through a series of divisions including asymmetric and symmetric divisions. Asymmetric entry division of meristemoid mother cellproduces two daughter cells, the smal er meristemoid and the larger sister cell, a stomatal lineage ground cell(SLGC). Stomatal lineage ground cells can differentiate into epidermal pavement cells but have the potential to divide asymmetrical y, spacing divisions, to create satel ite meristemoids. Peptide ligands and TOO MANY MOUTHS (TMM) and ERECTA family receptors regulate the initiation of stomatal lineages, activity, and orientation of spacing divisions. Here, we reported that a natural mutant 28y displayed an increased stomatal density and index. Using map-based cloning, we identified mutation in ARGONAUTE1 (AGO1) as the cause of 28y phenotypes. Time-lapse tracing of stomatal lineage cells reveals that stomatal overproduction in 28y is caused by the excessive asymmetric spacing division of SLGCs.Further genetic results demonstrated that AGO1 acts down-stream of TMM and negatively regulates the SPCH transcripts, but in a brassinosteroid-independent manner. Upregulation of AGAMOUS-LIKE16 (AGL16) in 28y mutants suggests that AGO1 is required to restrict AGL16-mediated stomatal spacing divisions, an miRNA pathway in addition to ligand-receptor signaling modules.

  9. Markov State Models Reveal a Two-Step Mechanism of miRNA Loading into the Human Argonaute Protein: Selective Binding followed by Structural Re-arrangement

    KAUST Repository

    Jiang, Hanlun

    2015-07-16

    Argonaute (Ago) proteins and microRNAs (miRNAs) are central components in RNA interference, which is a key cellular mechanism for sequence-specific gene silencing. Despite intensive studies, molecular mechanisms of how Ago recognizes miRNA remain largely elusive. In this study, we propose a two-step mechanism for this molecular recognition: selective binding followed by structural re-arrangement. Our model is based on the results of a combination of Markov State Models (MSMs), large-scale protein-RNA docking, and molecular dynamics (MD) simulations. Using MSMs, we identify an open state of apo human Ago-2 in fast equilibrium with partially open and closed states. Conformations in this open state are distinguished by their largely exposed binding grooves that can geometrically accommodate miRNA as indicated in our protein-RNA docking studies. miRNA may then selectively bind to these open conformations. Upon the initial binding, the complex may perform further structural re-arrangement as shown in our MD simulations and eventually reach the stable binary complex structure. Our results provide novel insights in Ago-miRNA recognition mechanisms and our methodology holds great potential to be widely applied in the studies of other important molecular recognition systems.

  10. A Flexible Domain-Domain Hinge Promotes an Induced-fit Dominant Mechanism for the Loading of Guide-DNA into Argonaute Protein in Thermus thermophilus

    KAUST Repository

    Zhu, Lizhe

    2016-02-24

    Argonaute proteins (Ago) are core components of the RNA Induced Silencing Complex (RISC) that load and utilize small guide nucleic acids to silence mRNAs or cleave foreign DNAs. Despite the essential role of Ago in gene regulation and defense against virus, the molecular mechanism of guide-strand loading into Ago remains unclear. We explore such a mechanism in the bacterium Thermus thermophilus Ago (TtAgo), via a computational approach combining molecular dynamics, bias-exchange metadynamics, and protein-DNA docking. We show that apo TtAgo adopts multiple closed states that are unable to accommodate guide-DNA. Conformations able to accommodate the guide are beyond the reach of thermal fluctuations from the closed states. These results suggest an induced-fit dominant mechanism for guide-strand loading in TtAgo, drastically different from the two-step mechanism for human Ago 2 (hAgo2) identified in our previous study. Such a difference between TtAgo and hAgo2 is found to mainly originate from the distinct rigidity of their L1-PAZ hinge. Further comparison among known Ago structures from various species indicates that the L1-PAZ hinge may be flexible in general for prokaryotic Agos but rigid for eukaryotic Agos. © 2016 American Chemical Society.

  11. Salt Stress Reveals a New Role for ARGONAUTE1 in miRNA Biogenesis at the Transcriptional and Posttranscriptional Levels1[OPEN

    Science.gov (United States)

    Niedojadlo, Katarzyna; Niedojadlo, Janusz; Walczak, Weronika; Szweykowska-Kulinska, Zofia; Jarmolowski, Artur

    2016-01-01

    Plants as sessile organisms have developed prompt response mechanisms to react to rapid environmental changes. In addition to the transcriptional regulation of gene expression, microRNAs (miRNAs) are key posttranscriptional regulators of the plant stress response. We show here that the expression levels of many miRNAs were regulated under salt stress conditions. This regulation occurred at the transcriptional and posttranscriptional levels. During salinity stress, the levels of miRNA161 and miRNA173 increased, while the expression of pri-miRNA161 and pri-miRNA173 was down-regulated. Under salt stress conditions, miRNA161 and miRNA173 were stabilized in the cytoplasm, and the expressions of MIR161 and MIR173 were negatively regulated in the nucleus. ARGONAUTE1 (AGO1) participated in both processes. We demonstrated that AGO1 cotranscriptionally controlled the expression of MIR161 and MIR173 in the nucleus. Our results suggests that AGO1 interacts with chromatin at MIR161 and MIR173 loci and causes the disassembly of the transcriptional complex, releasing short and unpolyadenylated transcripts. PMID:27385819

  12. Reactor simulator development. Workshop material

    International Nuclear Information System (INIS)

    The International Atomic Energy Agency (IAEA) has established a programme in nuclear reactor simulation computer programs to assist its Member States in education and training. The objective is to provide, for a variety of advanced reactor types, insight and practice in reactor operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the supply or development of simulation programs and training material, sponsors training courses and workshops, and distributes documentation and computer programs. This publication consists of course material for workshops on development of such reactor simulators. Participants in the workshops are provided with instruction and practice in the development of reactor simulation computer codes using a model development system that assembles integrated codes from a selection of pre-programmed and tested sub-components. This provides insight and understanding into the construction and assumptions of the codes that model the design and operational characteristics of various power reactor systems. The main objective is to demonstrate simple nuclear reactor dynamics with hands-on simulation experience. Using one of the modular development systems, CASSIMtm , a simple point kinetic reactor model is developed, followed by a model that simulates the Xenon/Iodine concentration on changes in reactor power. Lastly, an absorber and adjuster control rod, and a liquid zone model are developed to control reactivity. The built model is used to demonstrate reactor behavior in sub-critical, critical and supercritical states, and to observe the impact of malfunctions of various reactivity control mechanisms on reactor dynamics. Using a PHWR simulator, participants practice typical procedures for a reactor startup and approach to criticality. This workshop material consists of an introduction to systems used for developing reactor simulators, an overview of the dynamic simulation

  13. Reactor water spontaneous circulation structure in reactor pressure vessel

    International Nuclear Information System (INIS)

    The gap between the inner wall of a reactor pressure vessel of a BWR type reactor and a reactor core shroud forms a down comer in which reactor water flows downwardly. A feedwater jacket to which feedwater at low temperature is supplied is disposed at the outer circumference of the pressure vessel just below a gas/water separator. The reactor water at the outer circumferential portion just below the air/water separator is cooled by the feedwater jacket, and the feedwater after cooling is supplied to the feedwater entrance disposed below the feedwater jacket by way of a feedwater introduction line to supply the feedwater to the lower portion of the down comer. This can cool the reactor water in the down comer to increase the reactor water density in the down comer thereby forming strong downward flows and promote the recycling of the reactor water as a whole. With such procedures, the reactor water can be recycled stably only by the difference of the specific gravity of the reactor water without using an internal pump. In addition, the increase of the height of the pressure vessel can be suppressed. (I.N.)

  14. Chernobyl reactor transient simulation study

    International Nuclear Information System (INIS)

    This paper deals with the Chernobyl nuclear power station transient simulation study. The Chernobyl (RBMK) reactor is a graphite moderated pressure tube type reactor. It is cooled by circulating light water that boils in the upper parts of vertical pressure tubes to produce steam. At equilibrium fuel irradiation, the RBMK reactor has a positive void reactivity coefficient. However, the fuel temperature coefficient is negative and the net effect of a power change depends upon the power level. Under normal operating conditions the net effect (power coefficient) is negative at full power and becomes positive under certain transient conditions. A series of dynamic performance transient analysis for RBMK reactor, pressurized water reactor (PWR) and fast breeder reactor (FBR) have been performed using digital simulator codes, the purpose of this transient study is to show that an accident of Chernobyl's severity does not occur in PWR or FBR nuclear power reactors. This appears from the study of the inherent, stability of RBMK, PWR and FBR under certain transient conditions. This inherent stability is related to the effect of the feed back reactivity. The power distribution stability in the graphite RBMK reactor is difficult to maintain throughout its entire life, so the reactor has an inherent instability. PWR has larger negative temperature coefficient of reactivity, therefore, the PWR by itself has a large amount of natural stability, so PWR is inherently safe. FBR has positive sodium expansion coefficient, therefore it has insufficient stability it has been concluded that PWR has safe operation than FBR and RBMK reactors

  15. Construction of an external electrode for determination of electrochemical corrosion potential in normal operational conditions of an BWR type reactor for hot cells; Construccion de un electrodo externo para determinacion del potencial de corrosion electroquimico en condiciones normales de operacion de un reactor tipo BWR para celdas calientes

    Energy Technology Data Exchange (ETDEWEB)

    Aguilar T, J.A.; Rivera M, H.; Hernandez C, R. [Departamento de Sintesis y Caracterizacion de Materiales, Instituto Nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    2001-07-01

    The behavior of the corrosion processes at high temperature requires of external devices that being capable to resist a temperature of 288 Centigrade and a pressure of 80 Kg/cm{sup 2}, to give stable and reproducible results of some variable and resisting physically and chemically the radiation. The external electrode of Ag/AgCl fulfils all the requirements in the determination of the electrochemical corrosion potential under normal operational conditions of a BWR type reactor in hot cells. (Author)

  16. NEUTRONIC REACTOR FUEL ELEMENT

    Science.gov (United States)

    Picklesimer, M.L.; Thurber, W.C.

    1961-01-01

    A chemically nonreactive fuel composition for incorporation in aluminum- clad, plate type fuel elements for neutronic reactors is described. The composition comprises a mixture of aluminum and uranium carbide particles, the uranium carbide particles containing at least 80 wt.% UC/sub 2/.

  17. Reactors. Nuclear propulsion ships

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  19. Behavior of antimony isotopes in the primary coolant of WWER-1000-type nuclear reactors in NPP Kozloduy during operation and shutdown

    Energy Technology Data Exchange (ETDEWEB)

    Dobrevski, Ivan D.; Zaharieva, Neli N. [Bulgarian Academy of Sciences, Sofia (Bulgaria). Inst. for Nuclear Research and Nuclear Energy; Minkova, Katia F.; Gerchev, Nikolay B. [Kozloduy Nuclear Power Plant, Kozloduy (Bulgaria). Dept. of Chemistry and Radiochemistry

    2009-05-15

    This paper focuses on the behavior of the antimony isotopes {sup 122}Sb and {sup 124}Sb in the coolant of the WWER reactors in the nuclear power plant Kozloduy (Bulgaria) during operation and shutdown. It is concluded that the chemical properties of their actual precursor, the isotope {sup 121}Sb, determine the behavior of {sup 122}Sb and {sup 124}Sb during operation, load fluctuations, and shutdown as well as during the reactor coolant purification process. It is supposed that differences between the reactor bulk and the core fuel cladding surface chemistry as well as the presence of sub-cooled nucleate boiling at the fuel cladding may create conditions under which a local oxidizing environment may come into existence. (orig.)

  20. NEUTRONIC REACTOR

    Science.gov (United States)

    Anderson, H.L.

    1960-09-20

    A nuclear reactor is described comprising fissionable material dispersed in graphite blocks, helium filling the voids of the blocks and the spaces therebetween, and means other than the helium in thermal conductive contact with the graphite for removing heat.

  1. Chemical Reactors.

    Science.gov (United States)

    Kenney, C. N.

    1980-01-01

    Describes a course, including content, reading list, and presentation on chemical reactors at Cambridge University, England. A brief comparison of chemical engineering education between the United States and England is also given. (JN)

  2. Monolithic reactor: higher yield, less energy

    OpenAIRE

    Mols, B.

    2004-01-01

    The production of margarine, the desulphurisation of crude oil, and the manufacture of synthetic diesel fuel, these are only three of the many industrial processes in which a three-phase reactor is used. Traditionally, this type of reactor is rather ill-defined. Success with a lab scale set-up is no guarantee that a large commercial reactor will work. Scalability is less than perfect, one might say. Researchers at the Reactor & Catalysis Engineering epartment of the Chemical Technology facult...

  3. Reactor safety in Eastern Europe. Proceedings

    International Nuclear Information System (INIS)

    The papers presented to the GRS colloquium refer to the cooperative activities for reactor accident analysis and modification of the GRS computer codes for their application to reactors of the Russian design types of WWER or RBMK. Another topic is the safety of RBMK reactors in particular, and the current status of investigations and studies addressing the containment of unit 4 of the Chernobyl reactor station. (HP)

  4. Dynamic calculations of pressurized water reactor internals

    International Nuclear Information System (INIS)

    A mathematical model is briefly described for the calculation of oscillations in the WWER-440 reactor internals. The model was developed for improved safety of the type of reactors. It allows calculating vibrations resistance of reactor components, mainly during accidents, such as loss of coolant accidents. Some results are given of the calculation of forces acting in the rupture of the reactor inlet and outlet pipes. (Z.M.)

  5. Analysis of some antecipated transients without scram for PWR type reactors by coupling of the CORAN code to the ALMOD code system

    International Nuclear Information System (INIS)

    This study investigates some antecipated transients without scram for a pressurized water cooled reactor, using coupling of the containment CORAN code to the ALMOD code system, under severe random conditions. This coupling has the objective of including containment model as part of an unified code system. These severe conditions include failure of reactor scram, following a station black-out and emergency power initiation for the burn-up status at the beginning and end of the cycle. Furthermore, for the burn-up status at the end of the cycle, a failure in the closure of the pressurizer relief valve was also investigated. (Author)

  6. NUCLEAR REACTOR

    Science.gov (United States)

    Anderson, C.R.

    1962-07-24

    A fluidized bed nuclear reactor and a method of operating such a reactor are described. In the design means are provided for flowing a liquid moderator upwardly through the center of a bed of pellets of a nentron-fissionable material at such a rate as to obtain particulate fluidization while constraining the lower pontion of the bed into a conical shape. A smooth circulation of particles rising in the center and falling at the outside of the bed is thereby established. (AEC)

  7. Nuclear reactor

    International Nuclear Information System (INIS)

    In order to reduce neutron embrittlement of the pressue vessel of an LWR, blanked off elements are fitted at the edge of the reactor core, with the same dimensions as the fuel elements. They are parallel to each other, and to the edge of the reactor taking the place of fuel rods, and are plates of neutron-absorbing material (stainless steel, boron steel, borated Al). (HP)

  8. PROJECT-SPECIFIC TYPE A VERIFICATION FOR THE HIGH FLUX BEAM REACTOR UNDERGROUND UTILITIES REMOVAL PHASE 3 TRENCH 5, BROOKHAVEN NATIONAL LABORATORY UPTON, NEW YORK

    International Nuclear Information System (INIS)

    The Oak Ridge Institute for Science and Education (ORISE) has reviewed the project documentation and data for the High Flux Beam Reactor (HFBR) Underground Utilities removal Phase 3; Trench 5 at Brookhaven National Laboratory (BNL) in Upton, New York. The Brookhaven Survey Group (BSG) has completed removal and performed Final Status Survey (FSS) of the concrete duct from Trench 5 from Building 801 to the Stack. Sample results have been submitted as required to demonstrate that the cleanup goal of (le)15 mrem/yr above background to a resident in 50 years has been met. Four rounds of sampling, from pre-excavation to FSS, were performed as specified in the Field Sampling Plan (FSP) (BNL 2010a). It is the policy of the U.S. Department of Energy (DOE) to perform independent verifications of decontamination and decommissioning activities conducted at DOE facilities. ORISE has been designated as the organization responsible for this task for the HFBR Underground Utilities. ORISE, together with DOE, determined that a Type A verification of Trench 5 was appropriate based on recent verification results from Trenches 2, 3, and 4, and the minimal potential for residual radioactivity in the area. The removal of underground utilities is being performed in three stages to decommission the HFBR facility and support structures. Phase 3 of this project included the removal of at least 200 feet of 36-inch to 42-inch pipe from the west side to the south side of Building 801, and the 14-inch diameter Acid Waste Line that spanned from 801 to the Stack within Trench 5. Based on the pre-excavation sample results of the soil overburden the potential for contamination of the soil surrounding the pipe is minimal (BNL 2010a). ORISE reviewed the BNL FSP and identified comments for consideration (ORISE 2010). BNL prepared a revised FSP that resolved each ORISE comment adequately (BNL 2010a). ORISE referred to the revised HFBR Underground Utilities FSP FSS data to conduct the Type A verification

  9. Second-generation thermionic reactor (TRICE)

    International Nuclear Information System (INIS)

    During the past decade, advanced types of thermionic converters have been investigated and experimentally demonstrated. In addition to their substantially increased efficiency, these advanced (second-generation) converters also can condition the output power such as to remove many design constraints on the performance and complexity of the in-core thermionic reactor. A second-generation thermionic in-core reactor concept (TRICE) is described in which advanced-mode thermionic converters are integrated into a reactor core, and their output inductively-coupled to an external high-voltage AC electromagnetic output circuit. Because the electrical insulators and intercell gaps that are required in the flashlight type of in-core reactor are absent from the TRICE core, the size and complexity of the TRICE reactor are substantially less than for a flashlight type reactor of the same output power. The increased performance potentially available via the TRICE concept is estimated for a variety of reactor output power levels

  10. Operating experiences of the research reactors

    International Nuclear Information System (INIS)

    Nuclear research reactors are devices of wide importance, being used for different scientific research tasks, for testing and improving reactor systems and components, for the production of radioisotopes, for the purposes of defence, for staff training and for other purposes. There are three research reactors in Yugoslavia: RA, RB and TRIGA. Reactors RA and RB at the 'Boris Kidric' Institute of Nuclear Sciences are of heavy water type power being 6500 and 10 kW, and maximum thermal neutron flux of 1014 and 1011(n/cm2s), respectively. TRIGA reactor at the 'Jozef Stefan' Institute in Ljubljana is of 250 kW power and maximum thermal neutron flux of 1013(n/cm2s). Reactors RA and RB use soviet fuel in the form of uranium dioxide (80% enriched) and metallic uranium (2%). Besides, RB reactor operates with natural uranium too. TRIGA reactor uses american uranium fuel 70% and 20% enriched, uranium being mixed homogeneously with moderator (ZrH). Experiences in handling and controlling the fuel before irradiation in the reactor, in reactor and after it are numerous and valuable, involving either the commercial arrangements with foreign producers, or optimal burn up in reactor or fuel treatment after the reactor irradiation. Twenty years of operating experience of these reactors have great importance especially having in mind the number of trained staff. Maintenance of reactors systems and fluids in continuous operation is valuable experience from the point of view of water reactor utilization. The case of the RA reactor primary cycle cobalt decontamination and other events connected with nuclear and radiation security for all three reactors are also specially emphasized. Owing to our research reactors, numerous theoretical, numerical and experimental methods are developed for nuclear and other analyses and design of research and power reactors,as well as methods for control and protection of radiation. (author)

  11. Validation of WIMS-SNAP code systems for calculations in TRIGA-MARK II type reactors; Validacion del sistema de codigos WIMS-SNAP para calculos en reactores nucleares tipo TRIGA-MARK II

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez Valle, S.; Lopez Aldama, D. [Centro de Investigaciones Nucleares, Tecnologicas y Ambientales, La Habana (Cuba). E-mail: svalle@ctn.isctn.edu.cu

    2000-07-01

    The following paper contributes to validate the Nuclear Engineering Department methods to carry out calculations in TRIGA reactors solving a Benchmark. The benchmark is analyzed with the WIMS-D/4-SNAP/3D code system and using the cross section library WIMS-TRIGA. A brief description of the DSN method is presented used in WIMS/d{sup 4} code and also the SNAP-3d code is shortly explained. The results are presented and compared with the experimental values. In other hand the possible error sources are analyzed. (author)

  12. Backfitting swimming pool reactors

    International Nuclear Information System (INIS)

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

  13. Reactor operation experience

    International Nuclear Information System (INIS)

    Since the TRIGA Users Conference in Helsinki 1970 the TRIGA reactor Vienna was in operation without any larger undesired shutdown. The integrated thermal power production by August 15 1972 accumulated to 110 MWd. The TRIGA reactor is manly used for training of students, for scientific courses and research work. Cooperation with industry increased in the last two years either in form of research or in performing training courses. Close cooperation is also maintained with the IAEA, samples are irradiated and courses on various fields are arranged. Maintenance work was performed on the heat exchanger and to replace the shim rod magnet. With the view on the future power upgrading nine fuel elements type 110 have been ordered recently. Experiments, performed currently on the reactor are presented in details

  14. Maintenance operation by divers on a swimming-pool type reactor (Osiris, CEN Saclay). Technical and medical prevention: an example of multidisciplinary ergonomic step

    International Nuclear Information System (INIS)

    Maintenance works in a swimming-pool reactor was performed by a team of divers. A multidisciplinary ergonomic study had previously defined the working procedure. The ergonomic approach is analysed. The divers' working techniques are described. After work, medical tests showed that previsions were verified and proved the methods as safe. This technique by divers' interventions should open new possibilities in nuclear industry

  15. Generalities about nuclear reactors

    International Nuclear Information System (INIS)

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

  16. Analysis of Microbial Communities in Biofilms from CSTR-Type Hollow Fiber Membrane Biofilm Reactors for Autotrophic Nitrification and Hydrogenotrophic Denitrification.

    Science.gov (United States)

    Shin, Jung-Hun; Kim, Byung-Chun; Choi, Okkyoung; Kim, Hyunook; Sang, Byoung-In

    2015-10-01

    Two hollow fiber membrane biofilm reactors (HF-MBfRs) were operated for autotrophic nitrification and hydrogenotrophic denitrification for over 300 days. Oxygen and hydrogen were supplied through the hollow fiber membrane for nitrification and denitrification, respectively. During the period, the nitrogen was removed with the efficiency of 82-97% for ammonium and 87-97% for nitrate and with the nitrogen removal load of 0.09-0.26 kg NH4(+)-N/m(3)/d and 0.10-0.21 kg NO3(-)-N/m(3)/d, depending on hydraulic retention time variation by the two HF-MBfRs for autotrophic nitrification and hydrogenotrophic denitrification, respectively. Biofilms were collected from diverse topological positions in the reactors, each at different nitrogen loading rates, and the microbial communities were analyzed with partial 16S rRNA gene sequences in denaturing gradient gel electrophoresis (DGGE). Detected DGGE band sequences in the reactors were correlated with nitrification or denitrification. The profile of the DGGE bands depended on the NH4(+) or NO3(-) loading rate, but it was hard to find a major strain affecting the nitrogen removal efficiency. Nitrospira-related phylum was detected in all biofilm samples from the nitrification reactors. Paracoccus sp. and Aquaspirillum sp., which are an autohydrogenotrophic bacterium and an oligotrophic denitrifier, respectively, were observed in the denitrification reactors. The distribution of microbial communities was relatively stable at different nitrogen loading rates, and DGGE analysis based on 16S rRNA (341f /534r) could successfully detect nitrate-oxidizing and hydrogen-oxidizing bacteria but not ammonium-oxidizing bacteria in the HF-MBfRs.

  17. Experimental studies of local coolant hydrodynamics using a scaled model of cassette-type fuel assembly of a KLT-40S reactor

    Science.gov (United States)

    Dmitriev, S. M.; Barinov, A. A.; Varentsov, A. V.; Doronkov, D. V.; Solntsev, D. N.; Khrobostov, A. E.

    2016-08-01

    The results of experimental studies of local hydrodynamic and mass exchange characteristics of the coolant flow behind the spacer grid in the fuel assembly of a KLT-40S reactor are presented. The experiments were aimed at the investigation of representative domains of the fuel assembly with three tracer injection regions. The studies were performed at the aerodynamic test facility using the tracer gas diffusion method. According to the theory of hydrodynamic similarity, the obtained experimental results can be transferred to full-scale coolant flow conditions in standard fuel assemblies. The analysis of the tracer concentration propagation made it possible to determine in detail the flow pattern and find the main regularities and specific features of the coolant flow behind the plate spacer grid of KLT-40S fuel assembly. The hydraulic resistance coefficient of the spacer grid was experimentally determined. The coefficients of mass exchange between cells for representative cells of the displacer region in the KLT-40S fuel assembly were calculated for the first time; these results are presented in the form of the "mixing matrix." The results of studies of local coolant flow hydrodynamics in the KLT-40S fuel assembly are used at AO Afrikantov OKBM for estimation of thermotechnical reliability of active cores for reactors of floating nuclear power stations. The experimental data on hydrodynamic and mass exchange characteristics are included in the database for verification of CDF codes and detailed cell-wise calculation of the active core for KLT-40S reactor installation. The results of these studies can be used at FSUE RFNC-VNIIEF for testing and verification of domestic three-dimensional hydrodynamic CFD codes ("Logos") that are applied for substantiation of newly designed reactor installations. Practical recommendations on the application of the obtained results in thermohydraulic calculations of the active core for the KLT-40S reactor will be worked out. Proposals

  18. Research reactors: a tool for science and medicine

    International Nuclear Information System (INIS)

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

  19. Progress of China Experimental Fast Reactor in 2011

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

    1 Background Fast reactor is the reactor which realized the chain fission with fast neutron.As an optional type of generation Ⅳ reactor,fast reactor has three characters:1) It can change 238U to 239Pu and raise the uranium resource utilization

  20. Cooling system for reactor container

    International Nuclear Information System (INIS)

    Purpose: To effectively cool a reactor container upon reactor shutdown with no intrusion of metal corrosion products in coolants into the main steam pipe in a BWR type reactor. Constitution: A clean up system comprising a pipeway, a recycling pump, a non-regenerative heat exchanger and a primary coolant purifier and a regenerative heat exchanger is provided branched from a residual heat removing system and the clean up system is connected by way of a valve to a feedwater pipeway, as well as connected by way of the pipeway to the main steam pipeway at the midway of two main steam separation valves outside of the reactor container. This enables to prevent metal corrosion products floating on the surface of reactor water from introducing into the main steam pipe when the pressure vessel is filled with water. Then, since the pressure vessel is filled with primary coolants, the pressure vessel can be cooled uniformly in a short time. (Ikeda, J.)

  1. Antineutrino Monitoring of Thorium Reactors

    CERN Document Server

    Akindele, Oluwatomi A; Norman, Eric B

    2015-01-01

    Various groups have demonstrated that antineutrino monitoring can be successful in assessing the plutonium content in water-cooled nuclear reactors for nonproliferation applications. New reactor designs and concepts incorporate nontraditional fuels types and chemistry. Understanding how these properties affect the antineutrino emission from a reactor can extend the applicability of antineutrino monitoring.Thorium molten salt reactors (MSR) breed U-233, that if diverted constitute an IAEA direct use material. The antineutrino spectrum from the fission of U-233 has been determined, the feasibility of detecting the diversion of a significant quantity, 8 kg of U-233, within the IAEA timeliness goal of 30 days has been evaluated. The antineutrino emission from a thorium reactor operating under normal conditions is compared to a diversion scenario at a 25 meter standoff by evaluating the daily antineutrino count rate and the energy spectrum of the detected antineutrinos. It was found that the diversion of a signifi...

  2. Analysis of an accident type sbloca in reactor contention AP1000 with 8.0 Gothic code; Analisis de un accidente tipo Sbloca en la contencion del reactor AP1000 con el codigo Gothic 8.0

    Energy Technology Data Exchange (ETDEWEB)

    Goni, Z.; Jimenez Varas, G.; Fernandez, K.; Queral, C.; Montero, J.

    2016-08-01

    The analysis is based on the simulation of a Small Break Loss-of-Coolant-Accident in the AP1000 nuclear reactor using a Gothic 8.0 tri dimensional model created in the Science and Technology Group of Nuclear Fision Advanced Systems of the UPM. The SBLOCA has been simulated with TRACE 5.0 code. The main purpose of this work is the study of the thermo-hydraulic behaviour of the AP1000 containment during a SBLOCA. The transients simulated reveal close results to the realistic behaviour in case of an accident with similar characteristics. The pressure and temperature evolution enables the identification of the accident phases from the RCS point of view. Compared to the licensing calculations included in the AP1000 Safety Analysis, it has been proved that the average pressure and temperature evolution is similar, yet lower than the licensing calculations. However, the temperature and inventory distribution are significantly heterogeneous. (Author)

  3. Stability analysis of a recycling circuit of a BWR type reactor. Theoretical study; Analisis de estabilidad de un circuito de recirculacion de un reactor del tipo BWR. Estudio teorico

    Energy Technology Data Exchange (ETDEWEB)

    Salinas H, J.G.; Espinosa P, G. [Universidad Autonoma Metropolitana-Iztapalapa, 09000 Mexico D.F. (Mexico); Gonzalez M, V.M. [Comision Nacional de Seguridad Nuclear y Salvaguardias, 04000 Mexico D.F. (Mexico)

    2000-07-01

    The Technology, Regulation and Services Management of the National Commission of Nuclear Safety and Safeguards financed and in coordinate form with the I.P.H. Department of the Metropolitan Autonomous-Iztapalapa University developed the present project with the purpose of studying the effect of the recycling system on the linear stability of a BWR reactor whose reference central is the Laguna Verde power station. The present project forms part of a work series focused to the linear stability of the nuclear reactor of the Unit 1 at Laguna Verde power station. The components of the recycling system considered for the study of stability are the recycling external circuit (recycling pumps, valves) and the internal circuit (downcomer, jet pumps, lower full, driers, separators). The mathematical model is obtained applying mass balances and movement quantity in each one of the mentioned circuits. With respect to the nucleus model two regions are considered, the first one is made of a flow in one phase and the second one of a flow in two phases. For modelling the biphasic region it is considered homogenous flow. Generally it is studied the system behavior in the frequency domain starting from the transfer function applied to four operational states which correspond to the lower stability zone in the map power-flow of the Unit 1 of Laguna Verde power station. The Nyquist diagrams corresponding to each state as well as their characteristic frequency were determined. The results show that exists a very clear dependence of the power-flow relation on the stability of the system. It was found that the boiling length is an important parameter for the linear stability of the system. The obtained results show that the characteristic frequencies in unstability zones are similar to the reported data of the Unit 1 of the Laguna Verde power station in the event of power oscillations carried out in January 1995. (Author)

  4. Development of a neutronic model for the fuel of a high temperature gas reactor type PBMR; Desarrollo de un modelo neutronico para el combustible de un reactor de gas de alta temperatura tipo PBMR

    Energy Technology Data Exchange (ETDEWEB)

    Oropeza C, I.; Carmona H, R.; Francois L, J. L. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac 8532, Jiutepec, Morelos 62550 (Mexico)]. e-mail: ivonucci@prodigy.net.mx

    2008-07-01

    In this work was developed the neutronic model of a fuel sphere of a nuclear reactor of gas of high temperature to modulate of bed of spheres (PBMR), using the Monte Carlo method with the MCNPx code. In order to be able to verify the fuel model constructed in this investigation, it is used a case of reference, based on an international exercise {sup b}enchmark{sup .} The benchmark report contains the results sent by different international participants for five phases with respect to the high temperature gas reactor (HTR), fed with uranium, plutonium and thorium. In particular, in first stage of benchmark an infinite adjustment of uranium compound fuel spheres is considered unique, with which our results were compared. This first stage considers two cases: cell calculations with spherical external frontier and cell calculations with cubic external frontier. The objective is to identify any increase in the uncertainty, related to the uranium fuel, that is associated with the plutonium and thorium fuels. In order to validate our results, the values of the neutron multiplication factor were taken in account, in cold and in the heat of the moment from the participants who sent their results obtained with Monte Carlo and deterministic calculations. The model of the fuel sphere developed in this work considers a regular distribution of 15000 Triso particles, in a cubic mesh centered within the sphere. For it was necessary to define the step firstly or {sup p}itch{sup o}f the cubic mesh. Generally, the results obtained by the participants of benchmark and those of this investigation present good agreement, nevertheless, appear some discrepancies, attributed to factors like different libraries of cross sections used, the nature of the solution: Monte Carlo or deterministic, and the difficulty of some participants to model the external frontier condition of reflection. (Author)

  5. Analysis of Discharge Fault in 35 kV Dry-type Air-Core Reactor%一起35 kV干式空心电抗器放电故障分析

    Institute of Scientific and Technical Information of China (English)

    张宁; 李洪伟

    2014-01-01

    本文介绍了一起500 kV变电站35 kV干式空心电抗器在运行当中发生放电烧损的故障情况,通过现场检查、试验,结合故障电抗器的解体检查结果,对故障原因进行了深入分析,发现故障的主要原因是由于在强磁场下涡流产生温升,破坏了电抗器本身绝缘,本文对防止同类故障的发生具有一定的借鉴意义。%In this paper, a fault of burn-out of 35 kV dry-type air-core reactor due to discharge in the operation at 500 kV substation is introduced. The reason of the fault is analyzed through site inspec-tion, test and disassembly inspection result of the faulty reactor. It is caused by high temperature rise caused by eddy current at strong magnetic field resulting to insulation damage of the reactor it-self. The analysis result has a definite reference to similar fault to be prevented.

  6. 不同挡流板形式紫外线消毒仪杀菌效果模拟%Numerical simulation of sterilizing efficiency of ultraviolet disinfection reactors with different flow baffle types

    Institute of Scientific and Technical Information of China (English)

    牛培平; 丁日升; 宋卫堂; 王媛

    2015-01-01

    Nutrient solution recycling has become one of the essential techniques of soilless cultivation. But nutrient solutions are susceptible to be polluted by infectious diseases during the recycling process, so it is needed to disinfect nutrient solutions before recycling use. Compared to other disinfection methods of common nutrient solutions, ultraviolet (UV) disinfection has many advantages, such as high efficiency, low cost, not changing the physical and chemical properties of nutrient solutions, so UV disinfection is an environment-friendly technology of nutrient solution treatment.In general, experimental research and numerical simulation are the most common methods of UV disinfection. Although experimental results of the performance of UV disinfection reactor are credible, direct measurement is difficult and costly and thus seldom done. On the other hand, one can use numerical simulation techniques to model the UV disinfection. Computational fluid dynamics (CFD) has been widely used for simulating the UV disinfection. In previous studies, some researchers analyzed the performance of small-scale horizontal UV disinfection reactors with different flow baffle numbers and flow areas. They designed an alternately arranged flat-type flow baffle and analyzed the performance of the UV disinfection device for nutrient solutions using the CFD simulations and the measurements of biological bacterial disinfection. However, the effects of different flow baffle types on the performance of the UV disinfection reactor have seldom been investigated. In this study, we designed 5 different flow baffle types of the UV disinfection reactor. The main objectives were to increase disinfection efficiency by optimizing the flow baffle arrangement and to test the performance of the UV disinfection reactor with different flow baffle types. The 5 different flow baffle types were proposed: 2 circular channels and annulus alternation, 2 circular channels, 4 circular channels and annulus

  7. Study and characterization of noble metal deposits on similar rusty surfaces to those of the reactor U-1 type BWR of nuclear power station of Laguna Verde

    International Nuclear Information System (INIS)

    In the present investigation work, were determined the parameters to simulate the conditions of internal oxidation reactor circulation pipes of the nuclear power plant of Laguna Verde in Veracruz. We used 304l stainless steel cylinders with two faces prepared with abrasive paper of No. 600, with the finality to obtain similar surface to the internal circulation piping nuclear reactor. Oxides was formed within an autoclave (Autoclave MEX-02 unit B), which is a device that simulates the working conditions of the nuclear reactor, but without radiation generated by the fission reaction within the reactor. The oxidation conditions were a temperature of 280 C and pressure of 8 MPa, similar conditions to the reactor operating in nuclear power plant of Laguna Verde in Veracruz, Mexico (BWR conditions), with an average conductivity of 4.58 ms / cm and 2352 ppb oxygen to simulate normal water chemistry NWC. Were obtained deposits of noble metal oxides formed on 304l stainless steel samples, in a 250 ml autoclave at a temperature range of 180 to 200 C. The elements that were used to deposit platinum-rhodium (Pt-Rh) with aqueous Na2Pt (OH)6 and Na3Rh (NO2)6, Silver (Ag) with an aqueous solution of AgNO3, zirconium (Zr) with aqueous Zr O (NO3) and ZrO2, and zinc (Zn) in aqueous solution of Zn (NO3)2 under conditions of normal water chemistry. Also there was the oxidation of 304l stainless steel specimens in normal water chemistry with a solution of Zinc (Zn) (NWC + Zn). Oxidation of the specimens in water chemistry with a solution of zinc (Zn + NWC) was prepared in two ways: within the MEX-02 autoclave unit A in a solution of zinc and a flask at constant temperature in zinc solution. The oxides formed and deposits were characterized by scanning electron microscopy, energy dispersive X-ray analysis, elemental field analysis and X-ray diffraction. By other hand was evaluated the electrochemical behavior of the oxides formed on the surface of 304l stainless steel in normal water

  8. Test and Coupling Calculation of Temperature Field for UHV Dry-Type Air-Core Smoothing Reactor%特高压干式空心平波电抗器温度场耦合计算与试验

    Institute of Scientific and Technical Information of China (English)

    姜志鹏; 文习山; 王羽; 陈瑞珍; 曹继丰; 陈图腾

    2015-01-01

    为了研究特高压干式空心平波电抗器的温升分布特性,该文基于计算流体力学和传热学理论,建立了电抗器稳态流体与固体耦合温度场的数学计算模型.采用有限容积法对三维模型进行稳态流体场与温度场直接求解,获得其温度场分布特性,研究了包封轴向及径向温度分布规律.最后采用光纤测温法对自然对流下的电抗器进行温升测量.对比分析表明,计算与试验结果吻合较好,验证温度场数值计算的合理性和准确性,为特高压干式空心平波电抗器温升监测提供参考.%To research the distribution characteristics of temperature rise for UHV dry-type air-core smoothing reactor, according to computational fluid dynamics and heat transfer theory, this paper presented the mathematical model of temperature field coupling steady fluid and solid for the reactor. The finite volume method was employed to solve the steady flow and temperature fields of 3D model directly, and the temperature distribution characteristics of the reactor were obtained. Then the axial and radial temperature distributions of encapsulations were studied separately. Finally, optical fiber temperature measurement method was used to test temperature rise for the reactor under natural convection condition. Comparative analysis shows that the calculated results are in good agreement with the experiment, which verifies the rationality and accuracy of the temperature field numerical calculation. And it can provide references for the temperature rise monitoring of UHV dry-type air-core smoothing reactor.

  9. DeveIopment of interactive safety anaIysis program for pooI type sodium cooIed fast reactor%池式钠冷快堆交互式安全分析软件开发

    Institute of Scientific and Technical Information of China (English)

    钱鸿涛; 李政昕; 胡文军; 宫宇

    2015-01-01

    为建立适用于池式钠冷快堆的仿真机,开发了基于法国快堆系统分析程序 OASIS 的交互式安全分析系统,实现了实时绘图、动态显示等可视化功能。利用该系统模拟了中国实验快堆的堆芯、主热传输系统、事故余热排出系统,以及控制调节系统和保护系统,分析了各个功率台阶的稳态及满功率下流量阶跃瞬态工况。分析结果与设计值符合度良好,表明该系统具有良好的适用性,可用于人员培训与安全审评等。%An interactive safety analysis program was developed and integrated into the simulation system for pool type sodium cooled fast reactor based on a French fast reactor system analysis code OASIS.The visualized functions of real-time plotting and dynamic display were provided.The core,main power transfer system,decay heat removal system,control and regulation system and reactor protection system of China Experimental Fast Reactor were simulated by the system.The various power level steady states and the transient of flow step at full power state were analyzed.The calculation results match well with the design data.It can be indicated that the program had a good applicability,and can be used for personnel training and safety review.

  10. Study on the design of the new type of flocculation reactor and its performances%新型絮凝反应器的设计及应用性能研究

    Institute of Scientific and Technical Information of China (English)

    刘兴旺; 杨运泉

    2009-01-01

    A new type of flocculation reactor has been designed, whose performances have been evaluated by its residence time distribution. The new reactor's D/UL is 0.10, showing that it is close to plug flow reactor. Its flocculation efficiency is high. The new reactor is used for treating wastewater in iron and steel works. Dynamic tests are conducted according to the flocculant dosage ratio under static conditions. The hardness of the wastewater decreases from 143 mg/L to below 100 mg/L, its alkalinity decreases from 118 mg/L to below 100 mg/L, and its SS decreases from 213 mg/L to below 30 mg/L. The cost of coagulant can be controlled to 0.10 yuan/m3, which accords with the recycle standards of water quality and cost requirements of iron and steel works.%设计了一种新型结构的絮凝反应器,利用停留时间分布对其性能进行评价,测得它的分散数为0.10,接近于活塞流反应器,絮凝效率高.采用该絮凝器进行钢铁企业实际废水处理试验,结果表明:按静态条件下的药剂配比进行动态试验,废水的硬度、碱度和SS可以分别由进水时的143、118、213 mg/L控制在100、100、30 mg/L以内,药剂成本可控制在0.10元/m3,达到企业的回用水质和成本要求.

  11. Oscillatory flow chemical reactors

    Directory of Open Access Journals (Sweden)

    Slavnić Danijela S.

    2014-01-01

    Full Text Available Global market competition, increase in energy and other production costs, demands for high quality products and reduction of waste are forcing pharmaceutical, fine chemicals and biochemical industries, to search for radical solutions. One of the most effective ways to improve the overall production (cost reduction and better control of reactions is a transition from batch to continuous processes. However, the reactions of interests for the mentioned industry sectors are often slow, thus continuous tubular reactors would be impractically long for flow regimes which provide sufficient heat and mass transfer and narrow residence time distribution. The oscillatory flow reactors (OFR are newer type of tube reactors which can offer solution by providing continuous operation with approximately plug flow pattern, low shear stress rates and enhanced mass and heat transfer. These benefits are the result of very good mixing in OFR achieved by vortex generation. OFR consists of cylindrical tube containing equally spaced orifice baffles. Fluid oscillations are superimposed on a net (laminar flow. Eddies are generated when oscillating fluid collides with baffles and passes through orifices. Generation and propagation of vortices create uniform mixing in each reactor cavity (between baffles, providing an overall flow pattern which is close to plug flow. Oscillations can be created by direct action of a piston or a diaphragm on fluid (or alternatively on baffles. This article provides an overview of oscillatory flow reactor technology, its operating principles and basic design and scale - up characteristics. Further, the article reviews the key research findings in heat and mass transfer, shear stress, residence time distribution in OFR, presenting their advantages over the conventional reactors. Finally, relevant process intensification examples from pharmaceutical, polymer and biofuels industries are presented.

  12. Experience feedback examination in PWR type reactors operating for the 1997-1999 period; Examen du retour d'experience en exploitation des reacteurs a eau sous pression pour la periode 1997-1999

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2004-07-01

    The present report is relative to the examination that the permanent group has made on the experience feedback in operation for PWR type reactors for the period 1997-1999 that was on eleven themes chosen by the Nuclear Safety and Radiation Protection Authority. It used analysis reports made by I.R.S.N. in support of four meetings of the permanent group devoted to this examination from April 2001 to June 2002. The different themes were operating uncertainties, machining to vibrations, analysis of incidents and gaseous releases, circuits, human factors, behaviour of electric batteries, risk of cold source loss. (N.C.)

  13. Component failure data base of TRIGA reactors

    International Nuclear Information System (INIS)

    This compilation provides failure data such as first criticality, component type description (reactor component, population, cumulative calendar time, cumulative operating time, demands, failure mode, failures, failure rate, failure probability) and specific information on each type of component of TRIGA Mark-II reactors in Austria, Bangladesh, Germany, Finland, Indonesia, Italy, Indonesia, Slovenia and Romania. (nevyjel)

  14. Fast reactors: the future of nuclear energy

    International Nuclear Information System (INIS)

    The main problems to be solved for FBR type reactors become viable economically, presenting the research programs of Europe, United States of America, Japan and Brazil are described. The cooperations between interested countries for improving FBR type reactors, and the financial and human resources necessaries for the development of programs, are evaluated. The fuel cycle is also analysed. (M.C.K.)

  15. Fueling of tandem mirror reactors

    International Nuclear Information System (INIS)

    This paper summarizes the fueling requirements for experimental and demonstration tandem mirror reactors (TMRs), reviews the status of conventional pellet injectors, and identifies some candidate accelerators that may be needed for fueling tandem mirror reactors. Characteristics and limitations of three types of accelerators are described; neutral beam injectors, electromagnetic rail guns, and laser beam drivers. Based on these characteristics and limitations, a computer module was developed for the Tandem Mirror Reactor Systems Code (TMRSC) to select the pellet injector/accelerator combination which most nearly satisfies the fueling requirements for a given machine design

  16. Decree n. 2007-534 of the 4. april 2007 allowing the creation of the base nuclear installation named Flamanville 3, including a EPR type reactor, on the site of Flamanville (Manche); Decret no 2007-534 du 10 avril 2007 autorisant la creation de l'installation nucleaire de base denommee Flamanville 3, comportant un reacteur nucleaire de type EPR, sur le site de Flamanville (Manche)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2007-04-15

    This decree gives the authorization to EDF to create on the site of Flamanville a nuclear installation including a PWR type reactor for a power of 4500 MW and devoted to the electric production. This reactor will can use uranium oxide or a mixture of uranium oxide and plutonium oxide. Considerations concerning the safety are given, as well as the control of the impact of this exploitation on the populations and the environment. (N.C.)

  17. Application of the MOVE algorithm for the identification of reduced order models of a core of a BWR type reactor; Aplicacion del algoritmo MOVE para la identificacion de modelos de orden reducido del nucleo de un reactor tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Victoria R, M.A.; Morales S, J.B. [UNAM, DEPFI, Campus Morelos, en IMTA Jiutepec, Morelos (Mexico)]. e-mail: angelvr@gmail.com

    2005-07-01

    Presently work is applied the modified algorithm of the ellipsoid of optimal volume (MOVE) to a reduced order model of 5 differential equations of the core of a boiling water reactor (BWR) with the purpose of estimating the parameters that model the dynamics. The viability is analyzed of carrying out an analysis that calculates the global dynamic parameters that determine the stability of the system and the uncertainty of the estimate. The modified algorithm of the ellipsoid of optimal volume (MOVE), is a method applied to the parametric identification of systems, in particular to the estimate of groups of parameters (PSE for their initials in English). It is looked for to obtain the ellipsoid of smaller volume that guarantees to contain the real value of the parameters of the model. The PSE MOVE is a recursive identification method that can manage the sign of noise and to ponder it, the ellipsoid represents an advantage due to its easy mathematical handling in the computer, the results that surrender are very useful for the design of Robust Control since to smaller volume of the ellipsoid, better is in general the performance of the system to control. The comparison with other methods presented in the literature to estimate the reason of decline (DR) of a BWR is presented. (Author)

  18. Material test reactor fuel research at the BR2 reactor

    International Nuclear Information System (INIS)

    The construction of new, high performance material test reactor or the conversion of such reactors' core from high enriched uranium (HEU) to low enriched uranium (LEU) based fuel requires several fuel qualification steps. For the conversion of high performance reactors, high density dispersion or monolithic fuel types are being developed. The Uranium-Molybdenum fuel system has been selected as reference system for the qualification of LEU fuels. For reactors with lower performance characteristics, or as medium enriched fuel for high performance reactors, uranium silicide dispersion fuel is applied. However, on the longer term, the U-Mo based fuel types may offer a more efficient fuel alternative and-or an easier back-end solution with respect to the silicide based fuels. At the BR2 reactor of the Belgian nuclear research center, SCK-CEN in Mol, several types of fuel testing opportunities are present to contribute to such qualification process. A generic validation test for a selected fuel system is the irradiation of flat plates with representative dimensions for a fuel element. By flexible positioning and core loading, bounding irradiation conditions for fuel elements can be performed in a standard device in the BR2. For fuel element designs with curved plates, the element fabrication method compatibility of the fuel type can be addressed by incorporating a set of prototype fuel plates in a mixed driver fuel element of the BR2 reactor. These generic types of tests are performed directly in the primary coolant flow conditions of the BR2 reactor. The experiment control and interpretation is supported by detailed neutronic and thermal-hydraulic modeling of the experiments. Finally, the BR2 reactor offers the flexibility for irradiation of full size prototype fuel elements, as 200mm diameter irradiation channels are available. These channels allow the accommodation of various types of prototype fuel elements, eventually using a dedicated cooling loop to provide the

  19. STAR-H2: a battery-type lead-cooled fast reactor for hydrogen manufacture in a sustainable hierarchical hub-spoke energy infrastructure

    International Nuclear Information System (INIS)

    The Secure Transportable Autonomous Reactor for Hydrogen production STAR-H2 is designed to fit into a sustainable global, mid-21st century hierarchical hub-spoke nuclear energy supply architecture based on nuclear fuel, hydrogen, and electricity energy carriers and having favorable energy security, ecological and nonproliferation features. It will produce hydrogen, oxygen and potable water to service cities and their surrounding regions under an assumed electrical generation network based on fuel cells and microturbines and an assumed transportation sector using hydrogen fueled vehicles. STAR-H2 is a long refueling interval (Battery) turnkey heat supply reactor intended for production of hydrogen by thermochemical water cracking. The reactor is a Pb-cooled, mixed U-TRU-Nitride-fueled, fast spectrum reactor delivering 400 MWth of heat at 800degC core outlet temperature. The primary coolant circulates by natural circulation; the 400 MWth heat rating is set by dual requirements for natural circulation; the 400 MWth heat rating is set by dual requirements for natural circulation and for rail shippability of the vessel. An intermediate low pressure He loop carries the heat to a Ca-Br thermochemical water cracking cycle for the manufacture of H2 (and O2). The water cracking cycle rejects heat at 550degC and that heat is used in a supercritical CO2 Brayton cycle turbogenerator to provide hotel load electricity. A thermal desalinisation plant receives discharge heat at 125degC from the Brayton cycle and the brine provides for ultimate heat rejection from the cascaded thermodynamic cycles. The modified UT-3 cycle used in STAR-H2, called the Ca-Br cycle, operates at atmospheric pressure and 750-725degC, uses solid/gas separation steps and achieves about 44% efficiency. Unlike UT-3, it employs a single-stage HBr-dissociation step based on a plasma chemistry technique operating near ambient conditions. The STAR-H2 power plant will operate on a 20 year refueling interval

  20. Procedure of recovery of pin-by-pin fields of energy release in the core of VVER-type reactor for the BIPR-8 code

    Energy Technology Data Exchange (ETDEWEB)

    Gordienko, P. V., E-mail: gorpavel@vver.kiae.ru; Kotsarev, A. V.; Lizorkin, M. P. [National Research Center Kurchatov Institute (Russian Federation)

    2014-12-15

    The procedure of recovery of pin-by-pin energy-release fields for the BIPR-8 code and the algorithm of the BIPR-8 code which is used in nodal computation of the reactor core and on which the recovery of pin-by-pin fields of energy release is based are briefly described. The description and results of the verification using the module of recovery of pin-by-pin energy-release fields and the TVS-M program are given.

  1. Procedure of recovery of pin-by-pin fields of energy release in the core of VVER-type reactor for the BIPR-8 code

    Science.gov (United States)

    Gordienko, P. V.; Kotsarev, A. V.; Lizorkin, M. P.

    2014-12-01

    The procedure of recovery of pin-by-pin energy-release fields for the BIPR-8 code and the algorithm of the BIPR-8 code which is used in nodal computation of the reactor core and on which the recovery of pin-by-pin fields of energy release is based are briefly described. The description and results of the verification using the module of recovery of pin-by-pin energy-release fields and the TVS-M program are given.

  2. PROJECT-SPECIFIC TYPE A VERIFICATION FOR THE BROOKHAVEN GRAPHITE RESEARCH REACTOR ENGINEERED CAP, BROOKHAVEN NATIONAL LABORATORY UPTON, NEW YORK DCN 5098-SR-07-0

    Energy Technology Data Exchange (ETDEWEB)

    Evan Harpenau

    2011-07-15

    The Oak Ridge Institute for Science and Education (ORISE) has reviewed the project documentation and data for the Brookhaven Graphite Research Reactor (BGRR) Engineered Cap at Brookhaven National Laboratory (BNL) in Upton, New York. The Brookhaven Science Associates (BSA) have completed removal of affected soils and performed as-left surveys by BSA associated with the BGRR Engineered Cap. Sample results have been submitted, as required, to demonstrate that remediation efforts comply with the cleanup goal of {approx}15 mrem/yr above background to a resident in 50 years (BNL 2011a).

  3. Thermal-hydraulic interfacing code modules for CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Liu, W.S.; Gold, M.; Sills, H. [Ontario Hydro Nuclear, Toronto (Canada)] [and others

    1997-07-01

    The approach for CANDU reactor safety analysis in Ontario Hydro Nuclear (OHN) and Atomic Energy of Canada Limited (AECL) is presented. Reflecting the unique characteristics of CANDU reactors, the procedure of coupling the thermal-hydraulics, reactor physics and fuel channel/element codes in the safety analysis is described. The experience generated in the Canadian nuclear industry may be useful to other types of reactors in the areas of reactor safety analysis.

  4. Optimally moderated nuclear fission reactor and fuel source therefor

    Science.gov (United States)

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

    2008-07-22

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

  5. 反应器类型对生物厌氧发酵产氢的影响研究进展%Progress in Studying Effects of Bio-reactor Type on Anaerobic Fermentative Bio-hydgrogen Production

    Institute of Scientific and Technical Information of China (English)

    王磊; 谢丽; 罗刚; 周琪

    2012-01-01

    Recently, fermentative hydrogen production with organic wastewater or waste becomes a research focus in the field of bio-energy production, for it is a good way of garbage disposal and energy production. The type of bio-reactor is considered to have important effects on both microbial community structures and mass transfer efficiency between biomass and substrates, which could influence the hydrogen production and stability of the process. In this paper, immobilized-and suspended-cell Hj-producing systems are compared based on biomass growth in forms of granular, bio-film, gel-entrapped bio-particle. Reactor configurations such as continuous stirred tank reactor, anaerobic sequencing batch reactor, up-flow anaerobic sludge blanket, expanded granular sludge bed, anaerobic packed bed, anaerobic fluidized-bed, hybrid bio-reactor and their impacts on bio-hydrogen production are summarized. Finally, suggestions and perspectives are proposed to provide guidance for further research.%利用有机废水或废弃物厌氧发酵制取氢气可以同时达到除废和产能的双重目的,是近年来生物质能的研究热点.产氢反应器的类型会影响反应器内产氢菌的数量和种类以及微生物与底物之间的传质作用,进而影响产氢系统的稳定性和产氢效果.对厌氧产氢悬浮细胞系统和固定化细胞系统(颗粒污泥、生物膜、微生物包埋体)保留生物量的效果进行了对比,总结了连续搅拌釜式反应器( CSTR)、厌氧序批反应器(ASBR)、上流式厌氧污泥床(UASB)、膨胀颗粒污泥床(EGSB)、厌氧填充床、厌氧流化床(AFBR)以及复合生物反应器等不同反应器类型对厌氧产氢的影响,并提出了一些建议与展望,以期为以后的研究提供指导作用.

  6. Development of an innovative plate-type SG for fast breeder reactor. Proposal of the concept and the evaluation of the fabricating method by the test fabrication of the partial model

    International Nuclear Information System (INIS)

    The concept of an innovative plate type SG for the fast reactor fabricated by using the HIP (Hot Isostatic Pressing) method was proposed. The heat transfer plate, which is assembled with rectangular tubes and is fabricated by HIP method, is surrounded by leakage detection spaces. It is possible to apply it to both the pool-type and the loop-type LMFR. In this report, the fabrication technique was studied about the concept for the loop-type LMFR, and the following results were obtained. Hip tests, tensile tests, and structure observation were performed to clarify the suitable HIP condition for the modified 9Cr-1Mo steel. As a result, the optimum condition of 1150 deg C x 1200 kgf/cm2 x 3 hr was found. Nickel-type solder (BNi-5) and gold-type solder (BAu-4) were selected as a joining material to laminate the heat transfer tube plates. Through the comparison of tensile tests, BAu-4 that showed a more excellent joining performance was selected on the assumption of the margin of 5 mm from the welding line. After buckling load had been clarified, the BAu-4 brazing of the heat transfer tube plates was performed using a hot pressing method. Problems were not observed in the welding of simulated header, and in the fabricating of the partial model of SG. (author)

  7. Reactor container

    International Nuclear Information System (INIS)

    A reactor container has a suppression chamber partitioned by concrete side walls, a reactor pedestal and a diaphragm floor. A plurality of partitioning walls are disposed in circumferential direction each at an interval inside the suppression chamber, so that independent chambers in a state being divided into plurality are formed inside the suppression chamber. The partition walls are formed from the bottom portion of the suppression chamber up to the diaphragm floor to isolate pool water in a divided state. Operation platforms are formed above the suppression chamber and connected to an access port. Upon conducting maintenance, inspection or repairing, a pump is disposed in the independent chamber to transfer pool water therein to one or a plurality of other independent chambers to make it vacant. (I.N.)

  8. Analysis of the radiometric survey during the Argonauta reactor operation; Analise do levantamento radiometrico durante operacao do reator Argonauta

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, Eara de S.L.; Cardozo, Katia K.M.; Silva, Joao Carlos P.; Santos, Joao Regis dos, E-mail: esluz@ien.gov.br, E-mail: cardozo@ien.gov.br, E-mail: jcarlos@ien.gov.br, E-mail: regis@ien.gov.br [Instituto de Engenharia Nuclear (CNEN-IEN/RJ), Rio de Janeiro - RJ (Brazil)

    2013-07-01

    The Argonaut reactor at the Institute of Nuclear Engineering-IEN/CNEN, operates normally, the powers between 1.7 and 340 W on neutrongraphy procedures, production of radionuclides and experimental reactor physics lessons to postgraduate courses. The doses from neutrons and gamma radiation are measured when the reactor is critical, inside the reactor hall and surrounding regions. A study of the data obtained was performed to evaluate the daily need of this survey in the reactor hall. Taking into account the principle ALARA, which aims to optimize and minimize the dose received by the individual, we propose, in this work, through an analysis of the acquired data in occupational radiometric surveys, a reformulation of the area monitoring routine practiced by the team of radiological protection of the Institute of Nuclear Engineering - IEN/CNEN-RJ, whereas other monitoring routines regarding the radiological protection are also applied in the routine of the reactor. The operations under review occurred with the reactor operating 340 W power at intervals of 60, 120 and 180 minutes, in monitoring points in controlled areas, supervised and free. The results showed significant dose values in the output of the J-Channel 9 when the operation occurs with this open. With 180 minutes of operation, the measured values of dose rate were lower than the values at 60 min and 120 operations min. At the point in the supervised area, offsite to the reactor hall, situated in the direction of the J-Channel 9, the value reduces more than 14% in any operating time in relation to the dose rate measured at the point opposite the canal. There is a 50% reduction in the dose rates for operations with and J-9 closed. The results suggest a new frequency of radiometric survey whose mode of operation is maintained in similar conditions, since combined with other relevant practices of radiation protection.

  9. NEUTRONIC REACTORS

    Science.gov (United States)

    Anderson, J.B.

    1960-01-01

    A reactor is described which comprises a tank, a plurality of coaxial steel sleeves in the tank, a mass of water in the tank, and wire grids in abutting relationship within a plurality of elongated parallel channels within the steel sleeves, the wire being provided with a plurality of bends in the same plane forming adjacent parallel sections between bends, and the sections of adjacent grids being normally disposed relative to each other.

  10. Some economic aspects of natural uranium graphite gas reactor types. Present status and trends of costs in France; Quelques aspects economiques de la filiere uranium naturel - Graphite - gaz. Etat actuel et tendance des couts en France

    Energy Technology Data Exchange (ETDEWEB)

    Gaussens, J.; Tanguy, P. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Leo, B. [Electricite de France (EDF), 75 - Paris (France)

    1964-07-01

    The first part of this report defines the economic advantages of natural uranium fuels, which are as follows: the restricted number and relatively simple fabrication processes of the fuel elements, the low cost per kWh of the finished product and the reasonable capital investments involved in this type of fuel cycle as compared to that of enriched uranium. All these factors combine to reduce the arbitrary nature of cost estimates, which is particularly marked in the case of enriched uranium due to the complexity of its cycle and the uncertainties of plutonium prices). Finally, the wide availability of yellowcake, as opposed to the present day virtual monopoly of isotope separation, and the low cost of natural uranium stockpiling, offer appreciable guarantees in the way of security of supply and economic and political independence as compared with the use of enriched uranium. As far as overall capital investments are concerned, it is shown that, although graphite-gas reactor costs are higher than those of light water reactors in certain capacity ranges, the situation becomes far less clear when we start taking into account, in the interest of national independence, the cost of nuclear fuel production equipment in the case of each of these types of reactor. Finally, the marginal cost of the power capacity of a graphite-gas reactor is low and its technological limitations have receded (owing particularly to the use of prestressed concrete). It is a well known fact that the trend is now towards larger power station units, which means that the rentability of natural uranium graphite reactors as compared to other types of reactors will become more and more pronounced. The second section aims at presenting a realistic short and medium term view of the fuel, running, and investment costs of French natural uranium graphite gas, reactors. Finally, the economic goals which this type of reactor can reach in the very near future are given. It is thus shown that considerable

  11. Safe Operation of Research Reactors in Germany

    International Nuclear Information System (INIS)

    In Germany, experience was gained in the field of safe operation of research reactors during the last five decades. In this time, in total 46 research reactors were built and operated safely. Concerning the design, there is, or has been, a very broad range of different types of research reactors. The variety of facilities includes large pool or tank reactors with a thermal power of several tens of megawatt as well as small educational reactors with a negligible thermal power and critical assemblies. At present, 8 research reactors are still in operation. The other facilities are permanently shutdown, in decommissioning or have already been dismantled completely and released from regulatory control. In this paper, four selected facilities still being operated are presented as examples for safe operation of research reactors in Germany, including especially a description of the safety reviews and safety upgrades for the older facilities. (author)

  12. U.S. Nuclear Power Reactor Plant Status

    Data.gov (United States)

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

  13. 外置式与内置式秸秆生物反应堆对番茄生长及光合性能的影响%Effects of outer type and built-in type straw bio-reactors on tomato growth and photosynthetic performance

    Institute of Scientific and Technical Information of China (English)

    卞中华; 王玉; 胡晓辉; 邹志荣; 张静; 燕飞

    2013-01-01

    Taking the tomato (Solanum lycopersicum) cultivar "Kuiguanl08" as test object, a comparative study was made on the effects of outer type and built-in type straw bio-reactors on the CO2 concentration, air relative humidity , air vapor pressure deficit in the solar greenhouse during the tomato growth over autumn-delayed cultivation as well as the effects of the bio-reactors on the tomato growth and photosynthetic performance.As compared with that in CK, the average CO2 concentration in the greenhouse with outer type straw bio-reactor at 9:30-11:30 and 14:30-15 :00 on sunny days was increased significantly by 207.3 and 103 μmol·mol-1 , respectively, and the ave-rage CO2 concentration in the greenhouse with built-in straw bio-reactor at 9:30 -11:30 on sunny days was raised by 19.0 μmol·mol-1 .Both the outer type and the built-in type straw bio-reactors promoted the tomato plant height growth and early flowering, enhanced the plant net photosynthetic rate and the yield per plant and per unit area significantly, and decreased the plant transpiration rate at the stages of vegetative growth and fruit- bearing significantly.Nevertheless, as compared with built-in type straw bio-reactor, outer type straw bio-reactor was more suitable for the autumn- delayed cultivation of tomato in solar greenhouse.%以“魁冠108”番茄为试验材料,对比研究了外置式和内置式秸秆生物反应堆在秋延后番茄生产过程中对日光温室内CO2浓度、空气相对湿度、空气饱和水汽压差以及番茄生长和光合性能的影响.结果表明:与对照相比,晴天9:30-11:30和14:30-15:00外置式秸秆生物反应堆温室内CO2浓度平均提高了207.3和103 μmol·mol-1,差异显著;晴天9:30-11:30内置式秸秆生物反应堆温室内CO2浓度平均提高了19.0 μmol·mol-1;外置式和内置式秸秆生物反应堆促进了番茄株高生长,使植株提早开花,显著提高了番茄净光合速率、单株产量及单位面积产量,显著降

  14. Official announcement of the directive on protection of nuclear power plant equipped with LWR-type reactors from human intrusion or other interference by third parties. Announcement of BMU (German Federal Ministry Environment), of 6 Dec. 1995 - RS I 3 13151 - 6/14

    International Nuclear Information System (INIS)

    An operating permit for a nuclear power plant is to be granted only if the applicant and facility operator presents evidence guaranteeing the legally required physical protection and other security measures for protection from human instrusion and other type of interference. As a basis for review and licensing, the competent authorities in 1987 have issued a directive specifying the requirements to be met for physical protection of nuclear power plant equipped with PWR-type reactors, and in 1994 followed a second, analogous directive relating to nuclear power plant with BWR-type reactors. The directive now announced for physical protection of nuclear power plant equipped with LWR-type reactors combines and replaces the two former ones, and from the date of the announcement is the only applicable directive. The text of the directive is not reproduced for reasons of secrecy protection. (orig./CB)

  15. Manual for the operation of research reactors

    International Nuclear Information System (INIS)

    The great majority of the research reactors in newly established centres are light-water cooled and are often also light-water moderated. Consequently, the IAEA has decided to publish in its Technical Reports Series a manual dealing with the technical and practical problems associated with the safe and efficient operation of this type of reactor. Even though this manual is limited to light-water reactors in its direct application and presents the practices and experience at one specific reactor centre, it may also be useful for other reactor types because of the general relevance of the problems discussed and the long experience upon which it is based. It has, naturally, no regulatory character but it is hoped that it will be found helpful by staff occupied in all phases of the practical operation of research reactors, and also by those responsible for planning their experimental use. 23 refs, tabs

  16. Assessment of subcriticality during PWR-type reactor refueling; Evaluation de la sous-criticite lors des operations de chargement d'un reacteur nucleaire REP

    Energy Technology Data Exchange (ETDEWEB)

    Verdier, A

    2005-04-15

    During the core loading period of a PWR, any fuel assembly misplacements may significantly reduce the existing criticality margin. The Dampierre 4-18 event showed the present monitoring based on the variations of the outside-core detector counting rate cannot detect such misplacements. In order to circumvent that, a more detailed analysis of the available signal was done. We particularly focused on the neutronic noise analysis methods such as MSM (modified source multiplication), MSA (amplified source multiplication), Rossi-{alpha} and Feynman-{alpha} methods. The experimental part of our work was dedicated to the application of those methods to a research reactor. Finally, our results showed that those methods cannot be used with the present PWR instrumentation. Various detector positions were then studied using Monte Carlo calculations capable of following the neutron origin. Our results showed that the present technology does not allow us to use any solution based on neutron detection for monitoring core loading. (author)

  17. 螺旋升流塔式光催化反应器的设计研究%Design of Spiral Up-flow Tower-type Photocatalysis Reactor

    Institute of Scientific and Technical Information of China (English)

    徐璇; 吉芳英; 范子红

    2009-01-01

    Based on the cyclone separator model for granular pollutants, a spiral up-flow tower-type photocatalysis reactor was designed to increase the recovery rate of photocatalyst. In order to reach a high A/V ratio (the ratio of illumination area to reaction solution volume), a tower structure was used on this reactor, and the A/V ratio can reach 12.95 in this experimental condition. This reactor was used to treat the nitrobenzene simulated wastewater. When the initial concentration of nitrobenzene is 466 mg/L, the removal rate of nitrobenzene is about 60%, and the photocatalyst recovery rate reaches 92.80% after 12h.%采用颗粒污染物的旋流分离模型设计了螺旋升流塔式光催化反应器,能够在悬浮态光催化反应系统中提高光催化剂的回收率.反应器采用塔式结构布置,能有效提高反应器的光照面积与反应液体积之比(A/V值),在本试验条件下A/V值可达到12.95.采用该反应器处理硝基苯模拟废水,当硝基苯初始浓度为466 mg/L时,反应器对硝基苯的去除率稳定在60%左右,运行12 h后对光催化剂的回收率为92.80%.

  18. Study on an Automatic Knife Switch of New Type High-Voltage Small Current Reactor%新型高电压小电流电抗器自动刀闸的研究

    Institute of Scientific and Technical Information of China (English)

    许杰; 李世武; 孙伟

    2014-01-01

    针对当前电抗器投切刀闸存在的投切行程小、适用范围受限制、远程操控不方便等问题,研发了一种新型的高电压小电流电抗器自动刀闸,给出了自动刀闸结构图,分析了其工作原理。对该自动刀闸进行峰值耐受电流试验和机械操作试验,测试结果符合要求,有效通断率100%,提高了电抗器串并联自动化程度,易于远程控制且性能稳定,可广泛用于电力系统机构的各项试验。%Aiming at several problems such that the existing reactor knife switch is small in travel distance for knife making and breaking, limited in applicable range, not convenient in remote operating etc., this paper developed an automatic knife switch of new type high-voltage small current reactor and gave the structural diagram of the automatic knife switch, analyzing its working principle. The peak value withstand current and mechanical operating tests were carried out for the automatic knife switch. The test results are in conformity to the requirements, with 100% valid making-breaking rate, which raises the reactor serial and parallel connection automation degree, easy for remote control and stable in performance. The switch could be widely used in each test of electric power system organization.

  19. Nuclear reactor

    International Nuclear Information System (INIS)

    A nuclear reactor is described in which the core components, including fuel-rod assemblies, control-rod assemblies, fertile rod-assemblies, and removable shielding assemblies, are supported by a plurality of separate inlet modular units. These units are referred to as inlet module units to distinguish them from the modules of the upper internals of the reactor. The modular units are supported, each removable independently of the others, in liners in the supporting structure for the lower internals of the reactor. The core assemblies are removably supported in integral receptacles or sockets of the modular units. The liners, units, sockets and assemblies have inlet openings for entry of the fluid. The modular units are each removably mounted in the liners with fluid seals interposed between the opening in the liner and inlet module into which the fluid enters in the upper and lower portion of the liner. Each assembly is similarly mounted in a corresponding receptacle with fluid seals interposed between the openings where the fluid enters in the lower portion of the receptacle or fitting closely in these regions. As fluid flows along each core assembly a pressure drop is produced along the fluid so that the fluid which emerges from each core assembly is at a lower pressure than the fluid which enters the core assembly. However because of the seals interposed in the mountings of the units and assemblies the pressures above and below the units and assemblies are balanced and the units are held in the liners and the assemblies are held in the receptacles by their weights as they have a higher specific gravity than the fluid. The low-pressure spaces between each module and its liner and between each core assembly and its module is vented to the low-pressure regions of the vessel to assure that fluid which leaks through the seals does not accumulate and destroy the hydraulic balance

  20. Nuclear reactor physics course for reactor operators

    International Nuclear Information System (INIS)

    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