WorldWideScience

Sample records for agr type reactors

  1. 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

  2. CO2 direct cycles suitable for AGR type reactors

    International Nuclear Information System (INIS)

    The perspectives given by the gas turbines under pressure, to build simple nuclear power plants and acieving significantly high yield, are specified. The CO2 is characterised by by good efficiency under moderate temperature (500 to 750 Celsius degrees), compactness and the simpleness of machines and the safe exploitation (supply, storage, relief cooling, thermosyphon). The revision of thermal properties of the CO2 and loss elements show that several direct cycles would fit in particular to the AGR type reactors. Cycles that would diverge a little from classical models and able to lead to power and heat generation can lead by simple means to the best results. Several satisfying solutions present for the starting up, the power regulation and the stopping. The nuclear power plant components and the functioning safety are equally considered in the present report. The conclusions stimulate the studies and realizations of carbon dioxide gas turbines in when approprite

  3. Advanced gas-cooled reactors (AGR)

    International Nuclear Information System (INIS)

    The paper describes the advanced gas-cooled reactor system, Hunterston ''B'' power station, which is a development of the earlier natural uranium Magnox type reactor. Data of construction, capital cost, operating performance, reactor safety and also the list of future developments are given

  4. Rapid Staphylococcus aureus agr Type Determination by a Novel Multiplex Real-Time Quantitative PCR Assay

    OpenAIRE

    Francois, Patrice; Koessler, Thibaud; Huyghe, Antoine; Harbarth, Stephan; Bento, Manuela; Lew, Daniel; Etienne, Jérôme; Pittet, Didier; Schrenzel, Jacques

    2006-01-01

    The accessory gene regulator (agr) is a crucial regulatory component of Staphylococcus aureus involved in the control of bacterial virulence factor expression. We developed a real-time multiplex quantitative PCR assay for the rapid determination of S. aureus agr type. This assay represents a rapid and affordable alternative to sequence-based strategies for assessing relevant epidemiological information.

  5. Rapid Staphylococcus aureus agr type determination by a novel multiplex real-time quantitative PCR assay.

    Science.gov (United States)

    Francois, Patrice; Koessler, Thibaud; Huyghe, Antoine; Harbarth, Stephan; Bento, Manuela; Lew, Daniel; Etienne, Jérôme; Pittet, Didier; Schrenzel, Jacques

    2006-05-01

    The accessory gene regulator (agr) is a crucial regulatory component of Staphylococcus aureus involved in the control of bacterial virulence factor expression. We developed a real-time multiplex quantitative PCR assay for the rapid determination of S. aureus agr type. This assay represents a rapid and affordable alternative to sequence-based strategies for assessing relevant epidemiological information. PMID:16672433

  6. Rapid Staphylococcus aureus agr Type Determination by a Novel Multiplex Real-Time Quantitative PCR Assay

    Science.gov (United States)

    Francois, Patrice; Koessler, Thibaud; Huyghe, Antoine; Harbarth, Stephan; Bento, Manuela; Lew, Daniel; Etienne, Jérôme; Pittet, Didier; Schrenzel, Jacques

    2006-01-01

    The accessory gene regulator (agr) is a crucial regulatory component of Staphylococcus aureus involved in the control of bacterial virulence factor expression. We developed a real-time multiplex quantitative PCR assay for the rapid determination of S. aureus agr type. This assay represents a rapid and affordable alternative to sequence-based strategies for assessing relevant epidemiological information. PMID:16672433

  7. Status of the NGNP Fuel Experiment AGR-2 Irradiated in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Blaine Grover

    2012-10-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2), which utilized the same experiment design as well as control and monitoring systems as AGR-1, started irradiation in June 2010 and is currently scheduled to be completed in April 2013. The design of this experiment and support systems will be briefly discussed, followed by the progress and status of the experiment to date.

  8. Types of Nuclear Reactors

    International Nuclear Information System (INIS)

    The presentation is based on the following areas: Types of Nuclear Reactors, coolant, moderator, neutron spectrum, fuel type, pressurized water reactor (PWR), boiling water reactor (BWR) reactor pressurized heavy water (PHWR), gas-cooled reactor, RBMK , Nuclear Electricity Generation,Challenges in Nuclear Technology Deployment,EPR, APR1400, A P 1000, A PWR, ATMEA 1, VVER-1000, A PWR, VVER 1200, Boiling Water Reactor, A BWR, A BWR -II, ESBUR, Ke ren, AREVA, Heavy Water Reactor, Candu 6, Acr-1000, HWR, Bw, Iris, CAREM NuCcale, Smart, KLT-HOS, Westinghouse small modular Reactor, Gas Cooled Reactors, PBMR.

  9. Irradiation performance of AGR-1 high temperature reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Paul A. Demkowicz; John D. Hunn; Robert N. Morris; Charles A. Baldwin; Philip L. Winston; Jason M. Harp; Scott A. Ploger; Tyler Gerczak; Isabella J. van Rooyen; Fred C. Montgomery; Chinthaka M. Silva

    2014-10-01

    The AGR-1 experiment contained 72 low-enriched uranium oxide/uranium carbide TRISO-coated particle fuel compacts in six capsules irradiated to burnups of 11.2 to 19.5% FIMA, with zero TRISO coating failures detected during the irradiation. The irradiation performance of the fuel–including the extent of fission product release and the evolution of kernel and coating microstructures–was evaluated based on detailed examination of the irradiation capsules, the fuel compacts, and individual particles. Fractional release of 110mAg from the fuel compacts was often significant, with capsule-average values ranging from 0.01 to 0.38. Analysis of silver release from individual compacts indicated that it was primarily dependent on fuel temperature history. Europium and strontium were released in small amounts through intact coatings, but were found to be significantly retained in the outer pyrocrabon and compact matrix. The capsule-average fractional release from the compacts was 1×10 4 to 5×10 4 for 154Eu and 8×10 7 to 3×10 5 for 90Sr. The average 134Cs release from compacts was <3×10 6 when all particles maintained intact SiC. An estimated four particles out of 2.98×105 experienced partial cesium release due to SiC failure during the irradiation, driving 134Cs release in two capsules to approximately 10 5. Identification and characterization of these particles has provided unprecedented insight into the nature and causes of SiC coating failure in high-quality TRISO fuel. In general, changes in coating morphology were found to be dominated by the behavior of the buffer and inner pyrolytic carbon (IPyC), and infrequently observed SiC layer damage was usually related to cracks in the IPyC. Palladium attack of the SiC layer was relatively minor, except for the particles that released cesium during irradiation, where SiC corrosion was found adjacent to IPyC cracks. Palladium, silver, and uranium were found in the SiC layer of irradiated particles, and characterization

  10. The long term storage of advanced gas-cooled reactor (AGR) fuel

    International Nuclear Information System (INIS)

    The approach being taken by BNFL in managing the AGR lifetime spent fuel arisings from British Energy reactors is given. Interim storage for up to 80 years is envisaged for fuel delivered beyond the life of the Thorp reprocessing plant. Adopting a policy of using existing facilities, to comply with the principles of waste minimisation, has defined the development requirements to demonstrate that this approach can be undertaken safely and business issues can be addressed. The major safety issues are the long term integrity of both the fuel being stored and structure it is being stored in. Business related issues reflect long term interactions with the rest of the Sellafield site and storage optimisation. Examples of the development programme in each of these areas is given. (author)

  11. Natural convection type reactor

    International Nuclear Information System (INIS)

    In a natural convection type nuclear reactor, a reactor core is disposed such that the top of the reactor core is always situated in a flooded position even if pipelines connected to the pressure vessel are ruptured and the level at the inside of the reactor vessel is reduced due to flashing. Further, a lower dry well situated below the pressure vessel is disposed such that it is in communication with a through hole to a pressure suppression chamber situated therearound and the reactor core is situated at the level lower than that of the through hole. If pipelines connected to the pressure vessel are ruptured to cause loss of water, although the water level is lowered after the end of the flashing, the reactor core is always flooded till the operation of a pressure accummulation water injection system to prevent the top of the reactor core even from temporary exposure. Further, injected water is discharged to the outside of the pressure vessel, transferred to the lower dry well, and flows through the through hole to the pressure control chamber and cools the surface of the reactor pressure vessel from the outside. Accordingly, the reactor core is cooled to surely and efficiently remove the after-heat. (N.H.)

  12. FBR type reactor

    International Nuclear Information System (INIS)

    The present invention provides an FBR type reactor in which the combustion of reactor core fuels is controlled by reflectors, and the position of a reflector driving device can be controlled even during shut down of the reactor. Namely, the reflector driving device is attracted to the outer wall surface of a reactor core barrel by electromagnetic attraction force. An inertia body is disposed vertically movably to the upper portion of the reflector driving device. Magnetic repulsive coils generate instantaneous magnetic repulsive force between the inertia body and the reflector driving device. With such a constitution, the reflector driving device can be driven by using magnetic repulsion of the electromagnetic repulsive coils and inertia of the inertia body. As a result, not only the reflectors can be elevated at an ultraslow speed during normal reactor operation, but also fine position adjustment for the reflector driving device, as well as fine position adjustment of the reflectors required upon restart of the reactor can be conducted by lowering the reflector driving device during shut down of the reactor. (I.S.)

  13. 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.)

  14. 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.)

  15. Design and Status of the NGNP Fuel Experiment AGR-3/4 Irradiated in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Blaine Grover

    2012-10-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and is currently scheduled to be completed in April 2013. The third and fourth experiments have been combined into a single experiment designated AGR-3/4, which started its irradiation in December 2011 and is currently scheduled to be completed in November 2013. Since the purpose of this experiment is to provide data on fission product migration and retention in the NGNP reactor, the design of this experiment is

  16. 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.)

  17. 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.)

  18. Light water type reactor

    International Nuclear Information System (INIS)

    The nuclear reactor of the present invention prevents disruption of a reactor core even in a case of occurrence of entire AC power loss event, and even if a reactor core disruption should occur, it prevents a rupture of the reactor container due to excess heating. That is, a high pressure water injection system and a low pressure water injection system operated by a diesel engine are disposed in the reactor building in addition to an emergency core cooling system. With such a constitution, even if an entire AC power loss event should occur, water can surely be injected to the reactor thereby enabling to prevent the rupture of the reactor core. Even if it should be ruptured, water can be sprayed to the reactor container by the low pressure water injection system. Further, if each of water injection pumps of the high pressure water injection system and the low pressure water injection system can be driven also by motors in addition to the diesel engine, the pump operation can be conducted more certainly and integrally. (I.S.)

  19. FBR type reactor

    International Nuclear Information System (INIS)

    A circular neutron reflector is disposed vertically movably so as to surround the outer circumference of a reactor core barrel. A reflector driving device comprises a driving device main body attracted to the outer wall surface of the reactor barrel by electromagnetic attraction force and an inertia body disposed above the driving device main body vertically movably. A reflector is connected below the reactor driving device. At the initial stage, a spontaneous large current is supplied to upper electromagnetic repulsion coils of the reflector driving device, impact electromagnetic repulsion force is caused between the inertia body and the reflector driving device, so that the driving device main body moves downwardly by a predetermined distance and stopped. The reflector driving device can be lowered in a step-like manner to an appropriate position suitable to restart the reactor during stoppage of the reactor core by conducting spontaneous supply of current repeatedly to the upper electromagnetic repulsion coils. (I.N.)

  20. 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.)

  1. 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).

  2. HTGR type reactor

    International Nuclear Information System (INIS)

    A reactor core is disposed at the center of a reactor container, a reflector is disposed on the outer side thereof, a steam generator is disposed further outer side thereof coaxially, and they are constituted as an integrated one container. A gas circulator and control rod drives are protruded at the outer side of the lower portion of the integrated container. Heat insulators are disposed on the inner side of the container wall in the upper portion of the reactor container. Helium gas risen in the reactor core and heated to a high temperature descends in a circular steam generator and undergoes heat exchange with water, and is then pressurized in the gas circulator after the lowering of the temperature, and returned to the inlet of the reactor core from the lower central portion of the container. With such procedures, the helium gas as primary coolants circulates only in the container to improve confinement. The device can be reduced in the size and the cost. (I.N.)

  3. BWR type reactor

    International Nuclear Information System (INIS)

    An austenite/ferrite stainless steel is used for upstream of a condensate cleanup system and only austenitic stainless steel is used for the downstream. An iron concentration in feedwater is kept lower than 0.1ppb and a volume of a reactor cleanup system is increased, to remove Co before it is deposited on the surface of fuels. With such procedures, any of an ion 60Co concentration and a crud 60Co concentration in coolants can be kept low, thereby enabling to suppress radiation dose rate on the surface of equipments and pipelines. (T.M.)

  4. Reactor fueling of BWR type reactors

    International Nuclear Information System (INIS)

    Purpose: To enable the pattern exchange for control rods during burning in Control Cell Core type BWR reactors. Constitution: A plurality of control cells are divided into a plurality of groups such that the control cells is aparted from each other by way of at least two fuel assemblies other than the control cells with respect to the vertical and lateral directions of the reactor core cross section, as well as they are in adjacent with control cells of other groups with respect to the orthogonal direction. This enables to perform the pattern exchange for the control rods during burning in the control cell core with ease, and the control blade and the story effect harmful to the mechanical soundness of fuels can thus be suppressed. (Moriyama, K.)

  5. Molten salt reactor type

    International Nuclear Information System (INIS)

    This document is one of the three parts of a first volume devoted to the compilations of American data on the molten salt reactor concept. Emphasize is put essentially on the fuel salt of the primary circuit inside which fission reactions occur. The reasons why the (LiF-BeF2-ThF4-UF4) salt was chosen for the M.S.B.R. concept are examined; the physical, physicochemical and chemical properties of this salt are discussed with its interactions with the structural materials and its evolution in time. An important part of this volume is devoted to the continuous reprocessing of the active salt, the project designers having deemed advisable to take advantage at best from the availability of a continuous purification, in a thermal breeding. The problem of tritium formation and distribution inside the reactor is also envisaged and the fundamentals of the chemistry of the secondary coolant salt are given. The solutions proposed are: the hydrogen scavenging of the primary circuit, a reduction in metal permeability by an oxyde layer deposition on the side in contact with the vapor, and tritium absorption through an isotope exchange with the hydroxifluoroborate

  6. Liquid lithium control type LMFBR type reactor

    International Nuclear Information System (INIS)

    In a liquid lithium control type LMFBR type reactor, a fuel exchange device passing through the center of a stationary lid and capable of reaching a predetermined position of the reactor core is disposed. A control mechanism having a case in parallel with a reactor core shaft and a shrinkable sealed cylinder in the case is disposed in the outer circumferential region of the reactor core, and a tank for liquid lithium is connected to the sealed cylinder, and the pressure in the case is controlled by supplying or discharging coolants. Coolants in the reactor container are sucked and injected into the case. The sealed cylinder is shortened axially to attain balance of the pressure between the inner side and the outer side of the cylinder, and a portion of the liquid lithium is pulled out and recycled to a tank. Neutron absorbers rise by so much, to attain the same condition as in the case that control rods are drawn out. The pressure in the case can be optionally determined by a control device, and axial dimension of the sealed cylinder can be determined optionally. Then, a rotational plug for loading a fuel exchange device and control rod drives are not necessary to extremely simplify the structure of reactor upper structures. (N.H.)

  7. PWR type reactor

    International Nuclear Information System (INIS)

    Coolant discharging windows disposed to a control rod cluster guide tube are distributed in a region between the height of the lower end of a coolant exit nozzle and the height of the lower nozzle of an upper reactor core support column. The flow of coolants in the lateral direction toward an exit nozzle does not flow backwardly from the discharging windows to the inside of the control rod cluster guide tube, and the flow of coolants in the control rod cluster guide tube is discharged from each of the coolant discharging windows to the outside directly and rapidly while forming branched streams. As a result, the flow rate of coolants passing through a continuous portion is greatly reduced, and the flow rate of coolants in the direction traversing the control rods is greatly reduced. Accordingly, fluid vibrations for all the control rod clusters is reduced to reduce abrasion and the thickness reduction of the walls of a guide plate of the control rod cluster guide tube caused by contact with the control rods. (N.H.)

  8. BWR type reactor

    International Nuclear Information System (INIS)

    No channel box is mounted to a fuel assembly, but a partition plate for separating coolant flow channels between each of fuel bundles is disposed between each of fuel bundles along the direction of height for the reactor core instead of the channel box. The partition plate has a shape surrounding the fuel bundles only in a specific region, or so that coolant flow channels for a plurality of fuel bundles of identical output are integrated. As a result, cross-flow of coolants can be prevent without channel box and, further, radial expansion of the channel box can be eliminated. As the same time, the bending for the entire assembly due to the irradiation growth of the channel box is also eliminated and structural stability can be attained without using upper grid plates. Further, it is possible to minimize the pressure loss caused between the upper and lower portions of the assembly and it is possible to adjsut with respective thermohydrodynamic properties of the high conversion region and the burner region. (K.M.)

  9. HTGR type reactor

    International Nuclear Information System (INIS)

    A heat insulated high temperature gas rising pipe is disposed at the center of a steam generator and a helical heat exchanger is disposed at the periphery thereof. Helium coolants heated to a high temperature from the reactor core rises through the insulated high temperature gas rising pipe and then turns downward in the outer region of the helical heat exchange pipe, and a gas recycling device is disposed for discharging cooled gases to an annular portion below. On the other hand, feedwater from a liquid inlet nozzle is heated by the high temperature helium coolants during rising in the helical heat exchange pipe, to be a two-phase superheated flow. Accordingly, thermohydrodynamic instability due to downhill boiling is eliminated. Since a pipeline from a water reservoir is connected to the liquid inlet nozzle of the steam generator, the coolants sent from the water reservoir flow in the helical heat exchange pipe for a long period of time upon occurrence of accident such as troubles in an after-heat removal system, to cool the helium coolants at the outside of the pipe by utilizing heat dissipation due to the latent heat of coolants evaporation. (N.H.)

  10. Role of the Agr-like quorum-sensing system in regulating toxin production by Clostridium perfringens type B strains CN1793 and CN1795.

    Science.gov (United States)

    Chen, Jianming; McClane, Bruce A

    2012-09-01

    Clostridium perfringens type B causes enteritis and enterotoxemia in domestic animals. By definition, these bacteria must produce alpha toxin (CPA), beta toxin (CPB) and epsilon toxin (ETX) although most type B strains also produce perfringolysin O (PFO) and beta2 toxin (CPB2). A recently identified Agr-like quorum-sensing (QS) system in C. perfringens controls all toxin production by surveyed type A, C, and D strains, but whether this QS is involved in regulating toxin production by type B strains has not been explored. Therefore, the current study introduced agrB null mutations into type B strains CN1795 and CN1793. Both type B agrB null mutants exhibited reduced levels of CPB, PFO, and CPA in their culture supernatants, and this effect was reversible by complementation. The reduced presence of CPB in culture supernatant involved decreased cpb transcription. In contrast, the agrB null mutants of both type B strains retained wild-type production levels of ETX and CPB2. In a Caco-2 cell model of enteritis, culture supernatants of the type B agrB null mutants were less cytotoxic than supernatants of their wild-type parents. However, in an MDCK cell in vitro model for enterotoxemic effects, supernatants from the agrB null mutants or wild-type parents were equally cytotoxic after trypsin activation. Coupling these and previous results, it is now evident that strain-dependent variations exist in Agr-like QS system regulation of C. perfringens toxin production. The cell culture results further support a role for trypsin in determining which toxins contribute to disease involving type B strains. PMID:22689820

  11. Analysis of Gln223Agr Polymorphism of Leptin Receptor Gene in Type II Diabetic Mellitus Subjects among Malaysians

    Directory of Open Access Journals (Sweden)

    Chong Pei Pei

    2013-09-01

    Full Text Available Leptin is known as the adipose peptide hormone. It plays an important role in the regulation of body fat and inhibits food intake by its action. Moreover, it is believed that leptin level deductions might be the cause of obesity and may play an important role in the development of Type 2 Diabetes Mellitus (T2DM, as well as in cardiovascular diseases (CVD. The Leptin Receptor (LEPR gene and its polymorphisms have not been extensively studied in relation to the T2DM and its complications in various populations. In this study, we have determined the association of Gln223Agr loci of LEPR gene in three ethnic groups of Malaysia, namely: Malays, Chinese and Indians. A total of 284 T2DM subjects and 281 healthy individuals were recruited based on International Diabetes Federation (IDF criteria. Genomic DNA was extracted from the buccal specimens of the subjects. The commercial polymerase chain reaction (PCR method was carried out by proper restriction enzyme MSP I to both amplify and digest the Gln223Agr polymorphism. The p-value among the three studied races was 0.057, 0.011 and 0.095, respectively. The values such as age, WHR, FPG, HbA1C, LDL, HDL, Chol and Family History were significantly different among the subjects with Gln223Agr polymorphism of LEPR (p < 0.05.

  12. AGR-1 Data Qualification Report

    Energy Technology Data Exchange (ETDEWEB)

    Michael Abbott

    2010-03-01

    ABSTRACT Projects for the very high temperature reactor (VHTR) Technology Development Office (TDO) program provide data in support of Nuclear Regulatory Commission licensing of the VHTR. Fuel and materials to be used in the reactor are tested and characterized to quantify performance in high temperature and high fluence environments. The VHTR program has established the NGNP Data Management and Analysis System (NDMAS) to ensure that VHTR data are (1) qualified for use, (2) stored in a readily accessible electronic form, and (3) analyzed to extract useful results. This document focuses on the first NDMAS objective. It describes the data streams associated with the first Advanced Gas Reactor experiment (AGR-1), the processing of these data within NDMAS, and reports the qualification status of the data. Data qualification activities within NDMAS for specific types of data are determined by the data qualification category assigned by the data generator. They include: (1) capture testing, to confirm that the data stored within NDMAS are identical to the raw data supplied, (2) accuracy testing, to confirm that the data are an accurate representation of the system or object being measured, and (3) documentation that the data were collected under an NQA-1 or equivalent quality assurance program. The NDMAS database processing and qualification status of the following five data streams is reported in this document: 1. Fuel fabrication data. All data have been processed into the NDMAS database and qualified (1,819 records). 2. Fuel irradiation data. Data from all 13 AGR-1 reactor cycles have been processed into the NDMAS database and tested. Of these, 85% have been qualified and 15% have failed NDMAS accuracy testing. 3. FPMS data. Reprocessed (January 2010) data from all 13 AGR-1 reactor cycles have been processed into the database and capture tested. Final qualification of these data will be recorded after QA approval of an Engineering Calculations and Analysis Report

  13. Fourth generation type reactors - Synthesis note

    International Nuclear Information System (INIS)

    Six types of reactors have been studied: High or very high temperature helium cooled type reactors, fast neutrons sodium cooled type reactors, fast neutrons gas cooled type reactors, fast neutrons lead or lead-bismuth cooled type reactors, supercritical water type reactors, molten salt type reactors. For the high or very high temperature type reactors the questions of safety and radiation protection have been tackled through the fuel, the neutronics, the materials, the passive systems, safety and reliability of associated industrial processes, risks in relation with graphite, fire and explosion risks linked to hydrogen production; about the fast neutron sodium cooled type reactors the principal questions of safety are tackled through the specific risks linked to the metallic fuel, the neutronic effects in case of loss of coolant said sodium 'vacuum effect', risk of core meltdown, risks linked to sodium, passive systems, ability of structures inspection; concerning the fast neutron gas cooled type reactors, the questions of safety and radiation protection are the aspects linked to the reactor and the aspects linked to the fuel fabrication, this last question has been tackled for each reactor type. A part has been devoted to the production and the management of waste in the case of deployment of a fourth generation reactors park. (N.C.)

  14. Natural convection type BWR reactor

    International Nuclear Information System (INIS)

    In a natural convection type BWR reactor, a mixed stream of steams and water undergo a great flow resistance. In particular, pressure loss upon passing from an upper plenum to a stand pipe and pressure loss upon passing through rotational blades are great. Then, a steam dryer comprising laminated dome-like perforated plates and a drain pipe for flowing down separated water to a downcomer are disposed above a riser. The coolants heated in the reactor core are boiled, uprise in the riser as a gas-liquid two phase flow containing voids, release steams containing droplets from the surface of the gas-liquid two phase, flow into the steam dryer comprising the perforated plates and are separated into a gas and a liquid. The dried steams flow to a turbine passing through a main steam pipe and the condensated droplets flow down through the drain pipe and the downcomer to the lower portion of the reactor core. In this way, the conventional gas-liquid separator can be saved without lowering the quality of steam drying to reduce the pressure loss and to improve the operation performance. (N.H.)

  15. Evidence that the Agr-like quorum sensing system regulates the toxin production, cytotoxicity and pathogenicity of Clostridium perfringens type C isolate CN3685.

    Science.gov (United States)

    Vidal, Jorge E; Ma, Menglin; Saputo, Julian; Garcia, Jorge; Uzal, Francisco A; McClane, Bruce A

    2012-01-01

    Clostridium perfringens possesses at least two functional quorum sensing (QS) systems, i.e. an Agr-like system and a LuxS-dependent AI-2 system. Both of those QS systems can reportedly control in vitro toxin production by C. perfringens but their importance for virulence has not been evaluated. Therefore, the current study assessed whether these QS systems might regulate the pathogenicity of CN3685, a C. perfringens type C strain. Since type C isolates cause both haemorrhagic necrotic enteritis and fatal enterotoxemias (where toxins produced in the intestines are absorbed into the circulation to target other internal organs), the ability of isogenic agrB or luxS mutants to cause necrotizing enteritis in rabbit small intestinal loops or enterotoxemic lethality in mice was evaluated. Results obtained strongly suggest that the Agr-like QS system, but not the LuxS-dependent AI-2 QS system, is required for CN3685 to cause haemorrhagic necrotizing enteritis, apparently because the Agr-like system regulates the production of beta toxin, which is essential for causing this pathology. The Agr-like system, but not the LuxS-mediated AI-2 system, was also important for CN3685 to cause fatal enterotoxemia. These results provide the first direct evidence supporting a role for any QS system in clostridial infections. PMID:22150719

  16. Existence of two groups of Staphylococcus aureus strains isolated from bovine mastitis based on biofilm formation, intracellular survival, capsular profile and agr-typing.

    Science.gov (United States)

    Bardiau, Marjorie; Caplin, Jonathan; Detilleux, Johann; Graber, Hans; Moroni, Paolo; Taminiau, Bernard; Mainil, Jacques G

    2016-03-15

    Staphylococcus (S.) aureus is recognised worldwide as an important pathogen causing contagious acute and chronic bovine mastitis. Chronic mastitis account for a significant part of all bovine cases and represent an important economic problem for dairy producers. Several properties (biofilm formation, intracellular survival, capsular expression and group agr) are thought to be associated with this chronic status. In a previous study, we found the existence of two groups of strains based on the association of these features. The aim of the present work was to confirm on a large international and non-related collection of strains the existence of these clusters and to associate them with case history records. In addition, the genomes of eight strains were sequenced to study the genomic differences between strains of each cluster. The results confirmed the existence of both groups based on capsular typing, intracellular survival and agr-typing: strains cap8-positive, belonging to agr group II, showing a low invasion rate and strains cap5-positive, belonging to agr group I, showing a high invasion rate. None of the two clusters were associated with the chronic status of the cow. When comparing the genomes of strains belonging to both clusters, the genes specific to the group "cap5-agrI" would suggest that these strains are better adapted to live in hostile environment. The existence of these two groups is highly important as they may represent two clusters that are adapted differently to the host and/or the surrounding environment. PMID:26931384

  17. High conversion burner type reactor

    International Nuclear Information System (INIS)

    Purpose: To simply and easily dismantle and reassemble densified fuel assemblies taken out of a high conversion ratio area thereby improve the neutron and fuel economy. Constitution: The burner portion for the purpose of fuel combustion is divided into a first burner region in adjacent with the high conversion ratio area at the center of the reactor core, and a second burner region formed to the outer circumference thereof and two types of fuels are charged therein. Densified fuel assemblies charged in the high conversion ratio area are separatably formed as fuel assemblies for use in the two types of burners. In this way, dense fuel assembly is separated into two types of fuel assemblies for use in burner of different number and arranging density of fuel elements which can be directly charged to the burner portion and facilitate the dismantling and reassembling of the fuel assemblies. Further, since the two types of fuel assemblies are charged in the burner portion, utilization factor for the neutron fuels can be improved. (Kamimura, M.)

  18. Role of the Agr-Like Quorum-Sensing System in Regulating Toxin Production by Clostridium perfringens Type B Strains CN1793 and CN1795

    OpenAIRE

    Chen, Jianming; McClane, Bruce A.

    2012-01-01

    Clostridium perfringens type B causes enteritis and enterotoxemia in domestic animals. By definition, these bacteria must produce alpha toxin (CPA), beta toxin (CPB) and epsilon toxin (ETX) although most type B strains also produce perfringolysin O (PFO) and beta2 toxin (CPB2). A recently identified Agr-like quorum-sensing (QS) system in C. perfringens controls all toxin production by surveyed type A, C, and D strains, but whether this QS is involved in regulating toxin production by type B s...

  19. Livermore pool-type reactor

    International Nuclear Information System (INIS)

    The Livermore Pool-Type Reactor (LPTR) has served a dual purpose since 1958--as an instrument for fundamental research and as a tool for measurement and calibration. Our early efforts centered on neutron-diffraction, fission, and capture gamma-ray studies. During the 1960's it was used for extensive calibration work associated with radiochemical and physical measurements on nuclear-explosive tests. Since 1970 the principal applications have been for trace-element measurements and radiation-damage studies. Today's research program is dominated by radiochemical studies of the shorter-lived fission products and by research on the mechanisms of radiation damage. Trace-element measurement for the National Uranium Resource Evaluation (NURE) program is the major measurement application today

  20. Reactor core of FBR type reactor

    International Nuclear Information System (INIS)

    Gas sealed assemblies are disposed in rows between reactor core fuel assemblies. The gas sealed assembly incorporates inflowed sodium (coolants) and sealed gas in a gas sealing cylinder and an inner hollow of a wrapper tube. A cylindrical heat generating member is disposed in the gas sealing cylinder. The sealed gas is compressed by a discharging pressure of a pump by way of sodium in the wrapper tube. During normal operation, the liquid level of the coolants is present above than a backwarding flow hole, and the temperature of the coolants is raised by the cylindrical heat generation member to raise the temperature of sodium in the backwarding flow hole. High temperature sodium is mixed with low temperature sodium from a lower flow hole at the lower portion of the backwarding flow hole, and sodium at a leak flow hole becomes sodium at a middle temperature. The temperature of the middle temperature sodium is detected by a thermometer. With such procedures, the liquid level in the gas sealed assembly can be detected and confirmed during normal operation. (I.N.)

  1. 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

  2. AGR-1 Data Qualification Interim Report

    Energy Technology Data Exchange (ETDEWEB)

    Machael Abbott

    2009-08-01

    Projects for the very-high-temperature reactor (VHTR) program provide data in support of Nuclear Regulatory Commission licensing of the VHTR. Fuel and materials to be used in the reactor are tested and characterized to quantify performance in high temperature and high fluence environments. The VHTR Program has established the NGNP Data Management and Analysis System (NDMAS) to ensure that VHTR data are (1) qualified for use, (2) stored in a readily accessible electronic form, and (3) analyzed to extract useful results. This document focuses on the first NDMAS objective. It describes the data streams associated with the first Advanced Gas Reactor (AGR-1) experiment, the processing of these data within NDMAS, and reports the interim FY09 qualification status of the AGR-1 data to date. Data qualification activities within NDMAS for specific types of data are determined by the data qualification category, which is assigned by the data generator, and include: (1) capture testing, to confirm that the data stored within NDMAS are identical to the raw data supplied, (2) accuracy testing, to confirm that the data are an accurate representation of the system or object being measured, and (3) documentation that the data were collected under an NQA-1 or equivalent QA program. The interim qualification status of the following four data streams is reported in this document: (1) fuel fabrication data, (2) fuel irradiation data, (3) fission product monitoring system (FPMS) data, and (4) Advanced Test Reactor (ATR) operating conditions data. A final report giving the NDMAS qualification status of all AGR-1 data (including cycle 145A) is planned for February 2010.

  3. AGR-1 Thermocouple Data Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Jeff Einerson

    2012-05-01

    This report documents an effort to analyze measured and simulated data obtained in the Advanced Gas Reactor (AGR) fuel irradiation test program conducted in the INL's Advanced Test Reactor (ATR) to support the Next Generation Nuclear Plant (NGNP) R&D program. The work follows up on a previous study (Pham and Einerson, 2010), in which statistical analysis methods were applied for AGR-1 thermocouple data qualification. The present work exercises the idea that, while recognizing uncertainties inherent in physics and thermal simulations of the AGR-1 test, results of the numerical simulations can be used in combination with the statistical analysis methods to further improve qualification of measured data. Additionally, the combined analysis of measured and simulation data can generate insights about simulation model uncertainty that can be useful for model improvement. This report also describes an experimental control procedure to maintain fuel target temperature in the future AGR tests using regression relationships that include simulation results. The report is organized into four chapters. Chapter 1 introduces the AGR Fuel Development and Qualification program, AGR-1 test configuration and test procedure, overview of AGR-1 measured data, and overview of physics and thermal simulation, including modeling assumptions and uncertainties. A brief summary of statistical analysis methods developed in (Pham and Einerson 2010) for AGR-1 measured data qualification within NGNP Data Management and Analysis System (NDMAS) is also included for completeness. Chapters 2-3 describe and discuss cases, in which the combined use of experimental and simulation data is realized. A set of issues associated with measurement and modeling uncertainties resulted from the combined analysis are identified. This includes demonstration that such a combined analysis led to important insights for reducing uncertainty in presentation of AGR-1 measured data (Chapter 2) and interpretation of

  4. 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.

  5. 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.

  6. 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

  7. Improvements to PWR type reactors

    International Nuclear Information System (INIS)

    Improvements to pressurized water nuclear reactors are described, where the core coolant, called primary fluid, flows under the effect of a circulating pump in a primary loop between a steam generator and a pressure vessel containing the reactor core. The steam generator includes a bundle of tubes through which flows the primary fluid which exchanges calories with a secondary fluid, generally water, entering the generator as a liquid and issuing from it as steam. After expansion in turbines and recovery in a condenser, this steam is returned to the inside of the generator. Each primary fluid circulating pump is powered by a back-pressure turbine located in parallel with the high pressure section of the main turbine and hence fed with steam taken directly from the steam generator or the main steam pipe outside it

  8. Strategies of development of reactor types

    International Nuclear Information System (INIS)

    The development of nuclear energy in the coming decades will depend on the goals followed, on the available technologies and on the strategies implemented in the world in agreement with public acceptation. This article is limited to the technical aspects of the strategies of development of reactor types: 1 - objectives; 2 - common constraints to all reactor types: safety and terrorism risks, wastes, non-proliferation, economics; 3 - different reactor types: general considerations, proven technologies (PWR, BWR, Candu), non-proven technologies but having an important experience, technologies at the design stage; 4 - energy systems and 'Generation IV forum': systems based on thermal neutron reactors and low enrichment, systems for the valorization of 238U, systems for Pu burning, systems allowing the destruction of minor actinides, thorium-based systems, the Gen IV international forum; 5 - conclusion. (J.S.)

  9. ASTEC code adaptability to CANDU type reactors

    International Nuclear Information System (INIS)

    ASTEC integral code is dedicated for severe accident (SA) analysis, mainly for PWR type reactors. In the last years, in the FP-6 NoE SARNET project framework, important efforts were focused on the extension of the ASTEC use to other reactors: WWER, RBMK, BWR and CANDU. The use of ASTEC at CANDU type reactors introduces many difficulties especially for the core degradation phenomena. The paper shows some results obtained in exploratory calculation with the modules SOPHAEROS, CPA, IODE, CESAR and DIVA in order to investigate the possibility to use or to adapt the models at CANDU type reactors. An important part of the paper is focused on the models for CANDU core degradation to be implemented in DIVAC module. (authors)

  10. 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

  11. 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

  12. Evidence that the Agr-like Quorum Sensing System Regulates the Toxin Production, Cytotoxicity and Pathogenicity of Clostridium perfringens Type C Isolate CN3685

    OpenAIRE

    Vidal, Jorge E.; Ma, Menglin; Saputo, Julian; Garcia, Jorge; Uzal, Francisco A.; McClane, Bruce A.

    2011-01-01

    C. perfringens possesses at least two functional quorum sensing (QS) systems, i.e., an Agr-like system and a LuxS-dependent AI-2 system. Both of those QS systems can reportedly control in vitro toxin production by C. perfringens but their importance for virulence has not been evaluated. Therefore, the current study assessed whether these QS systems might regulate the pathogenicity of CN3685, a C. perfringens type C strain. Since type C isolates cause both hemorrhagic necrotic enteritis and fa...

  13. AGR-1, AGR-2 and AGR-3/4 Dimensional Change Data Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Herberger, Sarah E. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-02-01

    A series of Advanced Gas Reactor (AGR) experiments have been completed in the Advanced Test Reactor at Idaho National Laboratory in support of qualification and development of tristructural isotropic fuel. Each AGR test consists of multiple independently controlled and monitored capsules containing fuel compacts placed in a graphite cylinder. These capsules are instrumented with thermocouples embedded in the graphite, enabling temperature control. The fuel compacts are composed of fuel particles surrounded by a graphitic A3 matrix material. Dimensional change in AGR fuel compacts is vital because the swelling or shrinkage affects the size of the gas gaps that are used to control temperatures. Analysis of dimensional change in the AGR fuel compacts is needed to establish the variables directly relating to compact shrinkage. The variables initially identified for consideration were matrix density, compact density, fuel packing fraction, uranium loading, fuel particle diameter, cumulative fast neutron fluence, and volume average time average fuel temperature. In addition to the data available from the AGR experiments, the analysis included specimens formed from the same A3 matrix material used in Advanced Graphite Creep (AGC) experiments, which provide graphite creep data during irradiation for design and licensing purposes. The primary purpose of including the AGC specimens was to encompass dimensional behavior at zero packing fraction, zero uranium loading, and zero particle diameter. All possible combinations of first-order variable regressions were considered in the analysis. The study focused on identifying the best regression models for percent change in diameter, length, and volume. Bootstrap analysis was used to ensure the resulting regression models were robust and well-performing. The variables identified as very significant in predicting change in one or more dimensions (diameter, length, and volume) are volume average time average temperature, fast fluence

  14. Moving hot cell for LMFBR type reactor

    International Nuclear Information System (INIS)

    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.)

  15. 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.)

  16. AGR core safety assessment methodologies

    International Nuclear Information System (INIS)

    To demonstrate the safety of its gas-cooled graphite-moderated AGR reactors, nuclear safety assessments of the cores are based upon a methodology which demonstrates no component failures, geometrical stability of the structure and material properties bounded by a database. All AGRs continue to meet these three criteria. However, predictions of future core behaviour indicate that the safety case methodology will eventually need to be modified to deal with new phenomena. A new approach to the safety assessment of the cores is currently under development, which can take account of these factors while at the same time providing the same level of protection for the cores. This approach will be based on the functionality of the core: unhindered movement of control rods, continued adequate cooling of the fuel and the core, continued ability to charge and discharge fuel. (author). 5 figs

  17. Review of the use and state of development of the various reactor types

    International Nuclear Information System (INIS)

    The report gives a review of the reactor types being of importance from today's point of view for use as stationary power reactors. These are heavy water reactors, light water reactors (pressurized water reactor, Soviet pressurized water reactor, Soviet light-water-graphite reactors, boiling water reactors), gas-cooled reactors (gas-graphite reactors, high temperature reactors), and fast breeder reactors. (HJ)

  18. Scottish Nuclear Limited: the AGR - past, present and future

    International Nuclear Information System (INIS)

    This article reviews the historical development of the AGR (advanced gas-cooled reactor) in Scotland from its inception to its current successfully established position. It examines where the AGRs will go in the future and concludes with the strategic role of the existing and new plant in the market-led electricity supply industry. (Author)

  19. 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.)

  20. Fuel assembly for BWR type reactor

    International Nuclear Information System (INIS)

    A fuel assembly has a 9 x 9 square lattice arrangement having a water channel which occupies an area of 3 x 3 lattice pattern corresponding to 9 fuel rods. Fuel pellets comprise those of not more 7 kinds which have fission products at enrichment degrees different by a spun of not less than 10%. Fuel rods comprise from 4 to 12 first type fuel rods and remaining second type fuel rods. The first type fuel rod is loaded with fuel pellets of fissionable products having an enrichment degree axially different at the upper and the lower portions. The second type fuel rod is loaded with fuel pellets of fissionable products having the same enrichment degree in the vertical direction. With such a constitution, the enrichment degree of fissionable products of fuel pellets in the fuel assembly for a BWR type reactor having different reactor constitution and operation conditions can be used in common. Accordingly, the degree of freedom for the design of the distribution of the enrichment degree is increased. (I.N.)

  1. Safety testing of AGR-2 UO2 compacts 3-3-2 and 3-4-2

    Energy Technology Data Exchange (ETDEWEB)

    Hunn, John D. [Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States); Morris, Robert Noel [Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States); Baldwin, Charles A. [Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States); Montgomery, Fred C. [Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States)

    2015-09-01

    Post-irradiation examination (PIE) is in progress on tristructural-isotropic (TRISO) coated-particle fuel compacts from the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program second irradiation experiment (AGR-2) [Collin 2014]. The AGR-2 PIE will build upon new information and understanding acquired throughout the recently-concluded six-year AGR-1 PIE campaign [Demkowicz et al. 2015] and establish a database for the different AGR-2 fuel designs.

  2. Safety testing of AGR-2 UO2 compacts 3-3-2 and 3-4-2

    International Nuclear Information System (INIS)

    Post-irradiation examination (PIE) is in progress on tristructural-isotropic (TRISO) coated-particle fuel compacts from the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program second irradiation experiment (AGR-2) [Collin 2014]. The AGR-2 PIE will build upon new information and understanding acquired throughout the recently-concluded six-year AGR-1 PIE campaign [Demkowicz et al. 2015] and establish a database for the different AGR-2 fuel designs.

  3. Steam generating system in LMFBR type reactors

    International Nuclear Information System (INIS)

    Purpose: To suppress the thermal shock loads to the structures of reactor system and secondary coolant system, for instance, upon plant trip accompanying turbine trip in the steam generation system of LMFBR type reactors. Constitution: Additional feedwater heater is disposed to the pipeway at the inlet of a steam generator in a steam generation system equipped with a closed loop extended from a steam generator by way of a gas-liquid separator, a turbine and a condensator to the steam generator. The separated water at high temperature and high pressure from a gas-liquid separator is heat exchanged with coolants flowing through the closed loop of the steam generation system in non-contact manner and, thereafter, introduced to a water reservoir tank. This can avoid the water to be fed at low temperature as it is to the steam generator, whereby the thermal shock loads to the structures of the reactor system and the secondary coolant system can be suppressed. (Moriyama, K.)

  4. 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

  5. Fuel assembly for BWR type reactor

    International Nuclear Information System (INIS)

    A fuel assembly of an BWR type reactor of present invention, in which a plurality of fuel rods are arranged in a regular square lattice like configuration include vibration-filled fuel rods in which granular nuclear fuel materials and granular non-nuclear fuel materials having a smaller neutron absorbing cross sectional area are mixed and filled. With such a constitution, the content of the mixed and filled non-nuclear fuel materials in the vibration filled fuel rods is at least 20% by a volume ratio in average in fuel assemblies. In addition, a burnable poison is optionally added and mixed to the granular mixture of the nuclear fuel material and the diluting granules. With such a constitution, the manufacturing cost can be reduced, and the combustion rate of the nuclear fission materials is increased to improve reactor core characteristics, thereby enabling to obtain sufficient Pu loading amount per assembly, and fuel assemblies excellent in flexibility in design and economic property can be obtained. (T.M.)

  6. Design guide for Category III reactors: pool type reactors

    International Nuclear Information System (INIS)

    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

  7. AgrAbility Project

    Science.gov (United States)

    ... hours ago AgrAbility's 25 Years, 25 Stories Rosendo Ramirez is a farmworker in California, a job that ... Rosendo continues to work. www.agrability.org/25years/ramirez/ ... See More See Less Rosendo Ramirez: Ingenuity, perseverance, ...

  8. Determining a pool - type reactor fuel policy

    International Nuclear Information System (INIS)

    Refuelling the 10 to 15 MW pool type reactor considered here will occur frequently (some 10 elements every 3 to 4 weeks). It is therefore necessary to determine the most economic fuel policy. This study proposes to define a strategy that will make it possible to decide on the number and characteristics of the shipment containers, as well as on the means of storage, so as to reduce the risks as much as possible should the basic parameters of the study vary. Among these parameters, the respective influence of which is investigated, chemical reprocessing costs play a vital part. Two examples of optimum fuel management are given according to whether the reprocessing charges applied are those of the old or of the 1961 U.S. AEC base charges for reprocessing highly enriched irradiated fuel. (authors)

  9. Sodium cooling FBR type reactor plant

    International Nuclear Information System (INIS)

    In a sodium cooling FBR type reactor plant, a steam electrolysis type hydrogen forming device using high temperature steams as the starting material and a steam turbine power generator operated by high pressure/high temperature steams are disposed. During day time when a power demand is increased, a steam switch valve on the side of the hydrogen forming device is closed and the steam switch valve on the side of the power generator is opened to introduce substantially entire amount of steams to the steam turbine. During mid night when the power demand is decreased, the steam switch valve on the side of the power generator is closed and the steam switch valve on the side of the hydrogen forming device is opened to introduce substantially entire amount of steams to the steam electrolysis device, or the exhausted gases from the steam turbine type power generator is heated again by a heat exchanger and low pressure/high temperature steams are introduced to the steam electrolysis device. In this case, electric power is applied between a hydrogen electrode and an oxygen electrode to form a hydrogen gas and an oxygen gas, and the hydrogen gas is stored in a hydrogen storage vessel. It can easily cope with the fluctuation of the power demand, as well as hydrogen can be efficiently produced. (N.H.)

  10. AGR-1 Irradiation Test Final As-Run Report

    Energy Technology Data Exchange (ETDEWEB)

    Blaise P. Collin

    2012-06-01

    This document presents the as-run analysis of the AGR-1 irradiation experiment. AGR-1 is the first of eight planned irradiations for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. Funding for this program is provided by the US Department of Energy (DOE) as part of the Next-Generation Nuclear Plant (NGNP) project. The objectives of the AGR-1 experiment are: 1. To gain experience with multi-capsule test train design, fabrication, and operation with the intent to reduce the probability of capsule or test train failure in subsequent irradiation tests. 2. To irradiate fuel produced in conjunction with the AGR fuel process development effort. 3. To provide data that will support the development of an understanding of the relationship between fuel fabrication processes, fuel product properties, and irradiation performance. In order to achieve the test objectives, the AGR-1 experiment was irradiated in the B-10 position of the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) for a total duration of 620 effective full power days of irradiation. Irradiation began on December 24, 2006 and ended on November 6, 2009 spanning 13 ATR cycles and approximately three calendar years. The test contained six independently controlled and monitored capsules. Each capsule contained 12 compacts of a single type, or variant, of the AGR coated fuel. No fuel particles failed during the AGR-1 irradiation. Final burnup values on a per compact basis ranged from 11.5 to 19.6 %FIMA, while fast fluence values ranged from 2.21 to 4.39 ?1025 n/m2 (E >0.18 MeV). We’ll say something here about temperatures once thermal recalc is done. Thermocouples performed well, failing at a lower rate than expected. At the end of the irradiation, nine of the originally-planned 19 TCs were considered functional. Fission product release-to-birth (R/B) ratios were quite low. In most capsules, R/B values at the end of the irradiation were at or below 10-7 with only one

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

    International Nuclear Information System (INIS)

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

  12. AGR-1 Safety Test Predictions using the PARFUME code

    Energy Technology Data Exchange (ETDEWEB)

    Blaise Collin

    2012-05-01

    The PARFUME modeling code was used to predict failure probability of TRISO-coated fuel particles and diffusion of fission products through these particles during safety tests following the first irradiation test of the Advanced Gas Reactor program (AGR-1). These calculations support the AGR-1 Safety Testing Experiment, which is part of the PIE effort on AGR-1. Modeling of the AGR-1 Safety Test Predictions includes a 620-day irradiation followed by a 300-hour heat-up phase of selected AGR-1 compacts. Results include fuel failure probability, palladium penetration, and fractional release of fission products. Results show that no particle failure is predicted during irradiation or heat-up, and that fractional release of fission products is limited during irradiation but that it significantly increases during heat-up.

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

    International Nuclear Information System (INIS)

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

  14. Emergency reactor core cooling system of BWR type reactor

    International Nuclear Information System (INIS)

    The present invention provides an emergency reactor core cooling system which can reduce a capacity of a power source required upon occurrence of emergency, extending an start-up time of an emergency reactor core cooling system (ECCA) to provide a plant endurable to a common factor accident and can provide time margin up to the start-up time. Namely, the system of the present invention comprises a division I equipped with an isolation condenser (IC), an after-heat removing system (low pressure system)(LPFL/RHR) and an emergency gas turbine generator (GT), a division II equipped with a diesel driving water injection system (high pressure system)(HDIS), LPFL/RHR, and GT, and a division III equipped with a reactor isolation time cooling system (high pressure system)(ARCIC), LPFL/RHR and GT. With such a constitution, since the IC, HDIS and ARCIC are used in combination as a high pressure system, an electromotive pump required to be operated upon high pressure state can be saved. In addition, if a static reactor cooling system (PCCS) is adopted and is provided with a back-up function for LPFL/RHR with respect to heat removal of the container upon occurrence of an accident, the countermeasure for occurrence of severe accidents can be enhanced. (I.S.)

  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. Aging of reactor vessels in LWR type reactors

    International Nuclear Information System (INIS)

    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. Plenum separator system for pool-type nuclear reactors

    International Nuclear Information System (INIS)

    A pool-type liquid-metal-heat-transfer nuclear reactor is described which consists of: a vertically disposed generally cylindrical reactor vessel having a closed bottom portion, and a closure head atop the reactor vessel and closing the reactor vessel; the reactor vessel enclosing the major components of the nuclear reactor which include a reactor core supported at a centrally disposed lower portion of the reactor vessel; a bottom-supported gas-plenum-forming hollow cylindrical member closed at its upper end, the hollow cylindrical member sealed to and supported by the reactor vessel; a hollow cylindrically conformed neutron shield member spaced from and radially surrounding the reactor core; separate liquid-metal plena confining liquid metal during normal reactor operation and comprising a hot upper plenum, a cold lower plenum and intermediate temperature plena; the liquid metal intakes of the liquid metal pumps positioned in the cold lower plenum with the cooler liquid metal therein being pumped upwardly through the reactor core to be heated and exit therefrom in a turbulent fashion; and the liquid metal intakes of the heat exchangers positioned within the hot upper plenum and the liquid metal discharges of the heat exchangers positioned within the cold lower plenum to discharge cooled liquid metal into the cold lower plenum

  18. Safety protection device for BWR type reactors

    International Nuclear Information System (INIS)

    Purpose: To enable to prevent significant reduction in the reactor water level in a case where motor driven feedwater pumps used only at an extremely low frequency are saved and the remaining pumps are failed to be operated. Constitution: Cooling systems upon reactor isolation are extraordinarily operated by using a logic product signal between a trip signal for motor driven feedwater pumps (MDRFP) that are responsible to the small flow rate feed operation and a stop signal for turbine driven feedwater pumps that are responsible to the high flow rate feedwater operation. That is, in a case where motor driven feed water pumps used for a short period of time upon starting should fail to disable the water feeding, the cooling systems upon reactor isolation are started to suppress the reduction in the reactor water level if the condition that the turbine driven feedwater is interrupted is satisfied. (Horiuchi, T.)

  19. Reflector driving device of FBR type reactor

    International Nuclear Information System (INIS)

    The present invention provides a reflector driving device which eliminates driving at outside and rapidly start-up and shuts down the reactor. Namely, the device has an electromagnetic holding mechanism attracted to a reactor core barrel by electromagnetic attracting force. Electromagnetic repulsion coils for generating instantaneous electromagnetic repulsion force are disposed between the reflector and the electromagnet retaining mechanism. The reflector is driven using inertia force of the electromagnetic repulsion force and the reflector. Then, the reflector driving device is attracted and held at a start-up position of the reactor upon start-up of the reactor. The reflector elevated by fluid pressure is secured at a position of the start-up of the reactor while having the reflector driving device as a stopper. The reflector moves at an extremely slow speed while being secured to the reflector driving device along with the elevation of the reflector driving device. As a result, a driving shaft sealing material which supports a conventional driving device at the outside and the reflector is eliminated thereby simplifying the structure of the outside of the reactor. (I.S.)

  20. Uncertainty Quantification of Calculated Temperatures for the AGR-1 Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Binh T. Pham; Jeffrey J. Einerson; Grant L. Hawkes

    2013-03-01

    This report documents an effort to quantify the uncertainty of the calculated temperature data for the first Advanced Gas Reactor (AGR-1) fuel irradiation experiment conducted in the INL’s Advanced Test Reactor (ATR) in support of the Next Generation Nuclear Plant (NGNP) R&D program. Recognizing uncertainties inherent in physics and thermal simulations of the AGR-1 test, the results of the numerical simulations can be used in combination with the statistical analysis methods to improve qualification of measured data. Additionally, the temperature simulation data for AGR tests can be used for validation of the fuel transport and fuel performance simulation models. The crucial roles of the calculated fuel temperatures in ensuring achievement of the AGR experimental program objectives require accurate determination of the model temperature uncertainties. The report is organized into three chapters. Chapter 1 introduces the AGR Fuel Development and Qualification program and provides overviews of AGR-1 measured data, AGR-1 test configuration and test procedure, and thermal simulation. Chapters 2 describes the uncertainty quantification procedure for temperature simulation data of the AGR-1 experiment, namely, (i) identify and quantify uncertainty sources; (ii) perform sensitivity analysis for several thermal test conditions; (iii) use uncertainty propagation to quantify overall response temperature uncertainty. A set of issues associated with modeling uncertainties resulting from the expert assessments are identified. This also includes the experimental design to estimate the main effects and interactions of the important thermal model parameters. Chapter 3 presents the overall uncertainty results for the six AGR-1 capsules. This includes uncertainties for the daily volume-average and peak fuel temperatures, daily average temperatures at TC locations, and time-average volume-average and time-average peak fuel temperatures.

  1. Analysis of Gln223Agr Polymorphism of Leptin Receptor Gene in Type II Diabetic Mellitus Subjects among Malaysians

    OpenAIRE

    Chong Pei Pei; Ahmad Khairuddin Mohamed Yusof; Seyyed Reza Pishva; Ahmad Fazli Abdul Aziz; Farzad Heidari; Vasudevan Ramachandran; Ali Etemad; Patimah Ismail

    2013-01-01

    Leptin is known as the adipose peptide hormone. It plays an important role in the regulation of body fat and inhibits food intake by its action. Moreover, it is believed that leptin level deductions might be the cause of obesity and may play an important role in the development of Type 2 Diabetes Mellitus (T2DM), as well as in cardiovascular diseases (CVD). The Leptin Receptor (LEPR) gene and its polymorphisms have not been extensively studied in relation to the T2DM and its complications in ...

  2. AGR-2 AND AGR-3/4 RELEASE-TO-BIRTH RATIO DATA ANALYSIS

    Energy Technology Data Exchange (ETDEWEB)

    Pham, Binh T; Einerson, Jeffrey J; Scates, Dawn M; Maki, John T; Petti, David A

    2014-09-01

    A series of Advanced Gas Reactor (AGR) irradiation tests is being conducted in the Advanced Test Reactor at Idaho National Laboratory in support of development and qualification of tristructural isotropic (TRISO) low enriched fuel used in the High Temperature Gas-cooled Reactor (HTGR). Each AGR test consists of multiple independently controlled and monitored capsules containing fuel compacts placed in a graphite cylinder shrouded by a steel shell. These capsules are instrumented with thermocouples embedded in the graphite enabling temperature control. AGR configuration and irradiation conditions are based on prismatic HTGR technology distinguished primarily by the use of helium coolant, a low-power-density ceramic core capable of withstanding very high temperatures, and TRISO coated particle fuel. Thus, these tests provide valuable irradiation performance data to support fuel process development, qualify fuel for normal operating conditions, and support development and validation of fuel performance and fission product transport models and codes. The release-rate-to-birth-rate ratio (R/B) for each of fission product isotopes (i.e., krypton and xenon) is calculated from release rates in the sweep gas flow measured by the germanium detectors used in the AGR Fission Product Monitoring (FPM) System installed downstream from each irradiated capsule. Birth rates are calculated based on the fission power in the experiment and fission product generation models. Thus, this R/B is a measure of the ability of fuel kernel, particle coating layers, and compact matrix to retain fission gas atoms preventing their release into the sweep gas flow, especially in the event of particle coating failures that occurred during AGR-2 and AGR-3/4 irradiations. The major factors that govern gaseous radioactive decay, diffusion, and release processes are found to be material diffusion coefficient, temperature, and isotopic decay constant. For each of all AGR capsules, ABAQUS-based three

  3. Steam shut-off valves for PWR type reactors

    International Nuclear Information System (INIS)

    Fast acting closure means are requested in PWR type reactors as well as in BWR to safely shut-off the live steam at the turbine input in the event of accident. The design and control system of steam shut-off valves acted by the fluid system and intended for PWR type reactors, are described. The role of these valves in a PWR is discussed with the specified requirements involved

  4. Exploitation questions regarding channel type reactors: water graphite channel reactors (operation, reconstruction, advantages and disadvantages)

    International Nuclear Information System (INIS)

    An overview of up-grade of the RBMK-type reactors is given. I this paper the core design and core monitoring, pressure boundary integrity, RBMK basic design and safety improvements emergency core cooling system (ECCS) as well as reactor cavity overpressure protection system (RCOPS) are discussed

  5. 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.).

  6. Fabrication and Comparison of Fuels for Advanced Gas Reactor Irradiation Tests

    Energy Technology Data Exchange (ETDEWEB)

    Jeffrey Phillips; Charles Barnes; John Hunn

    2010-10-01

    As part of the program to demonstrate TRISO-coated fuel for the Next Generation Nuclear Plant, a series of irradiation tests of Advanced Gas Reactor (AGR) fuel are being performed in the Advanced Test Reactor (ATR) at the Idaho National Laboratory. In the first test, called “AGR-1,” graphite compacts containing approximately 300,000 coated particles were irradiated from December 2006 until November 2009. Development of AGR-1 fuel sought to replicate the properties of German TRISO-coated particles. No particle failures were seen in the nearly 3-year irradiation to a burn up of 19%. The AGR-1 particles were coated in a two-inch diameter coater. Following fabrication of AGR-1 fuel, process improvements and changes were made in each of the fabrication processes. Changes in the kernel fabrication process included replacing the carbon black powder feed with a surface-modified carbon slurry and shortening the sintering schedule. AGR-2 TRISO particles were produced in a six-inch diameter coater using a change size about twenty-one times that of the two-inch diameter coater used to coat AGR-1 particles. Changes were also made in the compacting process, including increasing the temperature and pressure of pressing and using a different type of press. Irradiation of AGR-2 fuel began in late spring 2010. Properties of AGR-2 fuel compare favorably with AGR-1 and historic German fuel. Kernels are more homogeneous in shape, chemistry and density. TRISO-particle sphericity, layer thickness standard deviations, and defect fractions are also comparable. In a sample of 317,000 particles from deconsolidated AGR-2 compacts, 3 exposed kernels were found in a leach test. No SiC defects were found in a sample of 250,000 deconsolidated particles, and no IPyC defects in a sample of 64,000 particles. The primary difference in properties between AGR-1 and AGR-2 compacts is that AGR-2 compacts have a higher matrix density, 1.6 g/cm3 compared to about 1.3 g/cm3 for AGR-1 compacts. Based on

  7. Fabrication and Comparison of Fuels for Advanced Gas Reactor Irradiation Tests

    International Nuclear Information System (INIS)

    As part of the program to demonstrate TRISO-coated fuel for the Next Generation Nuclear Plant, a series of irradiation tests of Advanced Gas Reactor (AGR) fuel are being performed in the Advanced Test Reactor (ATR) at the Idaho National Laboratory. In the first test, called 'AGR-1,' graphite compacts containing approximately 300,000 coated particles were irradiated from December 2006 until November 2009. Development of AGR-1 fuel sought to replicate the properties of German TRISO-coated particles. No particle failures were seen in the nearly 3-year irradiation to a burn up of 19%. The AGR-1 particles were coated in a two-inch diameter coater. Following fabrication of AGR-1 fuel, process improvements and changes were made in each of the fabrication processes. Changes in the kernel fabrication process included replacing the carbon black powder feed with a surface-modified carbon slurry and shortening the sintering schedule. AGR-2 TRISO particles were produced in a six-inch diameter coater using a change size about twenty-one times that of the two-inch diameter coater used to coat AGR-1 particles. Changes were also made in the compacting process, including increasing the temperature and pressure of pressing and using a different type of press. Irradiation of AGR-2 fuel began in late spring 2010. Properties of AGR-2 fuel compare favorably with AGR-1 and historic German fuel. Kernels are more homogeneous in shape, chemistry and density. TRISO-particle sphericity, layer thickness standard deviations, and defect fractions are also comparable. In a sample of 317,000 particles from deconsolidated AGR-2 compacts, 3 exposed kernels were found in a leach test. No SiC defects were found in a sample of 250,000 deconsolidated particles, and no IPyC defects in a sample of 64,000 particles. The primary difference in properties between AGR-1 and AGR-2 compacts is that AGR-2 compacts have a higher matrix density, 1.6 g/cm3 compared to about 1.3 g/cm3 for AGR-1 compacts. Based on fuel

  8. Control rod drives for HTGR type reactor

    International Nuclear Information System (INIS)

    The device of the present invention has a feature of having stable braking characteristics upon scram operation of control rods. That is, control rod drives are moved upon and down by a dram which rotates the control rod suspended from to a wire rope, and the dram is disconnected from the driving mechanism by a crutch mechanism upon scram, to rapidly insert the control rod in the reactor by its own weight. An electric generator is used as a braking mechanism for controlling the scram speed of the control rod. A plurality of resistors disposed outside of the reactor coolants boundary are connected in parallel between input/output terminals of the electric generator. With such a constitution, braking characteristics are determined by the intensity of the permanent magnet, number of the coil windings and values of the resistors constituting the power generator. Accordingly, the braking characteristics are less changed relative to the working circumstantial conditions, the history of use and the state of mounting. As a result, stable braking characteristics can always be obtained. Further, braking characteristics can easily be controlled by varying the resistance value. (I.S.)

  9. AGR-2 Data Qualification Report for ATR Cycle 154B

    Energy Technology Data Exchange (ETDEWEB)

    Binh Pham; Jeff Einerson

    2014-01-01

    This report provides the data qualification status of Advanced Gas Reactor-2 (AGR-2) fuel irradiation experimental data from Advanced Test Reactor (ATR) Cycle 154B as recorded in the Nuclear Data Management and Analysis System (NDMAS). This is the last cycle of AGR-2 irradiation, as the test train was pulled from the ATR core during the outage portion of ATR Cycle 155A. The AGR-2 data streams addressed in this report include thermocouple (TC) temperatures, sweep gas data (flow rates including new Fission Product Monitoring (FPM) downstream flows from Fission Product Monitoring System (FPMS) detectors, pressure, and moisture content), and FPMS data (release rates and release-to-birth rate ratios [R/Bs]) for each of the six capsules in the AGR-2 experiment. The final data qualification status for these data streams is determined by a Data Review Committee (DRC) comprised of AGR technical leads, Sitewide Quality Assurance (QA), and NDMAS analysts. The Data Review Committee reviewed the data acquisition process, considered whether the data met the requirements for data collection as specified in QA-approved Very High Temperature Reactor (VHTR) data collection plans, examined the results of NDMAS data testing and statistical analyses, and confirmed the qualification status of the data as given in this report.

  10. Serious accidents of PWR type reactors for power generation

    International Nuclear Information System (INIS)

    This document presents the great lines of current knowledge on serious accidents relative to PWR type reactors. First, is exposed the physics of PWR type reactor core meltdown and the possible failure modes of the containment building in such a case. Then, are presented the dispositions implemented with regards to such accidents in France, particularly the pragmatic approach that prevails for the already built reactors. Then, the document tackles the case of the European pressurized reactor (E.P.R.), for which the dimensioning takes into account explicitly serious accidents: it is a question of objectives conception and their respect must be the object of a strict demonstration, by taking into account uncertainties. (N.C.)

  11. 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.

  12. Safety device for separated type nuclear superheating reactor

    International Nuclear Information System (INIS)

    In the present invention, two nuclear reactors constituting a separated type nuclear superheating reactor can surely be cooled in any of events. That is, the device of the present invention comprises (1) a saturated steam generating portion on the upstream as a first reactor, (2) a superheated steam heating portion on the downstream as a second reactor, (3) an emergency condensator for introducing and condensing steams generated in the superheated steam heating portion, (4) a relief valve for releasing generated steams and (5) a double-walled pipe which communicates both of the reactors and has a leakage detection performance. Then, steams generated in the superheated steam heating portion are introduced into the emergency condensator to be condensed. Descending pipelines are disposed for leading the condensed condensates to the saturated steam generator. With such constitution, two reactors constituting the separated type superheating reactor can surely be cooled in any of events, including abnormal transient phase, medium and small-scale ruptures and large-scale ruptures. (I.S.)

  13. Three-dimensional reactor dynamics code for VVER type nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kyrki-Rajamaeki, R. [VTT Energy, Espoo (Finland)

    1995-10-01

    A three-dimensional reactor dynamics computer code has been developed, validated and applied for transient and accident analyses of VVER type nuclear reactors. This code, HEXTRAN, is a part of the reactor physics and dynamics calculation system of the Technical Research Centre of Finland, VTT. HEXTRAN models accurately the VVER core with hexagonal fuel assemblies. The code uses advanced mathematical methods in spatial and time discretization of neutronics, heat transfer and the two-phase flow equations of hydraulics. It includes all the experience of VTT from 20 years on the accurate three-dimensional static reactor physics as well as on the one-dimensional reactor dynamics. The dynamic coupling with the thermal hydraulic system code SMABRE also allows the VVER circuit-modelling experience to be included in the analyses. (79 refs.).

  14. Fuel assembly for BWR type reactor

    International Nuclear Information System (INIS)

    When fuel rods are suddenly oscillated by earthquakes, and a void ratio is abruptly reduced, it is forecast that feed back of negative reactivity due to generation of voids is delayed to cause power increase in a short period of time. Then, in a fuel assembly comprising a large number of fuel rods bundled by an upper tie plate, a lower tie plate and a plurality of spacers and contained in a channel box, stirring means for coolants flowing the periphery of fuel rods are disposed in a lower sub-cool boiling region. Coolants flown into the fuel assembly are directed to fuel rods by the coolant stirring means to mix the coolants, whereby the temperature difference between the periphery of the surface of the fuel rods and bulk coolants is reduced, to decrease a sub-cool void amount. Then, even if the fuel rods are oscillated, the reduction of a sub-cool void ratio is small, which scarcely gives influences of fuel rod oscillation on the power of the reactor core. (N.H.)

  15. 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.)

  16. 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.)

  17. Comparison of nuclear reactor types of the next generation

    International Nuclear Information System (INIS)

    The paper presents a comparison for a selected relevant set of parameters for different commercial nuclear reactor types at the next generation. This parameters overview could serve as the base for the semi-quantitative decision bases for the selection of the future nuclear strategy. The number of advanced reactor designs of the LWR, HWR, GCR and LMR type are presented. Even currently many of them are still on the drawing boards, the concepts and designs should be assessed in the sense of sensible approach for planning the possible future nuclear strategy. (author)

  18. Control method for water quality of BWR type reactor

    International Nuclear Information System (INIS)

    The present invention provides a method of suppressing radiation exposure upon periodical inspection of a BWR type reactor, suppressing leaching of radioactive materials deposited and activated on fuels, and reducing radioactive deposition on pipelines and equipments made of a carbon steel and austenite stainless steel. Namely, control of water quality described below is conducted under the conditions that the Ni metal ion concentration is from 2 to 10ppb and the Zn metal ion concentration of from 3 to 15ppb in reactor water. (1) controlling the water quality based on neutral/purified water during normal operation and upon injection of hydrogen, (2) using fuels having spring members made of a Ni based alloy processed by aging hardening in atmospheric air, (3) using reactor water recycling pipelines made of an electrolyzed and polished austenite stainless steel, and (4) using carbon steel or low alloy steel for pipelines and equipments of a reactor system. (I.S.)

  19. Development and enhancement potentials of Eastern design-type reactors

    International Nuclear Information System (INIS)

    There are approximately 125 nuclear power plants with Eastern design-type reactors in operation, under construction, or shut down in the countries of central and eastern Europe and in the CIS. Their backfitting is financed by worldwide support at a cost of currently 1.5 billion Deutschmarks. Enhancement activities performed in Russia concentrate on the three major designs, PWR reactors (WWER), breeder reactors (BN), and channel-cooled reactors (RBMK), in order to achieve improved designs for future construction of new plant to replace existing ones. The planning activities for new construction got as far as establishing outline plans, and there are only six more concrete plans providing for new construction of six nuclear power plants based on existing designs, with backfitting requirements to be met for engineered safety. (orig.)

  20. 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.)

  1. Safety criteria for plate type fuels in small power reactors

    International Nuclear Information System (INIS)

    Full text: Plate type fuels are generally used in research reactor, and could also be used in small or large power reactors, but would require that a comprehensive set of safety criteria be established. 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. Several alternatives of plate type fuels are considered for using 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, the heat transfer safety criteria were verified for all the alternatives, namely the DNBR, peak clad temperature to avoid clad embrittlement, meat temperature to avoid swelling, fuel temperature to avoid meat matrix reaction and coolant critical speed. It was found that all alternatives meet the safety criteria including the 0.5 mm monolithic U-Mo plate cladded with zircaloy. (author)

  2. Chemical cleaning of UK AGR boilers

    International Nuclear Information System (INIS)

    For a number of years, the waterside pressure drops across the advanced gas-cooled reactor (AGR) pod boilers have been increasing. The pressure drop increases have accelerated with time, which is the converse behaviour to that expected for rippled magnetite formation (rapid initial increase slowing down with time). Nonetheless, magnetite deposition remains the most likely cause for the increasing boiler resistances. A number of potential countermeasures have been considered in response to the boiler pressure drop increases. However, there was no detectable reduction in the rate of pressure drop increase. Chemical cleaning was therefore considered and a project to substantiate and then implement chemical cleaning was initiated. (authors)

  3. 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

  4. 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)

  5. Water injection system for turbine driven BWR type reactor

    International Nuclear Information System (INIS)

    The present invention provides a water injection system of a turbine driven nuclear reactor for maintaining the function thereof even upon occurrence of a severe accident in a BWR type nuclear reactor. That is, the system comprises a differential pressure detection means for measuring a pressure difference between the downstream of a the turbine and a reactor container and an interrupting means for stopping the supply of steams to the turbine when the differential pressure exceeds a predetermined value. With such a constitution, when the pressure in the turbine driven water injection system is locally increased, the differential pressure detection means detects the differential pressure, to interrupt the supply of the steams to the turbine. Further, upon occurrence of a severe accident that a pressure in the reactor container is abnormally elevated, differential pressure is not caused between the downstream of the turbine and the reactor container. Accordingly, a protection function is not operated by the differential pressure detection means. Accordingly, injection of coolants to the reactor can be continued even upon loss of AC power source. (I.S.)

  6. TMLB'-sequence simulation for a VVER-1000-type reactor

    International Nuclear Information System (INIS)

    During the last decade, extensive work has been performed to calculate the source term behavior for western-type reactors. However, the analysis of severe accidents for VVER-type reactors has just started. To investigate the source term behavior, a number of tools have been developed and tested with success. Among these tools, the Source Term Code Package (STCP) was selected to perform a source term analysis for VVER-type reactors. The input data for this case are based only on typical VVER-1000 features and not on a specific power plant design. For this first approach, no new models were added to the STCP. The selected accident sequence is a transient-initiated event with failure of all makeup to the primary and secondary systems as well as the failure of all active containment safety features (TMLB'). The goal of this work was to investigate the behavior of fission products and aerosols generated during a severe accident in a VVER-1000-type reactor

  7. Optical techniques for nuclear reactor inspection

    International Nuclear Information System (INIS)

    Optical inspection techniques available and relevant to the various stages of the life cycle of a nuclear reactor are briefly reviewed. Experience in the three main types of nuclear reactor of interest to the CEGB, Magnox, AGR and PWR, is discussed. Conventional optical systems and stereoscopic viewing systems are described together with specialized and novel techniques, mainly Marchwood Engineering Laboratory's developments, which have proved valuable in tackling a variety of inspection problems. (U.K.)

  8. Numerical simulation of melt behaviour in WWER type reactor vessel

    International Nuclear Information System (INIS)

    Event scenario, physical processes and models of SOKRAT code relating to the stage of melt location in WWER-1000 type reactor vessel at severe accident are described. The results of test calculations and calculations of thermal interaction of melt with reactor vessel for scenario Large leakage are given. Cross-verification of calculating code for simulating convection heat transfer and melt region propagation has been conducted, commercial code Fluent 6.3 has been used for it. The results obtained according to hydrodynamic code Fluent 6.3 and SOKRAT code agree well

  9. Neutron design of BN 600 type reactor with plutonium fuel

    International Nuclear Information System (INIS)

    Briefly described is the neutron physics design of a fast reactor of the BN 600 type burning plutonium fuel. The basic specifications of the reactor are given prior to steady-state refuelling and after it. Also presented are the indices of the fuel cycle, such as the balance of heavy isotopes during refuelling and data on fuel burnup. Computations of the reactivity of one compensation fuel assembly were made for a homogenized fuel assembly in the central part and the efficiency studied of the whole system of compensation. (B.S.)

  10. Some aspects of FBR type reactors in-service inspection

    International Nuclear Information System (INIS)

    Inspection of fast reactors during operation in France involves a general definition. Is objective is to ensure the ability of all the components to fulfill the functions they were designed for. The program of inspection during operation of Superphenix reactor is in the course of elaboration in cooperation with the safety regulatory authorities. It is possible at this stage to define in-service inspection of three types: inspection that would ensure safety of the environment and the population; inspection of functions whose loss could lead to instantaneous malfunctioning of the plant; and surveillance of the functions of the components whose loss could be irreversible

  11. AGR-1 Post Irradiation Examination Final Report

    International Nuclear Information System (INIS)

    The post-irradiation examination (PIE) of the Advanced Gas Reactor (AGR)-1 experiment was a multi-year, collaborative effort between Idaho National Laboratory (INL) and Oak Ridge National Laboratory (ORNL) to study the performance of UCO (uranium carbide, uranium oxide) tristructural isotropic (TRISO) coated particle fuel fabricated in the U.S. and irradiated at the Advanced Test Reactor at INL to a peak burnup of 19.6% fissions per initial metal atom. This work involved a broad array of experiments and analyses to evaluate the level of fission product retention by the fuel particles and compacts (both during irradiation and during post-irradiation heating tests to simulate reactor accident conditions), investigate the kernel and coating layer morphology evolution and the causes of coating failure, and explore the migration of fission products through the coating layers. The results have generally confirmed the excellent performance of the AGR-1 fuel, first indicated during the irradiation by the observation of zero TRISO coated particle failures out of 298,000 particles in the experiment. Overall release of fission products was determined by PIE to have been relatively low during the irradiation. A significant finding was the extremely low levels of cesium released through intact coatings. This was true both during the irradiation and during post-irradiation heating tests to temperatures as high as 1800°C. Post-irradiation safety test fuel performance was generally excellent. Silver release from the particles and compacts during irradiation was often very high. Extensive microanalysis of fuel particles was performed after irradiation and after high-temperature safety testing. The results of particle microanalysis indicate that the UCO fuel is effective at controlling the oxygen partial pressure within the particle and limiting kernel migration. Post-irradiation examination has provided the final body of data that speaks to the quality of the AGR-1 fuel, building

  12. AGR-1 Post Irradiation Examination Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Demkowicz, Paul Andrew [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-08-01

    The post-irradiation examination (PIE) of the Advanced Gas Reactor (AGR)-1 experiment was a multi-year, collaborative effort between Idaho National Laboratory (INL) and Oak Ridge National Laboratory (ORNL) to study the performance of UCO (uranium carbide, uranium oxide) tristructural isotropic (TRISO) coated particle fuel fabricated in the U.S. and irradiated at the Advanced Test Reactor at INL to a peak burnup of 19.6% fissions per initial metal atom. This work involved a broad array of experiments and analyses to evaluate the level of fission product retention by the fuel particles and compacts (both during irradiation and during post-irradiation heating tests to simulate reactor accident conditions), investigate the kernel and coating layer morphology evolution and the causes of coating failure, and explore the migration of fission products through the coating layers. The results have generally confirmed the excellent performance of the AGR-1 fuel, first indicated during the irradiation by the observation of zero TRISO coated particle failures out of 298,000 particles in the experiment. Overall release of fission products was determined by PIE to have been relatively low during the irradiation. A significant finding was the extremely low levels of cesium released through intact coatings. This was true both during the irradiation and during post-irradiation heating tests to temperatures as high as 1800°C. Post-irradiation safety test fuel performance was generally excellent. Silver release from the particles and compacts during irradiation was often very high. Extensive microanalysis of fuel particles was performed after irradiation and after high-temperature safety testing. The results of particle microanalysis indicate that the UCO fuel is effective at controlling the oxygen partial pressure within the particle and limiting kernel migration. Post-irradiation examination has provided the final body of data that speaks to the quality of the AGR-1 fuel, building

  13. Performance of AGR fuel - past, present and future

    International Nuclear Information System (INIS)

    The evolution of the Commercial Advanced Gas-cooled Reactor (CAGR) fuel element design, from trials in the Windscale AGR prototype to fuel currently being loaded, is briefly described. Fuel element performance has been monitored by an extensive post-irradiation examination programme, the results of which have been used to confirm or improve all aspects of the fuel design. As the CAGR system matures, there is less need for further design changes. Increases in discharge irradiation and improvements in manufacturing methods will further improve the economics of the AGR system. In order to achieve this, modest changes to the current successful fuel element design are being implemented or investigated. (Author)

  14. VGB water chemistry guideline for LWR type reactors

    International Nuclear Information System (INIS)

    The guideline for LWRs explains the quality standards to be met by the reactor feedwater and the primary water of BWR type reactors, and by the cooling water, steam generator feedwater and steam generator primary water of PWR type reactors. It also specifies quality standards for make-up water and steam used for the operation of turbines in LWR type power plant, which are subject to the same water purity requirements as fossil fueled power plant. The quality criteria are given as reference values, sometimes accompanied by values referring to specified normal operation, or limit values. The guideline applies to long-term operation, i.e. to the operating conditions at constant load. According to current knowledge and data, observation of the reference data given will exclude disturbance in the water or steam systems of reactors, steam generators and steam turbines. Certain operating conditions, such as load change or start-up and shut-down, may have effects on the water or steam quality which must not exceed the limit values. (orig./HP)

  15. 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)

  16. 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)

  17. 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.

  18. 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)

  19. AGR-1 Irradiation Test Final As-Run Report, Rev. 3

    International Nuclear Information System (INIS)

    This document presents the as-run analysis of the AGR-1 irradiation experiment. AGR-1 is the first of eight planned irradiations for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. Funding for this program is provided by the US Department of Energy (DOE) as part of the Next-Generation Nuclear Plant (NGNP) project. The objectives of the AGR-1 experiment are: 1. To gain experience with multi-capsule test train design, fabrication, and operation with the intent to reduce the probability of capsule or test train failure in subsequent irradiation tests. 2. To irradiate fuel produced in conjunction with the AGR fuel process development effort. 3. To provide data that will support the development of an understanding of the relationship between fuel fabrication processes, fuel product properties, and irradiation performance. In order to achieve the test objectives, the AGR-1 experiment was irradiated in the B-10 position of the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) for a total duration of 620 effective full power days of irradiation. Irradiation began on December 24, 2006 and ended on November 6, 2009 spanning 13 ATR cycles and approximately three calendar years. The test contained six independently controlled and monitored capsules. Each capsule contained 12 compacts of a single type, or variant, of the AGR coated fuel. No fuel particles failed during the AGR-1 irradiation. Final burnup values on a per compact basis ranged from 11.5 to 19.6 %FIMA, while fast fluence values ranged from 2.21 to 4.39 x 1025 n/m2 (E >0.18 MeV). We'll say something here about temperatures once thermal recalc is done. Thermocouples performed well, failing at a lower rate than expected. At the end of the irradiation, nine of the originally-planned 19 TCs were considered functional. Fission product release-to-birth (R/B) ratios were quite low. In most capsules, R/B values at the end of the irradiation were at or below 10-7 with only one

  20. AGR-1 Irradiation Test Final As-Run Report, Rev. 3

    Energy Technology Data Exchange (ETDEWEB)

    Collin, Blaise P.

    2015-01-01

    This document presents the as-run analysis of the AGR-1 irradiation experiment. AGR-1 is the first of eight planned irradiations for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. Funding for this program is provided by the US Department of Energy (DOE) as part of the Next-Generation Nuclear Plant (NGNP) project. The objectives of the AGR-1 experiment are: 1. To gain experience with multi-capsule test train design, fabrication, and operation with the intent to reduce the probability of capsule or test train failure in subsequent irradiation tests. 2. To irradiate fuel produced in conjunction with the AGR fuel process development effort. 3. To provide data that will support the development of an understanding of the relationship between fuel fabrication processes, fuel product properties, and irradiation performance. In order to achieve the test objectives, the AGR-1 experiment was irradiated in the B-10 position of the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) for a total duration of 620 effective full power days of irradiation. Irradiation began on December 24, 2006 and ended on November 6, 2009 spanning 13 ATR cycles and approximately three calendar years. The test contained six independently controlled and monitored capsules. Each capsule contained 12 compacts of a single type, or variant, of the AGR coated fuel. No fuel particles failed during the AGR-1 irradiation. Final burnup values on a per compact basis ranged from 11.5 to 19.6 %FIMA, while fast fluence values ranged from 2.21 to 4.39 x 1025 n/m2 (E >0.18 MeV). We’ll say something here about temperatures once thermal recalc is done. Thermocouples performed well, failing at a lower rate than expected. At the end of the irradiation, nine of the originally-planned 19 TCs were considered functional. Fission product release-to-birth (R/B) ratios were quite low. In most capsules, R/B values at the end of the irradiation were at or below

  1. Dimethylamine as a replacement for ammonia dosing in the secondary circuit of an advanced gas-cooled reactor (AGR) power station

    Energy Technology Data Exchange (ETDEWEB)

    Armstrong, Chris; Mitchell, Malcolm S. [EDF Energy, Hartlepool Power Station, Hartlepool (United Kingdom); Bull, Andrew E.A.; Quirk, Graham P.; Rudge, Andy [EDF Energy Nuclear Generation, Barnwood, Gloucester (United Kingdom). Central Technical Organisation; Woolsey, Ian S.

    2012-06-15

    Increasing flow resistance observed over recent years within the helical once-through boilers in the four advanced gas-cooled reactors at Hartlepool and Heysham 1 Power Stations have reduced boiler performance, resulting in reductions in feedwater flow, steam temperatures, and power output and in the need to carry out periodic chemical cleaning. The root cause is believed to be the development of magnetite deposits with high flow impedance in the 9Cr1Mo evaporator section of the boiler tubing. To prevent continued increases in boiler flow resistance, dimethylamine is being trialled, in one of the four affected units, as a replacement to the conventional ammonia dosing. Dimethylamine increases the pH at temperature around the secondary circuit and, based on full scale boiler rig simulations, is expected to reduce iron transport and prevent flow resistance increases within the evaporator section of the boiler. The dimethylamine plant trial commenced in January 2011 and is ongoing. The feedwater concentration of dimethylamine has been increased progressively towards a final target value of 900 {mu}g . kg{sup -1} and its effect on iron transport and boiler pressure loss is being closely monitored. The high steam temperature (> 500 C) of the secondary circuit leads to some decomposition of dimethylamine, which is being carefully monitored at various locations around the circuit. The decomposition products identified with dimethylamine dosing include ammonia, methylamine, formic acid, carbon dioxide and, as yet, unidentified neutral organic species. The effect of dimethylamine dosing on iron transport and boiler pressure drops and its decomposition behaviour around the secondary circuit during the plant trial will be presented in this paper. (orig.)

  2. Measurement of pressure pulsations in WWER-type reactors

    International Nuclear Information System (INIS)

    Measurements of pressure pulsations in WWER-type reactors are briefly described. A piezoelectric sensor and a charge sensitive amplifier forming the main measuring channel for pressure pulsation measurement are described. The charge sensitivity of the amplifier and its long-term drift are discussed. The freouency response and the design of the amplifier are given. The amplifier described was tested in laboratory; it represents the first stage in the development of the system for pressure pulsation measurements. (author)

  3. Results of the BREST-300 type reactor model fuel elements testing in the IGR reactor

    International Nuclear Information System (INIS)

    Testings of BREST-300 type fast reactor's model fuel elements with nitride fuel in the lead coolant in the central experimental channel of IGR reactor were carried out. In the testing the regime of non-controlled power burst was simulated. In the result of testing the seal failure of fuel elements with 2 % and 10 % 235U enrichment has been occurred, and fragmentation of the part of fuel pellets at interaction with coolant has been taken place. During the reactor testing the measurements and registration of experimental parameters (temperature of fuel, shell, coolant; pressure in fuel elements and testing ampoule; power release in the reactor) were conducted. The physical study of the 'fuel element - ampoule - reactor' was carried out, after-start-up spectrometric and material testing studies, calculated evaluation of temperature fields parameters in the testing ampoule were examined as well. Calculated and experimental values of breaking down specific power releases in the fuel are obtained. The assessment of both fuel fragmentation rate and it character is carried out. Distribution of fuel fragmentation within experimental ampoule volume is studied

  4. 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)

  5. Installation of a new type of nuclear reactor in Mexico: advantages and disadvantages

    International Nuclear Information System (INIS)

    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)

  6. Reactor building seismic analysis of a PWR type - NPP

    International Nuclear Information System (INIS)

    Earthquake engineering studies raised up in Brazil during design licensing and construction phases of Almirante Alvaro Alberto NPP, units 1 and 2. State of art of soil - structure interaction analysis with particular reference to the impedance function calculation analysis with particular reference to the impedance function calculation of a group of pile is presented in this M.Sc. Dissertation, as an example the reactor building dynamic response of a 1325 MWe NPP PWR type is calculated. The reactor building is supported by a pile foundation with 2002 end bearing piles. Upper and lower bound soil parameters are considered in order to observe their influence on dynamic response of structure. Dynamic response distribution on pile heads show pile-soil-pile interaction effects. (author)

  7. Verification tests performed for development of an integral type reactor

    International Nuclear Information System (INIS)

    SMART is an integral type reactor with innovative design features aimed at achieving a highly enhanced safety and improved economics. The SMART design is based on proven reactor design technologies with the use of new advanced design features. Most of the design features implemented into the SMART have been proven, however the advanced design features implemented into the SMART should be proven by testing. Various thermal hydraulic experiments have been carried out and also planned to assure the fundamental behavior of major concepts of the SMART and to prove the performance of the systems with new innovative technologies. This paper describes the thermal hydraulic test program for the SMART development and briefly discusses the typical test results. (author)

  8. Solid-Core Heat-Pipe Nuclear Batterly Type Reactor

    International Nuclear Information System (INIS)

    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). Like the SAFE 400 space nuclear reactor core, the HPENHS core is comprised of fuel rods and HPs embedded in a solid structure arranged in a hexagonal lattice in a 3:1 ratio. The core is oriented horizontally and has a square rather cylindrical cross section for effective heat transfer. The HPs extend from the two axial reflectors in which the fission gas plena are embedded and transfer heat to an intermediate coolant that flows by natural-circulation. The HP-ENHS is designed to preserve many features of the ENHS including 20-year operation without refueling, very small excess reactivity throughout life, natural circulation cooling, walkaway passive safety, and robust proliferation resistance. The target power level and specific power of the HP-ENHS reactor are those of the reference ENHS reactor. Compared to previous ENHS reactor designs utilizing a lead or lead-bismuth alloy natural circulation cooling system, the HP-ENHS reactor offers a number of advantageous features including: (1) significantly enhanced passive decay heat removal capability; (2) no positive void reactivity coefficients; (3) relatively lower corrosion of the cladding (4) a core that is more robust for transportation; (5) higher temperature potentially offering higher efficiency and hydrogen production capability. This preliminary study focuses on five areas: material compatibility analysis, HP performance analysis, neutronic analysis, thermal-hydraulic analysis and safety analysis. Of the four high-temperature structural materials evaluated, Mo TZM alloy is the preferred choice; its upper estimated feasible operating temperature is 1350 K. HP performance is evaluated as a function of working fluid type, operating temperature, wick design and HP diameter and length. Sodium is the

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

    Energy Technology Data Exchange (ETDEWEB)

    Jensen, S.E.; Nonboel, E

    1999-05-01

    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 (CO{sub 2}) 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)

  10. A CANDU-type small/medium power reactor

    International Nuclear Information System (INIS)

    This presentation reviews some of the main factors that will govern the design and operation of reactors in remote Northern Canadian communities, as applied to a small CANDU-type power plant. The central advantage of the CANDU is the fact that it is modular at the level of a single fuel channel. Examining each of the main features of this SMR plant on a hypothetical site in the Canadian Arctic reveals some of the unique characteristics that will be either desirable or mandatory for any such power plant applied to service in this remote region. (author)

  11. 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.)

  12. 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.

  13. AGR design for heysham II and Torness

    International Nuclear Information System (INIS)

    The background to the decision to build two AGR stations at Heysham and Torness is reviewed. The stations will each have twin 660MW(c) reactors based on the design of these at Hinkley Point B and Hunterston B, but including design changes introduced to meet enhanced safety requirements. The main features of the new design are described. Changes include: the provision of greater diversity in the shut-down and post-trip cooling systems; increased segregation of post-trip cooling and protection systems against fires and other hazards; measures for withstanding a major earthquake and extreme winds; greater provision for inservice inspection. The station layout is illustrated. Some generating costs are shown in comparison with a PWR and a coal-fired unit. (U.K.)

  14. RUTA pool-type reactor for heat supply and the possibility for its application area expansion

    International Nuclear Information System (INIS)

    RUTA, a reactor facility with a pool-type reactor, has been designed for heat supply of residential districts. A relatively low potential of the heat generated by the reactor requires a special approach to building up heat supply systems with RUTA facilities. The application of the RUTA facility as a heat source for seawater thermal distillation has been considered. It is possible to use the reactor for neutron therapy. The reactor optimization provides for the improvement of the facility's consumer qualities. (author)

  15. Transmutation of plutonium in pebble bed type high temperature reactors

    International Nuclear Information System (INIS)

    The pebble bed type High Temperature Reactor (HTR) has been studied as a uranium-free burner of reactor grade plutonium. In a parametric study, the plutonium loading per pebble as well as the type and size of the coated particles (CPs) have been varied to determine the plutonium consumption, the final plutonium burnup, the k∞ and the temperature coefficients as a function of burnup. The plutonium loading per pebble is bounded between 1 and 3 gr Pu per pebble. The upper limit is imposed by the maximal allowable fast fluence for the CPs. A higher plutonium loading requires a longer irradiation time to reach a desired burnup, so that the CPs are exposed to a higher fast fluence. The lower limit is determined by the temperature coefficients, which become less negative with increasing moderator-actinide ratio. A burnup of about 600 MWd/kgHM can be reached. With the HTR's high efficiency of 40%, a plutonium supply of 1520 kg/GWea is achieved. The discharges of plutonium and minor actinides are then 450 and 110 kg/GWea, respectively. (author)

  16. Overall plant concept for a tank-type fast reactor

    International Nuclear Information System (INIS)

    Japanese nuclear industries are expressing interest in the merits of the tank-type FBR as a large plant (demonstration) after JOYO (experimental, in operation) and MONJU (prototype, under construction). In response to this growing interest in a tank-type FBR demonstration plant, Hitachi has initiated a conceptual study of a 1000 MWe tank plant concept in collaboration with GE and Bechtel. Key objectives of this study have been: to select reliable and competitive tank plant concepts, with emphases on a seismic-resistant and compact tank reactor system;to select reliable shutdown heat removal system;and to identify R and D items needed for early 1990s construction. Design goals were defined as follows: capital costs must be less than twice, and as close as practical to 1.5 those of equivalent LWR plants;earthquake resistant structures to meet stringent Japanese seismic conditions must be as simple and reliable as practical;safety must be maintained at LWR-equivalent risks;and R and D needs must be limited to minimum cost for the limited time allowed. This paper summarizes the overall plant concepts with some selected topics, whereas detailed descriptions of the reactor assembly and the layout design are found in separate papers

  17. The water desalination complex based on ABV-type reactor plant

    International Nuclear Information System (INIS)

    A floating nuclear desalination complex with two barges, one for ABV type reactor plant, with twin reactor 2 x 6 MW(e), and one for reverse osmosis desalination plant, was described. The principal specifications of the ABV type reactor plant and desalination barge were given. The ABV type reactor has a traditional two-circuit layout using an integral type reactor vessel with all mode natural convection of primary coolant. The desalted water cost was estimated to be around US $0.86 per cubic meter. R and D work has been performed and preparations for commercial production are under way. (author)

  18. Liquid-metal-gas heat exchanger for HTGR type reactors

    International Nuclear Information System (INIS)

    The aim of this study is to investigate the heat transfer characteristics of a liquid metal heat exchanger (HE) for a helium-cooled high temperature reactor. A tube-type heat exchanger is considered as well as two direct exchangers: a bubble-type heat exchanger and a heat exchanger according to the spray principle. Experiments are made in order to determine the gas content of bubble-type heat exchangers, the dependence of the droplet diameter on the nozzle diameter, the falling speed of the droplets, the velocity of the liquid jet, and the temperature variation of liquid jets. The computer codes developed for HE calculation are structured so that they may be used for gas/liquid HE, too. Each type of HE that is dealt with is designed by accousting for a technical and an economic assessment. The liquid-lead jet spray is preferred to all other types because of its small space occupied and its simple design. It shall be used in near future in the HTR by the name of lead/helium HE. (GL)

  19. Effects of nuclear island connected buildings on seismic behaviour of reactor internals in a pool type fast breeder reactor

    International Nuclear Information System (INIS)

    The seismic analysis of reactor assembly housing the primary circuit of a typical 500 MWe capacity pool type fast breeder reactor (PFBR) is reported. The reactor assembly is supported on the reactor vault within the nuclear island connected buildings (NICB). The seismic responses, viz. critical displacements, sloshing heights, stresses and strain energy values in the vessels are determined for the reactor assembly by detailed finite element analysis including the fluid-structure interaction and sloshing effects. Analysis is carried out to quantify the effects of inter-connection of the reactor vault with the adjacent buildings under the assumptions that the reactor vault along with reactor assembly is: (1) an isolated structural system from the adjacent buildings within reactor containment building (RCB) and (2) connected with the adjacent civil structures through floor slabs. Analysis indicates that, by inter-connecting the vault with the NICB, there are overall increases of all the governing parameters which decide the seismic design criteria. The significant effects are increases of: (1) radial and axial displacements of core top and absorber rods and vertical accelerations of core subassemblies which are of concern to reactor safety, (2) primary membrane stress intensities for the inner vessel and (3) strain energies developed at the critical portions which can enhance the buckling risks of main vessel, inner vessel and thermal baffles. Hence, it is preferable to isolate the reactor vault, directly constructing from the base raft without inter-connecting it with the NICB, from the seismic loading considerations

  20. Advances in AGR fuel fabrication - now and the future

    International Nuclear Information System (INIS)

    To date, over 3 million AGR fuel pins have been manufactured at Springfields for the UK AGR programme. During this time, AGR fuel design and manufacture has developed and evolved in response to the needs of the reactor operators to enhance fuel reliability and performance. More recently, major advances have been made in the systems and organisational culture which support fuel manufacture at Fuel Division. The introduction of MRP II in 1989 into Fuel Division enabled significant reductions in stock and work-in-progress, together with reductions in manufacturing lead times. Other successful initiatives introduced into Fuel Division have been Just-in-Time (JIT) and AST (Additional Skills Training) which have built on the success of MRP II. All of these initiatives are evidence of Fuel Division's ''Total Quality'' approach to fabricating fuel. Fuel Division is currently in the final stages of commissioning the New Oxide Fuels Complex (NOFC) where both AGR and PWR fuel will be manufactured to the highest standards of quality, safety and environmental protection. NOFC is a totally integrated plant which represents a Pound 200M investment, demonstrating Fuel Division's commitment to building on its 40+ years of fuel fabrication experience and ensuring secure supply of fuel to its customers for years to come. (author)

  1. Utilization of the experimental reactor Osiris for the study and the development of fuels of the fast neutron reactor type

    International Nuclear Information System (INIS)

    Nuclear fuel tests for the fast neutron reactor type have been carried out at the Osiris reactor: thermal study of (U,Pu)O2 oxide by measurement with thermocouples in the core of the fuel pellet; study of the effects of power cycling on nuclear fuel; study of the mechanical interactions between oxide and cladding by measurement of the cladding deformation during irradiation

  2. Investigation of accident management strategies for VVER-1000-Type reactors

    International Nuclear Information System (INIS)

    The goal of this work is the search for an optimal accident management strategy to prevent containment failure and to stop the core/concrete interaction from hindering cavity bottom melt-through on the one hand and from ending the ex-vessel source term increase on the other hand, i.e., to terminate the accident. The work is based on the results of previous studies of physical and chemical phenomena during different accident scenarios for VVER-1000-type reactors. For a TMLB' sequence (an accident caused by a transient in which core melt occurs because the electric power cannot be restored before the pressure vessel melts through), a number of calculations were performed using the source term code package (STCP) to investigate the influence of several accident management measures on the core/concrete interaction and the containment integrity

  3. Inteligent control system for a CANDU 600 type reactor process

    International Nuclear Information System (INIS)

    The present paper is set on presenting a highly intelligent configuration, capable of controlling, without the need of the human factor, a complete nuclear power plant type of system, giving it the status of an autonomous system. The urge for such a controlling system is justified by the amount of drawbacks that appear in real life as disadvantages, loses and sometimes even inefficiency in the current controlling and comanding systems of the nuclear reactors. The application stands in the comand sent from the auxiliary feedwater flow control valves to the steam generators. As an environment fit for development I chose Matlab Simulink to simulate the behaviour of the process and the adjusted system. Comparing the results obtained after the fuzzy regulation with those obtained after the classical regulation, we can demonstrate the necessity of implementing artificial intelligence techniques in nuclear power plants and we can agree to the advantages of being able to control everything automatically. (authors)

  4. Radionuclide compositions of spent fuel and high level waste from commercial nuclear reactors

    International Nuclear Information System (INIS)

    This report provides information on radionuclide compositions of spent fuel and high level waste produced during reprocessing. The reactor types considered are Magnox, AGR, PWR and CFR. The activities of the radionuclides are calculated using the FISPIN code. The results are presented in a form suitable for radioactive waste management calculations. (author)

  5. AGR-2 IRRADIATION TEST FINAL AS-RUN REPORT

    Energy Technology Data Exchange (ETDEWEB)

    Blaise, Collin

    2014-07-01

    This document presents the as-run analysis of the AGR-2 irradiation experiment. AGR-2 is the second of the planned irradiations for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. Funding for this program is provided by the U.S. Department of Energy as part of the Very High Temperature Reactor (VHTR) Technical Development Office (TDO) program. The objectives of the AGR-2 experiment are to: (a) Irradiate UCO (uranium oxycarbide) and UO2 (uranium dioxide) fuel produced in a large coater. Fuel attributes are based on results obtained from the AGR-1 test and other project activities. (b) Provide irradiated fuel samples for post-irradiation experiment (PIE) and safety testing. (c) Support the development of an understanding of the relationship between fuel fabrication processes, fuel product properties, and irradiation performance. The primary objective of the test was to irradiate both UCO and UO2 TRISO (tri-structural isotropic) fuel produced from prototypic scale equipment to obtain normal operation and accident condition fuel performance data. The UCO compacts were subjected to a range of burnups and temperatures typical of anticipated prismatic reactor service conditions in three capsules. The test train also includes compacts containing UO2 particles produced independently by the United States, South Africa, and France in three separate capsules. The range of burnups and temperatures in these capsules were typical of anticipated pebble bed reactor service conditions. The results discussed in this report pertain only to U.S. produced fuel. In order to achieve the test objectives, the AGR-2 experiment was irradiated in the B-12 position of the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) for a total irradiation duration of 559.2 effective full power days (EFPD). Irradiation began on June 22, 2010, and ended on October 16, 2013, spanning 12 ATR power cycles and approximately three and a half calendar years. The test

  6. AGR-2 Safety Test Predictions Using the PARFUME Code

    Energy Technology Data Exchange (ETDEWEB)

    Blaise Collin

    2014-09-01

    This report documents calculations performed to predict failure probability of TRISO-coated fuel particles and diffusion of fission products through these particles during safety tests following the second irradiation test of the Advanced Gas Reactor program (AGR-2). The calculations include the modeling of the AGR-2 irradiation that occurred from June 2010 to October 2013 in the Advanced Test Reactor (ATR) and the modeling of a safety testing phase to support safety tests planned at Oak Ridge National Laboratory and at Idaho National Laboratory (INL) for a selection of AGR-2 compacts. The heat-up of AGR-2 compacts is a critical component of the AGR-2 fuel performance evaluation, and its objectives are to identify the effect of accident test temperature, burnup, and irradiation temperature on the performance of the fuel at elevated temperature. Safety testing of compacts will be followed by detailed examinations of the fuel particles to further evaluate fission product retention and behavior of the kernel and coatings. The modeling was performed using the particle fuel model computer code PARFUME developed at INL. PARFUME is an advanced gas-cooled reactor fuel performance modeling and analysis code (Miller 2009). It has been developed as an integrated mechanistic code that evaluates the thermal, mechanical, and physico-chemical behavior of fuel particles during irradiation to determine the failure probability of a population of fuel particles given the particle-to-particle statistical variations in physical dimensions and material properties that arise from the fuel fabrication process, accounting for all viable mechanisms that can lead to particle failure. The code also determines the diffusion of fission products from the fuel through the particle coating layers, and through the fuel matrix to the coolant boundary. The subsequent release of fission products is calculated at the compact level (release of fission products from the compact). PARFUME calculates the

  7. Advances in safety engineering for LWR type reactors (retrofitting, re-engineering, EPR, SWR 1000)

    International Nuclear Information System (INIS)

    The paper summarizes the activities of the Siemens company in the field of advanced LWR type reactor engineering and the company's commitments in international projects for the retrofitting and engineered safety improvements of reactor stations in countries of the former Soviet Union. The advances in reactor engineering and the novel design concepts are explained. (orig./CB)

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

    International Nuclear Information System (INIS)

    Most of the research reactors discussed in this document are pool-type reactors in which the reactor vessel and some of the reactor coolant systems are located in a pool of water. These reactors generally use fuel in plate assemblies formed by a compact layer of uranium (or U3Si2) and aluminium particles, sandwiched between two thin layers of aluminium serving as cladding. The fuel melting process begins at 660 deg. C when the aluminium melts, while the uranium (or U3Si2) particles may remain solid. The accident that occurred in the American SL-1 reactor in 1961, together with tests carried out in the United States as of 1954 in the BORAX-1 reactor and then, in 1962, in the SPERT-1 reactor, showed that a sudden substantial addition of reactivity in this type of reactor could lead to explosive mechanisms caused by degradation, or even fast meltdown, of part of the reactor core. This is what is known as a 'BORAX-type' accident. The aim of this document is first to briefly recall the circumstances of the SL-1 reactor accident, the lessons learned, how this operational feedback has been factored into the design of various research reactors around the world and, second, to describe the approach taken by France with regard to this type of accident and how, led by IRSN, this approach has evolved in the last decade. (authors)

  9. Design and development of steam generators for the AGR power stations at Heysham II/Torness

    International Nuclear Information System (INIS)

    The current AGR steam generator design is a development of the successful once-through units supplied for the Oldbury Magnox and Hinkley/Hunterston AGR power stations. These units have demonstrated proven control and reliability in service. In this paper the factors which have dictated the design and layout of the latest AGR steam generators are described and reference made to the latest high temperature design techniques that have been employed. Details of development work to support the design and establish the performance characteristics over the range of plant operating conditions are also given. To comply with current UK safety standards, the AGR steam generators and associated plant are designed to accommodate seismic loadings. In addition, provision is made for an independent heat removal system for post reactor trip operations. (author)

  10. 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

  11. Safety Approach of BORAX Type Accidents in French Research Reactors

    International Nuclear Information System (INIS)

    Most of pool type French research reactors are designed to withstand an explosive BORAX accident, defined as a pressure load on the pool walls. The purpose of this paper is to present the approach implemented at IRSN to analyse this accident by linking safety assessment and supporting studies. Examples of recent work on Jules Horowitz Reactor (JHR) and ORPHEE will be presented. Although all aspects of the accident are addressed, we will focus on the first two frames of the transient: the reactivity insertion and the consequences on the core. The first step of the BORAX analysis is to identify the most penalizing plausible reactivity insertion. This means characterising the sequences of events that can induce a reactivity surge and evaluate the worth of such variation. Neutronic computations are then required to quantify the reactivity increase. To comply with the geometrical specificities of research reactors, IRSN chose to use the homemade Monte Carlo code MORET5. The control rod worth calculations on the JHR were in good agreement with the operator results, whereas in ORPHEE, IRSN demonstrated that the beam channels reactivity worth was largely. In both cases the obtained results allowed an interesting dialogue with the operator and were used in the conclusions of the safety assessment. Following the accidental sequence of events, the second stage analysed by IRSN is the power transient occurring in the core and the consequences on the fuel. IRSN applied on JHR a homemade simplified model based on point kinetics and standard thermal balance equations to compute power evolution taking into account the temperatures of the fuel for feedback reactivity. As heat exchange coefficients between cladding and water for such fast transients are unknown, IRSN took the conservative hypothesis of adiabatic heating of the plates. The comparison the JHR power pulse calculation results against SPERT experimental measurements enabled IRSN to be optimistic about the possibility

  12. 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)

  13. Automatic pressure reduction device for BWR type reactor

    International Nuclear Information System (INIS)

    Purpose: To suppress the leakage amount from main steam isolation valves after the closure of the valves upon main steam pipeway rupture at the outside of a reactor container. Constitution: Main steam isolation valves disposed on both sides of the main steam pipeway penetration portion in a reactor container are closed, as well as a safety relief valve of the main steam pipeways disposed in the reactor container is opened by a rupture signal generated by the rupture of the main steam pipeways at the outside of the reactor container. Since the pressure in the reactor pressure vessel is automatically reduced after a predetermined of time even if rupture is caused to the main steam pipeways at the outside of the reactor pressure vessel, it is possible to suppress the leaking amount sufficiently after the entire closure of the main steam isolation valves and contribute to the improvement for the nuclear power plant safety. (Yoshihara, H.)

  14. Corrosion and hydridation features of RBMK type reactor technological channels

    International Nuclear Information System (INIS)

    Generalization results, obtained in the course of monitoring the corrosion state and hydridation of RBMK-1000 and RBMK-1500 reactor technological channels (TC) are presented. It is shown, that the corrosion behaviour of TC tube metal in reactors differs notably. Comparison of data on hybridization of RBMK-100 and RBMK-1500 reactor technological tubes allows one to suppose a possibly higher tendency to hydrogen absorption in Zr - 2.5% of Nb alloy under TMT-1 and TMT-2 states

  15. 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)

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

    International Nuclear Information System (INIS)

    The measurements carried out on reactor dosimetry are applied mainly at the study on the effects of the radiation in the materials of the reactor; a 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 tracks manufactured in the ININ is presented, for the environmental monitoring in penetrations around the primary container of the Unit 1 of the Laguna Verde power plant. The monitoring of neutrons carried out with ends of radiological protection, during those operational tests of the reactor. (author)

  17. 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)

  18. A conceptual design study on various types of HLMC fast reactor plant

    International Nuclear Information System (INIS)

    To seek for a promising concept of a heavy liquid metal coolant (HLMC) fast reactor plant, Japan Nuclear Cycle Development Institute (JNC) and the electric utilities conducted conceptual design study on various types of plant concepts and compared these concepts based on technical feasibility and economical perspective. Finally, Pb-Bi cooled medium tank type reactor was selected as a most promising concept. (author)

  19. Conceptual design studies on various types of HLMC fast reactor plants

    International Nuclear Information System (INIS)

    To seek for a promising concept of a heavy liquid metal coolant (HLMC) fast reactor plant, Japan Nuclear Cycle Development Institute (JNC) and the electric utilities conducted conceptual design studies on various types of plant concepts and compared these concepts based on technical feasibility and economical perspective. Finally, Pb-Bi cooled medium tank type reactor was selected as the most promising concept. (author)

  20. 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.)

  1. After heat removing device for FBR type reactor

    International Nuclear Information System (INIS)

    An annular liquid (water) pool is formed radially surrounding a reactor container and a reactor safety container. An annular cavity wall is formed in the liquid pool, and the inside of the cavity wall is formed as a liquid channel. If the temperature of liquid sodium in the reactor container rises by the after heat of the nuclear fuels, the temperature of the reactor safety container also rises to a high temperature, and the amount of heat radiated from the surface is increased. Water in the liquid channel heated by undergoing the radiation heat forms upward streams in the liquid channel by an air lift-effect caused by rising of boiling air bubbles. Namely, the water in the liquid pool rises the liquid channel while boiling to cool the reactor safety container. With such a constitution, after heat can be removed continuously by the spontaneously circulating water. (I.N.)

  2. DETERMINATION OF THE AGR-1 CAPSULE TO FPMS SPECTROMETER TRANSPORT VOLUMES FROM LEADOUT FLOW TEST DATA

    International Nuclear Information System (INIS)

    The AGR-1 experiment is a fueled multiple-capsule irradiation experiment being conducted in the Advanced Test Reactor (ATR) in support of the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. A flow experiment conducted during the AGR-1 irradiation provided data that included the effect of flow rate changes on the decay of a short-lived radionuclide (23Ne). This data has been analyzed to determine the capsule-specific downstream transport volume through which the capsule effluents must pass before arrival at the fission product monitoring system spectrometers. These resultant transport volumes when coupled with capsule outlet flow rates determine the transport times from capsule-to-detector. In this work an analysis protocol is developed and applied in order to determine capsule-specific transport volumes to precisions of better than +/- 7%

  3. 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.

  4. 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)

  5. 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).

  6. 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).

  7. 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

  8. 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).

  9. Nonlinear punctual dynamic applied to simulation of PWR type reactors

    International Nuclear Information System (INIS)

    In order to study some kinds of nuclear reactor accidents, a simulation is made using the punctual kinetics model to the reactor core. The following integration methods are used: Hansen's method in which a linearization is made and C S M P using a variable interval fourth-order Runge Kutta method. The results were good and were compared with those obtained by the code Dinamica I which uses a finite difference integration method of backward kind. (author)

  10. 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.

  11. 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

  12. AGR 3/4 Irradiation Test Final As Run Report

    Energy Technology Data Exchange (ETDEWEB)

    Collin, Blaise P. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-06-01

    Several fuel and material irradiation experiments have been planned for the Idaho National Laboratory Advanced Reactor Technologies Technology Development Office Advanced Gas Reactor Fuel Development and Qualification Program (referred to as the INL ART TDO/AGR fuel program hereafter), which supports the development and qualification of tristructural-isotropic (TRISO) coated particle fuel for use in HTGRs. The goals of these experiments are to provide irradiation performance data to support fuel process development, qualify fuel for normal operating conditions, support development and validation of fuel performance and fission product transport models and codes, and provide irradiated fuel and materials for post irradiation examination and safety testing (INL 05/2015). AGR-3/4 combined the third and fourth in this series of planned experiments to test TRISO coated low enriched uranium (LEU) oxycarbide fuel. This combined experiment was intended to support the refinement of fission product transport models and to assess the effects of sweep gas impurities on fuel performance and fission product transport by irradiating designed-to-fail fuel particles and by measuring subsequent fission metal transport in fuel-compact matrix material and fuel-element graphite. The AGR 3/4 fuel test was successful in irradiating the fuel compacts to the burnup and fast fluence target ranges, considering the experiment was terminated short of its initial 400 EFPD target (Collin 2015). Out of the 48 AGR-3/4 compacts, 42 achieved the specified burnup of at least 6% fissions per initial heavy-metal atom (FIMA). Three capsules had a maximum fuel compact average burnup < 10% FIMA, one more than originally specified, and the maximum fuel compact average burnup was <19% FIMA for the remaining capsules, as specified. Fast neutron fluence fell in the expected range of 1.0 to 5.5×1025 n/m2 (E >0.18 MeV) for all compacts. In addition, the AGR-3/4 experiment was globally successful in keeping the

  13. For sale: 7 AGR stations and a brand new PWR

    International Nuclear Information System (INIS)

    Britain's seven AGR stations and the Sizewell B PWR will pass to private ownership under the UK government's plan to privatise the two nuclear generators, Nuclear Electric and Scottish Nuclear, sometime next year. Under the new set-up, the two generators will become operating subsidiaries of a holding company which will be headquartered in Scotland. The companies' ageing Magnox gas-cooled reactors will remain in a separate public sector company before being transferred to British Nuclear Fuels (BNFL) at the time of privatisation. (author)

  14. PIE on Safety-Tested AGR-1 Compact 5-1-1

    Energy Technology Data Exchange (ETDEWEB)

    Hunn, John D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Morris, Robert Noel [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Baldwin, Charles A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Montgomery, Fred C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gerczak, Tyler J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-08-01

    Post-irradiation examination (PIE) is being performed in support of tristructural isotropic (TRISO) coated particle fuel development and qualification for High-Temperature Gas-cooled Reactors (HTGRs). AGR-1 was the first in a series of TRISO fuel irradiation experiments initiated in 2006 under the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program; this work continues to be funded by the Department of Energy's Office of Nuclear Energy as part of the Advanced Reactor Technologies (ART) initiative. AGR-1 fuel compacts were fabricated at Oak Ridge National Laboratory (ORNL) in 2006 and irradiated for three years in the Idaho National Laboratory (INL) Advanced Test Reactor (ATR) to demonstrate and evaluate fuel performance under HTGR irradiation conditions. PIE is being performed at INL and ORNL to study how the fuel behaved during irradiation, and to examine fuel performance during exposure to elevated temperatures at or above temperatures that could occur during a depressurized conduction cooldown event. This report summarizes safety testing of irradiated AGR-1 Compact 5-1-1 in the ORNL Core Conduction Cooldown Test Facility (CCCTF) and post-safety testing PIE.

  15. Uncertainty Quantification of Calculated Temperatures for the AGR 3/4 Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Pham, Binh Thi-Cam [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-09-01

    A series of Advanced Gas Reactor (AGR) irradiation experiments are being conducted within the Advanced Reactor Technology (ART) Fuel Development and Qualification Program. The main objectives of the fuel experimental campaign are to provide the necessary data on fuel performance to support fuel process development, qualify a fuel design and fabrication process for normal operation and accident conditions, and support development and validation of fuel performance and fission product transport models and codes (PLN 3636, “Technical Program Plan for INL Advanced Reactor Technologies Technology Development Office/Advanced Gas Reactor Fuel Development and Qualification Program”). The AGR 3/4 test was inserted in the Northeast Flux Trap position in the Advanced Test Reactor (ATR) core at Idaho National Laboratory (INL) in December 2011 and successfully completed irradiation in mid-April 2014, resulting in irradiation of the tristructural isotropic (TRISO) fuel for 369.1 effective full-power days (EFPDs) during approximately 2.4 calendar years. The AGR 3/4 data, including the irradiation data and calculated results, were qualified and stored in the Nuclear Data Management and Analysis System (NDMAS). To support the U.S. TRISO fuel performance assessment and to provide data for validation of fuel performance and fission product transport models and codes, the daily as run thermal analysis has been performed separately on each of twelve AGR 3/4 capsules for the entire irradiation as discussed in ECAR-2807, “AGR 3/4 Daily As Run Thermal Analyses”. The ABAQUS code’s finite element-based thermal model predicts the daily average volume average (VA) fuel temperature (FT), peak FT, and graphite matrix, sleeve, and sink temperature in each capsule. The JMOCUP simulation codes were also created to perform depletion calculations for the AGR 3/4 experiment (ECAR-2753, “JMOCUP As-Run Daily Physics Depletion Calculation for the AGR 3/4 TRISO Particle Experiment in ATR

  16. Uncertainty Quantification of Calculated Temperatures for the AGR 3/4 Experiment

    International Nuclear Information System (INIS)

    A series of Advanced Gas Reactor (AGR) irradiation experiments are being conducted within the Advanced Reactor Technology (ART) Fuel Development and Qualification Program. The main objectives of the fuel experimental campaign are to provide the necessary data on fuel performance to support fuel process development, qualify a fuel design and fabrication process for normal operation and accident conditions, and support development and validation of fuel performance and fission product transport models and codes (PLN 3636, 'Technical Program Plan for INL Advanced Reactor Technologies Technology Development Office/Advanced Gas Reactor Fuel Development and Qualification Program'). The AGR 3/4 test was inserted in the Northeast Flux Trap position in the Advanced Test Reactor (ATR) core at Idaho National Laboratory (INL) in December 2011 and successfully completed irradiation in mid-April 2014, resulting in irradiation of the tristructural isotropic (TRISO) fuel for 369.1 effective full-power days (EFPDs) during approximately 2.4 calendar years. The AGR 3/4 data, including the irradiation data and calculated results, were qualified and stored in the Nuclear Data Management and Analysis System (NDMAS). To support the U.S. TRISO fuel performance assessment and to provide data for validation of fuel performance and fission product transport models and codes, the daily as run thermal analysis has been performed separately on each of twelve AGR 3/4 capsules for the entire irradiation as discussed in ECAR-2807, 'AGR 3/4 Daily As Run Thermal Analyses'. The ABAQUS code's finite element-based thermal model predicts the daily average volume average (VA) fuel temperature (FT), peak FT, and graphite matrix, sleeve, and sink temperature in each capsule. The JMOCUP simulation codes were also created to perform depletion calculations for the AGR 3/4 experiment (ECAR-2753, 'JMOCUP As-Run Daily Physics Depletion Calculation for the AGR 3/4 TRISO Particle

  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

    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.

  19. 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.

  20. Convective cooling in a pool-type research reactor

    International Nuclear Information System (INIS)

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

  1. After-heat removal system in BWR type reactor

    International Nuclear Information System (INIS)

    An after-heat removal system having a duel low pressure coolant injection mode (LPCI) selects an integral recycling loop based on a pressure difference between reactor recycling loops to inject emergency cooling water to the reactor. In this case, if the pressure difference between the recycling loops is less than such a pressure difference as capable of injecting a sufficient amount of cooling water to the reactor core, injection lines to both of the recycling loops are lined up. With such a constitution, the injection lines of LPCI can be retarded in most of the cases of requiring LPCI, to remarkably improve the reliability and sufficiently utilize the retardation of the four pumps. (I.S.)

  2. 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)

  3. 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.

  4. Mathematical game type optimization of powerful fast reactors

    International Nuclear Information System (INIS)

    To obtain maximum speed of putting into operation fast breeders it is recommended on the initial stage of putting into operation these reactors to apply lower power which needs less fission materials. That is why there is an attempt to find a configuration of a high-power reactor providing maximum power for minimum mass of fission material. This problem has a structure of the mathematical game with two partners of non-zero-order total and is solved by means of specific aids of theory of games. Optimal distribution of fission and breeding materials in a multizone reactor first is determined by solution of competitive game and then, on its base, by solution of the cooperation game. The second problem the solution for which is searched is developed from remark on the fact that a reactor with minimum coefficient of flux heterogenity has a configuration different from the reactor with power coefficient heterogenity. Maximum burn-up of fuel needs minimum heterogenity of the flux coefficient and the highest power level needs minimum coefficient of power heterogenity. That is why it is possible to put a problem of finding of the reactor configuration having both coefficients with minimum value. This problem has a structure of a mathematical game with two partners of non-zero-order total and is solved analogously giving optimal distribution of fuel from the new point of view. In the report is shown that both these solutions are independent which is a result of the aim put in the problem of optimization. (author)

  5. Mechanical, chemical and radiological characterization of the graphite of the UNGG reactors type

    International Nuclear Information System (INIS)

    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.)

  6. The development of the physical conceptions of the FBR type reactors control methods

    International Nuclear Information System (INIS)

    The physical concepts and specific problems of the control elements for LMFBR type reactors are discussed in this paper. Typical temperature coefficient of reactivity, its dependency on reactor power and burnup level are given. The authors give us the most advisable methods of the reactivity coefficient compensation

  7. Supercritical CO2 direct cycle Gas Fast Reactor (SC-GFR) concept.

    Energy Technology Data Exchange (ETDEWEB)

    Wright, Steven Alan; Parma, Edward J., Jr.; Suo-Anttila, Ahti Jorma (Computational Engineering Analysis, Albuquerque, NM); Al Rashdan, Ahmad (Texas A& M University, College Station, TX); Tsvetkov, Pavel Valeryevich (Texas A& M University, College Station, TX); Vernon, Milton E.; Fleming, Darryn D.; Rochau, Gary Eugene

    2011-05-01

    This report describes the supercritical carbon dioxide (S-CO{sub 2}) direct cycle gas fast reactor (SC-GFR) concept. The SC-GFR reactor concept was developed to determine the feasibility of a right size reactor (RSR) type concept using S-CO{sub 2} as the working fluid in a direct cycle fast reactor. Scoping analyses were performed for a 200 to 400 MWth reactor and an S-CO{sub 2} Brayton cycle. Although a significant amount of work is still required, this type of reactor concept maintains some potentially significant advantages over ideal gas-cooled systems and liquid metal-cooled systems. The analyses presented in this report show that a relatively small long-life reactor core could be developed that maintains decay heat removal by natural circulation. The concept is based largely on the Advanced Gas Reactor (AGR) commercial power plants operated in the United Kingdom and other GFR concepts.

  8. Study on regimes of nuclear power plants with WWER-type reactors

    International Nuclear Information System (INIS)

    The problems are considered of optimization of nuclear fuel loading, the peculiarities of the NPP operation at decreased power, and also the problem of stability operation of NPP with WWER type reactors taking into account specific features of these reactors (partial fuel overloads, change in reactor reactivity with power changes). The two particular interconnected problems discussed are: choice of such a sequence of partial rechargings which ensures the minimum cost of the electric power generated, and increasing the reactor operating time by reducing its power output. Besides the technical and economic estimates, much attention is given to analysing the stability of NPP operation

  9. The conceptual design of IPR1000 reactor pressure vessel for PWR type

    International Nuclear Information System (INIS)

    The conceptual design of IPR1000 reactor pressure vessel for PWR type has been configured, material selection and dimension parameter are designed to cover the reactor cooling system (RCS), nuclear fuel assembly, and others internals reactor. The reactor pressure vessel consist of closure head assembly, vessel shell upper head assembly, vessel shell lower assembly, and inlet and outlet nozzle. These are designed capable to support weight of RPV, at pressures and temperature of each 2485 psig and 650 °F. The design refers to AP1000 as according to ASME code and industrials standard applicable for Nuclear Power Plants (NPP). (author)

  10. Spent fuel management in Czechoslovak WWER-440 type reactors

    International Nuclear Information System (INIS)

    The main aspects of the present WWER-440 reactors spent fuel management are described in the paper. Experimental results of fuel integrity studies which are carried out under conditions of a long-term storage are also presented. (author). 5 refs, 5 figs

  11. Plutonium utilization in different reactor types in France

    International Nuclear Information System (INIS)

    Many liquid-metal fast breeder reactor (LMFBR) penetration studies were carried out in France after the 1973 oil crisis. Since that period, several new facts have modified the possible scenarios. The energy (especially electricity) use increase was lower than anticipated. Consequently, uranium supplying difficulties and uranium cost increases will not occur until later. As a result, pressurized water reactor (PWR) plutonium recycling and studies of advanced pressurized water reactors (APWRs) putting plutonium to use will begin. In France many PWRs were built between 1973 and 1989. After the first lifetime studies, a 35- or 40-yr lifetime can be expected. Then after 2010 the first reactors must be changed and many new power plants must be started. At this time PWRs will have produced a great amount of plutonium. How will this plutonium be put to use? As regards the use of fissile materials, LMFBRs are the most efficient. But LMFBRs are more expensive than PWRs, and as a first step, uses of plutonium such as recycling in PWRs or APWRs can be considered. In this paper the authors consider PWRs, APWRs, and LMFBRs, from a physicist's point of view, which focuses on the problems of in-core and out-of-core plutonium isotopic composition evolution and plutonium supplying safeguards. In these studies, all the plutonium produced is used as soon as possible, not taking into consideration the economic aspects of plutonium utilization

  12. 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

  13. 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.)

  14. 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)

  15. Public acceptance of fusion energy and scientific feasibility of a fusion reactor. Design of helical-type fusion reactors

    International Nuclear Information System (INIS)

    Currently, there is active interest in the research and development of helical systems. New large devices using superconducting magnets (LHD in Japan and W7-X in Germany) are expected to produce highly improved plasmas comparable to those recently obtained in the large tokamaks. Because current-less steady operation is advantageous, these aggressive programs have accelerated several design studies of helical-type reactors, which are promising alternatives to demonstration reactors. A reference design for the Force Free Helical Reactor (FFHR) is presented, the main feature of which is the force-free-like configuration of the helical coils. Another feature is the selection of molten-salt Flibe as a self-cooling tritium breeder, which enhances safety. Demo-relevant engineering issues in the concept definition phase are discussed. (author)

  16. 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

  17. The HTGR type reactor ready for entering the market place

    International Nuclear Information System (INIS)

    ASEA Brown Boveri Mannheim/HRB have readied the pebble-bed high-temperature reactor for entering the market-place with their successful commissioning of the THTR-300 in Hamm-Uentrop handed over to their customer HKG on 1st June 1987. Interim results of the performance of the THTR-300 are excellent since it went into output operation at the end of 1985. This plant can be used as a reference plant for an innovative advanced reactor technology with particular technical and safety-related features at demand-effective output sizes (HTR-500, HTR-100, GHR-10) which will be available for future energy supply to cover not only power generation but also cogeneration of heat and power, and applications in the heat market involving temperatures between ca. 1000 and 9500C. (orig.)

  18. 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.)

  19. 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)

  20. 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.

  1. 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)

  2. 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)

  3. Quantity of 135I Released from the AGR 1, AGR 2, and AGR 3/4 Experiments and Discovery of 131I at the FPMS Traps during the AGR-3/4 Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Dawn Scates

    2014-09-01

    A series of three Advanced Gas Reactor (AGR) experiments have been conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). From 2006 through 2014, these experiments supported the development and qualification of the new U.S. tristructural isotropic (TRISO) particle fuel for Very High Temperature Reactors (VHTR). Each AGR experiment consisted of multiple fueled capsules, each plumbed for independent temperature control using a mix of helium and neon gases. The gas leaving a capsule was routed to individual Fission Product Monitor (FPM) detectors. For intact fuel particles, the TRISO particle coatings provide a substantial barrier to fission product release. However, particles with failed coatings, whether because of a minute percentage of initially defective particles, those which fail during irradiation, or those designed to fail (DTF) particles, can release fission products to the flowing gas stream. Because reactive fission product elements like iodine and cesium quickly deposit on cooler capsule components and piping structures as the effluent gas leaves the reactor core, only the noble fission gas isotopes of Kr and Xe tend to reach FPM detectors. The FPM system utilizes High Purity Germanium (HPGe) detectors coupled with a thallium activated sodium iodide NaI(Tl) scintillator. The HPGe detector provides individual isotopic information, while the NaI(Tl) scintillator is used as a gross count rate meter. During irradiation, the 135mXe concentration reaching the FPM detectors is from both direct fission and by decay of the accumulated 135I. About 2.5 hours after irradiation (ten 15.3 minute 135mXe half lives) the directly produced 135mXe has decayed and only the longer lived 135I remains as a source. Decay systematics dictate that 135mXe will be in secular equilibrium with its 135I parent, such that its production rate very nearly equals the decay rate of the parent, and its concentration in the flowing gas stream will appear to decay

  4. 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

  5. Corrosion Inhibition Studies in Support of the Long Term Storage of AGR Fuel

    International Nuclear Information System (INIS)

    Thorp receipt and storage (TR&S), Sellafield, UK has been chosen as the optimum facility to be used for the long term, interim wet storage of Advanced Gas Reactor (AGR) fuel. TR&S is currently a demineralised water pond due to a materials compatibility issue. However, a proportion of AGR spent fuel is known to be susceptible to corrosion through inter-granular attack (IGA). To avoid this, the chosen interim storage regime for AGR fuel is sodium hydroxide dosed pond water to pH 11.4. Some fuel has been safely stored in these conditions for 24 years in other Sellafield fuel ponds. For the period until the materials that are incompatible with pH 11.4 have been removed from the pond, a temporary pond water chemistry will be introduced to offer some corrosion protection. Lead container studies have been used to evaluate sodium nitrate (10 ppm) and ‘Low Dose’ sodium hydroxide (pH 9). ‘Low Dose’ conditions have been chosen for implementation. Objectives: • To underpin and implement temporary pond water chemistry for AGR storage; • To develop and deliver a programme of work to support the transition and safety case delivery for interim storage of AGR fuel

  6. Survival of Listeria monocytogenes in Soil Requires AgrA-Mediated Regulation.

    Science.gov (United States)

    Vivant, Anne-Laure; Garmyn, Dominique; Gal, Laurent; Hartmann, Alain; Piveteau, Pascal

    2015-08-01

    In a recent paper, we demonstrated that inactivation of the Agr system affects the patterns of survival of Listeria monocytogenes (A.-L. Vivant, D. Garmyn, L. Gal, and P. Piveteau, Front Cell Infect Microbiol 4:160, http://dx.doi.org/10.3389/fcimb.2014.00160). In this study, we investigated whether the Agr-mediated response is triggered during adaptation in soil, and we compared survival patterns in a set of 10 soils. The fate of the parental strain L. monocytogenes L9 (a rifampin-resistant mutant of L. monocytogenes EGD-e) and that of a ΔagrA deletion mutant were compared in a collection of 10 soil microcosms. The ΔagrA mutant displayed significantly reduced survival in these biotic soil microcosms, and differential transcriptome analyses showed large alterations of the transcriptome when AgrA was not functional, while the variations in the transcriptomes between the wild type and the ΔagrA deletion mutant were modest under abiotic conditions. Indeed, in biotic soil environments, 578 protein-coding genes and an extensive repertoire of noncoding RNAs (ncRNAs) were differentially transcribed. The transcription of genes coding for proteins involved in cell envelope and cellular processes, including the phosphotransferase system and ABC transporters, and proteins involved in resistance to antimicrobial peptides was affected. Under sterilized soil conditions, the differences were limited to 86 genes and 29 ncRNAs. These results suggest that the response regulator AgrA of the Agr communication system plays important roles during the saprophytic life of L. monocytogenes in soil. PMID:26002901

  7. 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.)

  8. Sloshing and fluid-structure interaction in a 400-MWe pool-type advanced fast reactor

    International Nuclear Information System (INIS)

    This paper describes the seismic analysis of a 400-MWe advanced fast reactor under 0.3 g SSE ground excitation. Two types of analyses are performed - the sloshing analysis and the fluid-structure interaction analysis. In the sloshing analysis, the sloshing frequency and wave patterns are calculated. The maximum wave height and the sloshing forces exerted on the submerged components and the primary tank are evaluated. In the fluid-structure interaction analysis, the maximum horizontal acceleration for the reactor core and the relative displacement between the reactor core and UIS are examined. The fluid-coupling phenomena between various components are investigated. Seismic stresses at critical areas are examined. The results obtained from this study are very useful to the design of the advanced reactors. Meanwhile, the computer code FLUSTR-ANL has proved to be a useful analytical tool for assessing the complicated seismic fluid-structure interactions and sloshing in the fast reactor systems. 10 refs., 25 figs

  9. Dismantling of Loop-Type Channel Equipment of MR Reactor in NRC 'Kurchatov Institute' - 13040

    International Nuclear Information System (INIS)

    In 2009 the project of decommissioning of MR and RTF reactors was developed and approved by the Expert Authority of the Russian Federation (Gosexpertiza). The main objective of the decommissioning works identified in this project: - complete dismantling of reactor equipment and systems; - decontamination of reactor premises and site in accordance with the established sanitary and hygienic standards. At the preparatory stage (2008-2010) of the project the following works were executed: loop-type channels' dismantling in the storage pool; experimental fuel assemblies' removal from spent fuel repositories in the central hall; spent fuel assembly removal from the liquid-metal-cooled loop-type channel of the reactor core and its placement into the SNF repository; and reconstruction of engineering support systems to the extent necessary for reactor decommissioning. The project assumes three main phases of dismantling and decontamination: - dismantling of equipment/pipelines of cooling circuits and loop-type channels, and auxiliary reactor equipment (2011-2012); - dismantling of equipment in underground reactor premises and of both MR and RTF in-vessel devices (2013-2014); - decontamination of reactor premises; rehabilitation of the reactor site; final radiation survey of reactor premises, loop-type channels and site; and issuance of the regulatory authorities' de-registration statement (2015). In 2011 the decommissioning license for the two reactors was received and direct MR decommissioning activities started. MR primary pipelines and loop-type facilities situated in the underground reactor hall were dismantled. Works were also launched to dismantle the loop-type channels' equipment in underground reactor premises; reactor buildings were reconstructed to allow removal of dismantled equipment; and the MR/RTF decommissioning sequence was identified. In autumn 2011 - spring 2012 results of dismantling activities performed are: - equipment from underground rooms (No. 66, 66

  10. 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.

  11. Assessment of subcriticality during PWR-type reactor refueling

    International Nuclear Information System (INIS)

    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-α and Feynman-α 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)

  12. 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.

  13. Experience with safety assessment of digital upgrading of IandC in VVER type reactors

    International Nuclear Information System (INIS)

    The digital upgrading of IandC systems important to safety in WWER type reactors requires a broad expertise in various knowledge fields. The approach of the Institute for safety Technology to the qualification and categorization of safety-critical software systems is highlighted. The role of the Institute in the qualification of the Teleperm XS and the type testing of its components is described. The aspects of the safety assessment of digital IandC systems in WWER type reactors is discussed in some detail. (A.K.)

  14. Design of film-type reactors with laminar film flow and slip at wall

    International Nuclear Information System (INIS)

    This paper proposes a method for calculating the height of a film-type reactor with allowance for the different residence times of individual layers of a draining laminar liquid film with slip at a wall. The probability, density function is found for the flow element distribution with respect to residence time in the reaction one. Formulas are suggested for calculating reactor length for zero- and-first-order chemical reactions. Optimal organization of liquid film flow in chemical reactors is discussed. 6 refs., 4 figs

  15. Model for spatial synthesis of automated control system of the GCR type reactor

    International Nuclear Information System (INIS)

    This paper describes the model which was developed for synthesis of spatial distribution of automated control elements in the reactor. It represents a general reliable mathematical model for analyzing transition states and synthesis of the automated control and regulation systems of GCR type reactors. One-dimensional system was defined under assumption that the time dependence of parameters of the neutron diffusion equation are identical in the total volume of the reactor and that spatial distribution of neutrons is time independent. It is shown that this assumption is satisfactory in case of short term variations which are relevant for safety analysis

  16. 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.

  17. Investigating The Neutron Flux Distribution Of The Miniature Neutron Source Reactor MNSR Type

    International Nuclear Information System (INIS)

    Neutron flux distribution is the important characteristic of nuclear reactor. In this article, four energy group neutron flux distributions of the miniature neutron source reactor MNSR type versus radial and axial directions are investigated in case the control rod is fully withdrawn. In addition, the effect of control rod positions on the thermal neutron flux distribution is also studied. The group constants for all reactor components are generated by the WIMSD code, and the neutron flux distributions are calculated by the CITATION code. The results show that the control rod positions only affect in the planning area for distribution in the region around the control rod. (author)

  18. 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)

  19. 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)

  20. 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)

  1. Optimization of the sipping test procedure for WWER-type reactors

    International Nuclear Information System (INIS)

    The endeavour to provide for a lower activity level of the fission products in the power reactor coolant during operation is associated with timely identification and unloading of failed fuel elements from the core. In this connection studies have been performed to optimize the sipping test procedures existing for the WWER-440 and WWER-1000 type reactors. Possible incorrectness in the sipping test is discussed, some methods for increasing accuracy and reliability of the results are proposed. (orig.)

  2. Generic Investigations on Transport Theory Modelling of High Temperature Reactors of Pebble Bed Type

    OpenAIRE

    Sureda Sureda, Antonio Jaime

    2008-01-01

    The GRS (Gesellschaft fuer Anlagen und Reaktorsicherheit = Company for Plant and Reactor Safety) maintains and further develops the code system DORT-TD/HERMIX-DIREKT, which is a complex tool for the simulation of coupled neutronics/thermal-hydraulics transients and accident scenarios of high-temperature gas cooled reactors of pebble bed type. With this tool, GRS takes part in the international benchmark activity "OECD/NEA PBMR400 Transient Benchmark”, which aims at the simulation of transient...

  3. About the transportation of WWER-type reactor spent nuclear fuel

    International Nuclear Information System (INIS)

    An intensive transition of existing NPPs with WWER-type reactors to new types of nuclear fuel and, therefore, to new fuel cycles, can be observed. New fuel cycles increase the burnup of nuclear fuel, and it increases the cost-effectiveness of its utilization

  4. 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.)

  5. Application of materials and fabrication with respect to current and next generation AGR steam generators (based on Hinkley B and Hunterston B AGRs)

    International Nuclear Information System (INIS)

    The Magnox Reactor experience is briefly reviewed showing the progressive development through Oldbury to the current A.G.R. The increasing power density has given rise to more restrictions in the fabrication of the steam generator. This has resulted in a steadily increasing demand for mechanised fabrication techniques to ensure adequate precision during manufacture. The Hinkley Hunterston A.G.R. reflects this in the development and utilisation of orbital welding, 1 D bends and remote repair devices together with controlled atmosphere heat treatment. Identification of fretting phenomena associated with the original design of Hinkley B steam generator supports resulted in a welded support design and an extensive substantiation programme. The theme of continuity of experience is maintained in the programme of work now in hand for the second generation A.G.R.'s (Heysham II and Torness). Material specifications have been strongly influenced by the current A.G.R programme and fabrication experience. Special attention is being given to improved fabrication techniques, typically improved orbital welding equipment will be implemented progressively in accordance with production acceptance trials. (author)

  6. Status of development - An integral type small reactor MRX in JAERI

    International Nuclear Information System (INIS)

    JAERI is conducting a design study on an integral type small reactor MRX for the use of nuclear ships. The basic concept of the reactor system is the integral type reactor with in-vessel steam generators and control rod drive systems, however, such new technologies as the water-filled containment, the passive decay heat removal system, the advanced automatic system, etc., are adopted to satisfy the essential requirements for the next generation ship reactors, i.e. compact, light, highly safe and easy operation. Research and development (R and D) works have being progressed on the peculiar components, the advanced automatic operation systems and the safety systems. Feasibility study and the economical evaluation of nuclear merchant ships have also being performed. The experiments and analysis of the safety carried out so far are proving that the passive safety features applied into the MRX are sufficient functions in the safety point of view. The MRX is a typical small type reactor realizing the easy operation by simplifying the reactor systems adopting the passive safety systems, therefore, it has wide variety of use as energy supply systems. This paper summarizes the present status on the design study of the MRX and the research and development activities as well as the some results of feasibility study. (author)

  7. Ceramographic Examinations of Irradiated AGR-1 Fuel Compacts

    Energy Technology Data Exchange (ETDEWEB)

    Paul Demkowicz; Scott Ploger; John Hunn

    2012-05-01

    The AGR 1 experiment involved irradiating 72 cylindrical fuel compacts containing tri-structural isotropic (TRISO)-coated particles to a peak burnup of 19.5% fissions per initial metal atom with no in-pile failures observed out of almost 300,000 particles. Five irradiated AGR 1 fuel compacts were selected for microscopy that span a range of irradiation conditions (temperature, burnup, and fast fluence). These five compacts also included all four TRISO coating variations irradiated in the AGR experiment. The five compacts were cross-sectioned both transversely and longitudinally, mounted, ground, and polished after development of careful techniques for preserving particle structures against preparation damage. Approximately 40 to 80 particles within each cross section were exposed near enough to mid-plane for optical microscopy of kernel, buffer, and coating behavior. The microstructural analysis focused on kernel swelling and porosity, buffer densification and fracture, debonding between the buffer and inner pyrolytic carbon (IPyC) layers, and fractures in the IPyC and SiC layers. Three basic particle morphologies were established according to the extent of bonding between the buffer and IPyC layers: complete debonding along the interface (Type A), no debonding along the interface (Type B), and partial debonding (Type AB). These basic morphologies were subdivided according to whether the buffer stayed intact or fractured. The resulting six characteristic morphologies were used to classify particles within each cross section, but no spatial patterns were clearly observed in any of the cross-sectional morphology maps. Although positions of particle types appeared random within compacts, examining a total of 830 classified particles allowed other relationships among morphological types to be established.

  8. Ceramographic Examinations of Irradiated AGR-1 Fuel Compacts

    Energy Technology Data Exchange (ETDEWEB)

    Paul Demkowicz; Scott Ploger; John Hunn; Jay S. Kehn

    2012-09-01

    The AGR 1 experiment involved irradiating 72 cylindrical fuel compacts containing tri-structural isotropic (TRISO)-coated particles to a peak burnup of 19.5% fissions per initial metal atom with no in-pile failures observed out of almost 300,000 particles. Six irradiated AGR 1 fuel compacts were selected for microscopy that span a range of irradiation conditions (temperature, burnup, and fast fluence). These six compacts also included all four TRISO coating variations irradiated in the AGR experiment. The six compacts were cross-sectioned both transversely and longitudinally, mounted, ground, and polished after development of careful techniques for preserving particle structures against preparation damage. From 36 to 79 particles within each cross section were exposed near enough to midplane for optical microscopy of kernel, buffer, and coating behavior. The microstructural analysis focused on kernel swelling and porosity, buffer densification and fracture, debonding between the buffer and inner pyrolytic carbon (IPyC) layers, and fractures in the IPyC and SiC layers. Three basic particle morphologies were established according to the extent of bonding between the buffer and IPyC layers: complete debonding along the interface (Type A), no debonding along the interface (Type B), and partial debonding (Type AB). These basic morphologies were subdivided according to whether the buffer stayed intact or fractured. The resulting six characteristic morphologies were used to classify particles within each cross section, but no spatial patterns were clearly observed in any of the cross-sectional morphology maps. Although positions of particle types appeared random within compacts, examining a total of 931 classified particles allowed other relationships among morphological types to be established.

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

    Energy Technology Data Exchange (ETDEWEB)

    Ali Khan, Liaquat; Ahmad, Nasir E-mail: epg.piaas@dgcc.org.pk; Zafar, M.S.; Ahmad, Ayaz

    2000-07-01

    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.

  10. 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

  11. 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

  12. 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)

  13. Hydride blister formation simulation in Candu type reactors

    International Nuclear Information System (INIS)

    We have developed a computer code for the probability study of hydride blister formation in pressure tubes named BLIFO. The basic hypothesis of the model are: the pressure tube is divided into five areas according to the existence of four garter springs. For each area the probability of blister formation is the probability of the hydrogen content exceeding a critical threshold when contact tube is present; the probability of a blister in a tube is the OR combination of the probabilities of a blister in each area; the tube contact is a function of the garter springs location, and the time; the critical hydrogen threshold is sorted over the areas within the pressure tube; hydrogen pick-up rate was sorted with a Gaussian distribution; the initial hydrogen content values for each tube were measured before the ensamble and they are used in the code. For Embalse evaluation, we build up a subroutine that simulate Gaussian distribution using the parameters of a typical nuclear power Candu reactor garter spring distribution. (author)

  14. Consideration of BORAX-type reactivity accidents applied to research reactors; Prise en compte des accidents de type 'BORAX' pour les reacteurs de recherche

    Energy Technology Data Exchange (ETDEWEB)

    Couturier, Jean; Meignen, Renaud; Bourgois, Thierry; Biaut, Guillaume; Mireau, Jean-Pierre [Direction de la surete des reacteurs, Institut de Radioprotection et de Surete Nucleaire - IRSN, 31, avenue de la Division Leclerc, 92260 Fontenay-aux-Roses (France); Natta, Marc [Direction de la strategie, du developpement et des partenariats, Institut de Radioprotection et de Surete Nucleaire - IRSN, 31, avenue de la Division Leclerc, 92260 Fontenay-aux-Roses (France)

    2011-08-08

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

  15. Effect of Core Configurations on Burn-Up Calculations For MTR Type Reactors

    International Nuclear Information System (INIS)

    Three-dimensional burn-up calculations of MTR-type research reactor were performed using different patterns of control rods , to examine their effect on power density and neutron flux distributions throughout the entire core and on the local burn-up distribution. Calculations were performed using the computer codes' package MTRPC system, using the cell calculation transport code WIMS-D4 and the core calculation diffusion code CITVAP. A depletion study was done and the effects on the reactor fuel were studied, then an empirical formula was generated for every fuel element type, to correlate irradiation to burn-up percentage. Keywords: Neutronic Calculations, Burn-Up, MTR-Type Research Reactors, MTRPC Package, Empirical Formula For Fuel Burn-Up.

  16. 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

  17. Coolant clean-up method in PWR type reactors

    International Nuclear Information System (INIS)

    Purpose: To perform coolant clean-up while climinating the need of replacing boric acid with plant primary coolants and using anionic exchange resins in which the amount of Cl anionic exchange resins as impurities is decreased. Method: OH type anionic exchange resins are previously treated with an aqueous boric acid not containing radioactivity at a place other than the equipment for use (that is desalting tower) into boric acid type anionic ion exchange resins and, thereafter, the boric acid type anionic exchange resins are filled into a desalting tower of the clean-up system to perform primary coolant clean-up. In this case, since the resins can be used directly for the purpose without performing boric acid replacement after charging into the equipment for use, the procedures in the plant being in operation can be saved. (Yoshino, Y.)

  18. 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

  19. Early detection of coolant boiling in research reactors with MTR-type fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kozma, R.; Turkcan, E.; Verhoef, J.P. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands)

    1992-10-01

    A reactor core monitoring system having the function of early detection of boiling in the coolant channels of research reactors with MTR-type fuel is introduced. The system is based on the on-line analysis of signals of various ex-core and in-core neutron detectors. Early detection of coolant boiling cannot be accomplished by the evaluation of the DC components of these detectors in a number of practically important cases of boiling anomaly. It is shown that the noise component of the available neutron detector signals can be used for the detection of boiling in these cases. Experiments have been carried out at a boiling setup in the research reactor HOR of the Interfaculty Reactor Institute, Technical University of Delft, The Netherlands. (author). 8 refs., 11 figs.

  20. Early detection of coolant boiling in research reactors with MTR-type fuel

    International Nuclear Information System (INIS)

    In this paper, a reactor core monitoring system having the function of early detection of boiling in the coolant channels of research reactors with MTR-type fuel is introduced. The system is based on the on-line analysis of signals of various ex-core and in-core neutron detectors. Early detection of coolant boiling cannot be accomplished by the evaluation of the DC components of these detectors in a number of practically important cases of boiling anomaly. It is shown that noise component of the available neutron detector signals can be used for the detection of boiling in these cases. Experiments have been carried out at a boiling setup in the research reactor HOR of the Interfaculty Reactor Institute, Technical University Delft, The Netherlands

  1. Early detection of coolant boiling in research reactors with MTR-type fuel

    International Nuclear Information System (INIS)

    A reactor core monitoring system having the function of early detection of boiling in the coolant channels of research reactors with MTR-type fuel is introduced. The system is based on the on-line analysis of signals of various ex-core and in-core neutron detectors. Early detection of coolant boiling cannot be accomplished by the evaluation of the DC components of these detectors in a number of practically important cases of boiling anomaly. It is shown that the noise component of the available neutron detector signals can be used for the detection of boiling in these cases. Experiments have been carried out at a boiling setup in the research reactor HOR of the Interfaculty Reactor Institute, Technical University of Delft, The Netherlands. (author). 8 refs., 11 figs

  2. 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)

  3. 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)

  4. 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

  5. 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)

  6. The measuring set: reactor power meter (type of SG-8), reactor energy meter (type of SG-11) and digital dose meter (type of SG-9) for reactor rigs operation

    International Nuclear Information System (INIS)

    A measuring set consisting of the Reactor Power Meter, Reactor Energy Meter and Digital Dose Meter is described. The gamma radiation of water in the reactor primary cooling circuit reaches the ionisation chamber and involves the output current, driving the Reactor Power Meter and Reactor Energy Meter. The Digital Dose Meter is controlled by the output current of the self-powered detector mounted inside the reactor rig. (author)

  7. 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

  8. Analysis of Fission Products on the AGR-1 Capsule Components

    Energy Technology Data Exchange (ETDEWEB)

    Paul A. Demkowicz; Jason M. Harp; Philip L. Winston; Scott A. Ploger

    2013-03-01

    The components of the AGR-1 irradiation capsules were analyzed to determine the retained inventory of fission products in order to determine the extent of in-pile fission product release from the fuel compacts. This includes analysis of (i) the metal capsule components, (ii) the graphite fuel holders, (iii) the graphite spacers, and (iv) the gas exit lines. The fission products most prevalent in the components were Ag-110m, Cs 134, Cs 137, Eu-154, and Sr 90, and the most common location was the metal capsule components and the graphite fuel holders. Gamma scanning of the graphite fuel holders was also performed to determine spatial distribution of Ag-110m and radiocesium. Silver was released from the fuel components in significant fractions. The total Ag-110m inventory found in the capsules ranged from 1.2×10 2 (Capsule 3) to 3.8×10 1 (Capsule 6). Ag-110m was not distributed evenly in the graphite fuel holders, but tended to concentrate at the axial ends of the graphite holders in Capsules 1 and 6 (located at the top and bottom of the test train) and near the axial center in Capsules 2, 3, and 5 (in the center of the test train). The Ag-110m further tended to be concentrated around fuel stacks 1 and 3, the two stacks facing the ATR reactor core and location of higher burnup, neutron fluence, and temperatures compared with Stack 2. Detailed correlation of silver release with fuel type and irradiation temperatures is problematic at the capsule level due to the large range of temperatures experienced by individual fuel compacts in each capsule. A comprehensive Ag 110m mass balance for the capsules was performed using measured inventories of individual compacts and the inventory on the capsule components. For most capsules, the mass balance was within 11% of the predicted inventory. The Ag-110m release from individual compacts often exhibited a very large range within a particular capsule.

  9. 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

  10. 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

  11. 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)

  12. AGR-5/6/7 LEUCO Kernel Fabrication Readiness Review

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, Douglas W. [Idaho National Lab. (INL), Idaho Falls, ID (United States). Fuel Design and Development; Bailey, Kirk W. [Idaho National Lab. (INL), Idaho Falls, ID (United States). ART Quality Assurance Engineer

    2015-02-01

    In preparation for forming low-enriched uranium carbide/oxide (LEUCO) fuel kernels for the Advanced Gas Reactor (AGR) fuel development and qualification program, Idaho National Laboratory conducted an operational readiness review of the Babcock & Wilcox Nuclear Operations Group – Lynchburg (B&W NOG-L) procedures, processes, and equipment from January 14 – January 16, 2015. The readiness review focused on requirements taken from the American Society Mechanical Engineers (ASME) Nuclear Quality Assurance Standard (NQA-1-2008, 1a-2009), a recent occurrence at the B&W NOG-L facility related to preparation of acid-deficient uranyl nitrate solution (ADUN), and a relook at concerns noted in a previous review. Topic areas open for the review were communicated to B&W NOG-L in advance of the on-site visit to facilitate the collection of objective evidences attesting to the state of readiness.

  13. DETERMINATION OF THE QUANTITY OF I-135 RELEASED FROM THE AGR-1 TEST FUELS AT THE END OF ATR OPERATING CYCLE 138B

    Energy Technology Data Exchange (ETDEWEB)

    J. K. Hartwell; D. M. Scates; J. B. Walter; M. W. Drigert

    2007-05-01

    The AGR-1 experiment is a multiple fueled-capsule irradiation experiment being conducted in the Advanced Test Reactor (ATR) in support of the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. The experiment began irradiation in the ATR with a cycle that reached full power on December 26, 2006 and ended with shutdown of the reactor for a brief outage on February 10, 2007 at 0900. The AGR-1 experiment will continue cyclical irradiation for about 2.5 years. In order to allow estimation of the amount of radioiodine released during the first cycle, purge gas flow to all capsules continued for about 4 days after reactor shutdown. The FPMS data acquired during part of that shutdown flow period has been analyzed to elucidate the level of 135I released during the operating cycle.

  14. 15 CFR 740.18 - Agricultural commodities (AGR).

    Science.gov (United States)

    2010-01-01

    ... 15 Commerce and Foreign Trade 2 2010-01-01 2010-01-01 false Agricultural commodities (AGR). 740.18... EXCEPTIONS § 740.18 Agricultural commodities (AGR). (a) Eligibility requirements. License Exception AGR permits the export of agricultural commodities to Cuba, as well as the reexport of U.S....

  15. Microscopic analysis of irradiated AGR-1 coated particle fuel compacts

    Energy Technology Data Exchange (ETDEWEB)

    Scott Ploger; Paul Demkowicz; John Hunn; Robert Morris

    2012-10-01

    The AGR-1 experiment involved irradiation of 72 TRISO-coated particle fuel compacts to a peak burnup of 19.5% FIMA with no in-pile failures observed out of 3×105 total particles. Irradiated AGR-1 fuel compacts have been cross-sectioned and analyzed with optical microscopy to characterize kernel, buffer, and coating behavior. Five compacts have been examined so far, spanning a range of irradiation conditions (burnup, fast fluence, and irradiation temperature) and including all four TRISO coating variations irradiated in the AGR-1 experiment. The cylindrical specimens were sectioned both transversely and longitudinally, then polished to expose between approximately 40-80 individual particles on each mount. The analysis focused primarily on kernel swelling and porosity, buffer densification and fracturing, buffer-IPyC debonding, and fractures in the IPyC and SiC layers. Characteristic morphologies have been identified, over 800 particles have been classified, and spatial distributions of particle types have been mapped. No significant spatial patterns were discovered in these cross sections. However, some trends were found between morphological types and certain behavioral aspects. Buffer fractures were found in approximately 23% of the particles, and these fractures often resulted in unconstrained kernel swelling into the open cavities. Fractured buffers and buffers that stayed bonded to IPyC layers appear related to larger pore size in kernels. Buffer-IPyC interface integrity evidently factored into initiation of rare IPyC fractures. Fractures through part of the SiC layer were found in only three particles, all in conjunction with IPyC-SiC debonding. Compiled results suggest that the deliberate coating fabrication variations influenced the frequencies of IPyC fractures, IPyC-SiC debonds, and SiC fractures.

  16. Loss Of Secondary coolant accident analysis for Pius type reactor using Relap5/MOD2

    International Nuclear Information System (INIS)

    Process inherent ultimate safety (Pius) reactor concept is a reactor concept that intrinsically based on passive safety. This reactor refer to Pressurized water reactor (PWR) wherein the primary system is submerged in a pool of poison water. the operating principle is to maintain the pressure balance, so that no inflow from pool to the primary system. On loss of secondary coolant accident, primary coolant temperature increases, it is followed by the increase of primary pump speed. When the upper limit is reached, the pump is tripped due to the pressure balance disturbance, poison water flows from pool to the primary system, then reactor shut down. this accident condition was simulated by experimental and numerical simulation using RELAP5/MOD2. Numerical simulation was done to the experimental apparatus nodalization that was set on the norm of RELAP5/MOD2. This nodalization consist of 119 volumes, 127 junctions, and 106 heat structures. Analysis was carried out using both experimental and numerical simulation results. it can be concluded that PIUS type reactor is able to anticipate loss of secondary coolant accident because its capability of self shut down

  17. A lumped parameter core dynamics model for MTR type research reactors under natural convection regime

    International Nuclear Information System (INIS)

    Highlights: ► A model is presented to simulate the reactivity insertion transient in MTR reactors. ► Transient dynamics of IAEA 10 MW MTR type research reactor are evaluated. ► Maximum unprotected reactivity insertion for safe condition is calculated. ► The model predictions are validated with corresponding results in the literature. - Abstract: On the basis of lumped parameter modeling of both the kinetic and thermal–hydraulic effects, a reasonably accurate simplified model has been developed to predict the dynamic response of MTR reactors following to an unprotected reactivity insertion under natural convection regime. By this model the reactor transient behavior at a given initial steady-state can be solved by a set of ordinary differential equations. The model predictions have an acceptable consent with corresponding results of reactivity insertion transients analyzed in the literature. The inherent safety characteristics of MTR research reactors utilizing natural convection is clearly demonstrated by the expanded model. The safety margin of reactor operating is selected ONB condition and thereby the proposed model determines that any slight increase in the value of $0.73 for inserted reactivity will cause the maximum cladding surface temperature to exceed the ONB condition

  18. 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.)

  19. Fast reactor fuel pin behavior analyses in a LOF type transient event

    International Nuclear Information System (INIS)

    In order to evaluate integrity limiting parameters of fuel pins during fast reactor core transient events, such as fuel center line temperature and cladding maximum temperature, fuel pin behavior calculations were made using the fast reactor fuel pin performance code CEDAR. The temperature histories of fuel pins during a loss of flow (LOF) type transient events was calculated based on Ross and Stoute type gap conductance model and constant gap conductance model, which is used in a core transient calculation code like HIPRAC. The calculated maximum temperatures of cladding and adjacent coolant channel were lower in the case with Ross and Stoute type model than in the case of constant gap conductance model due to the dynamic change of gap conductance of former case. It is indicated that core transient calculations with constant gap conductance give conservative cladding and coolant temperatures than that with Ross and Stoute type gap conductance model which is thought to be realistic. (author)

  20. 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.)

  1. Device for estimating neutron flux distribution in BWR type reactor

    International Nuclear Information System (INIS)

    Purpose: To convert the neutron flux distribution obtained by the migration of TIP (moving type neutron detector) into the power distribution and to accurately estimate the neutron flux distribution between LPRMs (neutron detectors) on the basis of the thus converted power distribution. Constitution: A computer calculates the infinite multiplication factor K sub(infinity)sup(+) (K) from a TIP indication signal and a control rod positioning signal, and converts the resulting value into a value corresponding to K sub(infinity)sup(+) (K) in a case where the control rods are not inserted, thus sending the value to a K sub(infinity)sup( b) (K) memory device, the TIP indication value computer estimates and calculates the TIP indication value from K sub(infinity)sup( b) (K) stored in the memory device when the surveilance of the neutron flux distribution is required, and the control rod positioning signal and LPRM indication value signal at that time, and sends the resulting value to an indicator and a recording device. (Nakamura, S.)

  2. Estudio de Salud Agrícola

    Science.gov (United States)

    En 1993, científicos del Instituto Nacional del Cáncer, Instituto Nacional de Ciencias Ambientales y la Agencia de Protección Ambiental de Estados Unidos iniciaron un estudio conocido como Estudio de Salud Agrícola (AHS).

  3. Development of A Conservative Method for A Feedwater Pipe Break Analysis of An Integral Type Reactor

    International Nuclear Information System (INIS)

    The development of advanced small and medium sized nuclear power plants for multipurpose appears before the footlights, and some of them are ready for construction. The SMART, which is an integral pressurized water reactor is one of those advanced types of small sized nuclear reactors. The basic design of SMART was completed at the Korea Atomic Energy Research Institute. A new phase in order to test and verify the SMART design is currently underway in Korea. The results of these tests and verifications will be fed back into the SMART design for a further improvement of the safety and reliability. The integral type reactor can be mitigated design basis events by a reactor protection system, or engineered safety features. The consequences of design basis events must be less than the established acceptance limits and provide an acceptable margin to protect the health and safety. The design basis events are divided into general categories corresponding to their effect on a plant. One of these categories is a decrease in a heat removal by the secondary system. There are a turbine trip, a main steam isolation valve closure, a loss of the primary component cooling system, and a feedwater pipe break for the decrease in the heat removal by the secondary system. The feedwater pipe break accident is one of the most important accidents in the safety of the integral type reactor. Decrease in the feedwater supply to the steam generators causes a decrease in the heat extraction rate from the reactor coolant system, resulting in an increase of a primary coolant temperature and a pressure, and the nuclear power decreases due to a reactivity feedback. Performed sensitivity analysis to find parameters affecting seriously in the integral reactor's feedwater pipe break accident. According to these parametric analysis results, a power level, an initial system pressure, a moderator reactivity coefficient and a break size are major parameters for the maximum system pressure. The detailed

  4. Data Compilation for AGR-1 Variant 3 Coated Particle Composite LEU01-49T

    Energy Technology Data Exchange (ETDEWEB)

    Hunn, John D [ORNL; Lowden, Richard Andrew [ORNL

    2006-07-01

    This document is a compilation of characterization data for the AGR-1 variant 3 coated particle composite LEU01-49T, a composite of three batches of TRISO-coated 350 {micro}m diameter 19.7% low enrichment uranium oxide/uranium carbide kernels (LEUCO). The AGR-1 TRISO-coated particles consist of a spherical kernel coated with a {approx} 50% dense carbon buffer layer (100 {micro}m nominal thickness) followed by a dense inner pyrocarbon layer (40 {micro}m nominal thickness) followed by a SiC layer (35 {micro}m nominal thickness) followed by another dense outer pyrcoarbon layer (40 {micro}m nominal thickness). The coated particles were produced by ORNL for the Advanced Gas Reactor Fuel Development and Qualification (AGR) program to be put into compacts for the fuel shakedown irradiation (AGR-1) experiment. The kernels were obtained from BWXT and identified as composite G73D-20-6302. The BWXT kernel lot G73D-20-69302 was riffled into sublots for characterization and coating by ORNL and identified as LEUO01-?? (where ?? is a series of integers beginning with 01).

  5. AGR-3/4 Final Data Qualification Report for ATR Cycles 151A through 155B-1

    Energy Technology Data Exchange (ETDEWEB)

    Pham, Binh T. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-03-01

    This report provides the qualification status of experimental data for the entire Advanced Gas Reactor 3/4 (AGR 3/4) fuel irradiation. AGR-3/4 is the third in a series of planned irradiation experiments conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) for the AGR Fuel Development and Qualification Program, which supports development of the advanced reactor technology under the INL ART Technology Development Office (TDO). The main objective of AGR-3/4 irradiation is to provide a known source of fission products for subsequent transport through compact matrix and structural graphite materials due to the presence of designed-to-fail fuel particles. Full power irradiation of the AGR 3/4 test began on December 14, 2011 (ATR Cycle 151A), and was completed on April 12, 2014 (end of ATR Cycle 155B) after 369.1 effective full power days of irradiation. The AGR-3/4 test was in the reactor core for eight of the ten ATR cycles between 151A and 155B. During the unplanned outage cycle, 153A, the experiment was removed from the ATR northeast flux trap (NEFT) location and stored in the ATR canal. This was to prevent overheating of fuel compacts due to higher than normal ATR power during the subsequent Powered Axial Locator Mechanism cycle, 153B. The AGR 3/4 test was inserted back into the ATR NEFT location during the outage of ATR Cycle 154A on April 26, 2013. Therefore, the AGR-3/4 irradiation data received during these 2 cycles (153A and 153B) are irrelevant and their qualification status isnot included in this report. Additionally, during ATR Cycle 152A the ATR core ran at low power for a short enough duration that the irradiation data are not used for physics and thermal calculations. However, the qualification status of irradiation data for this cycle is still covered in this report. As a result, this report includes data from 8 ATR Cycles: 151A, 151B, 152A, 152B, 154A, 154B, 155A, and 155B, as recorded in the Nuclear Data Management and

  6. Position indicator for movable coil type reactor control rod driving mechanism

    International Nuclear Information System (INIS)

    Purpose: To enable the accurate and continuous indication of the position of a movable coil type reactor control rod driving mechanism. Constitution: The position of an electromagnet magnetically coupled to a plunger connected to a reactor core control rod is detected by an electromagnet position detector, and the displacement of the positions of the electromagnet and the plunger is detected by a relative position detector connected to the electromagnet. The detected values of both the detectors are used to calculate the position of the driving mechanism. (Aizawa, K.)

  7. 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 ...

  8. Thermohydraulic analysis of loss of forced flow accident in a pool type reactor

    International Nuclear Information System (INIS)

    This paper shows a calculation model for the fuel and pool water temperatures, and the internal building pressure of a 5 MW pool-type reactor, in the hypothetical event of the forced flow interruption. it is obtained the solution of the thermal energy and the momentum conservation equations in one dimension, which represent the heat conduction and natural convection in the coolant. The reactor building pressure increment due to the partial pool water evaporation is also calculated, using a homogenous model with thermal equilibrium of the phases (liquid water and steam) and the existing air. The heat loss to the building walls is also considered. (Author)

  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. Hydrodynamic problems of heavy liquid metal coolant technology in loop-type or mono-block-type reactor installation

    International Nuclear Information System (INIS)

    Full text of publication follows: Reactor installations with heavy liquid metal coolant (HLMC) can be of loop-type and mono-block-type (integral) designs. The latter is usually characterized by fairly low HLMC velocity and pressure values, the availability of free levels at which efficient separation of solid and gaseous impurities is possible, the absence of auxiliary HLMC pumps and large HLMC volume. So far as the loop-type and integral reactors substantially differ in hydrodynamic conditions, the solution of their coolant technology problems is specific for the following. In loop-type reactors, shutoff and adjusting accessories can be used as part of coolant technology devices. Provision should be made in this case to prevent erosion of structural materials due to the HLMC impact as well as melt ingress into the gas lines. In integral reactors, - a more active impurity concentration and interaction is possible in the low velocity sections of the HLMC circulation. Therefore it is desirable to perform control over the HLMC state simultaneously by several sensors positioned in different parts of the loop. In this case, impurity deposition on the sensor surfaces and the influence of these deposits on sensor signal is not improbable; - it is necessary that the separation from the HLMC of solid impurities due to filtration be much more intensive than the separation of impurities on free surfaces; therefore it is desirable to employ filtering materials with high filtering characteristics, filters with large volume and/or filtering material surface and less hydraulic resistance as compared to those being used in loop-type reactors; - it is reasonable to adjust the content of oxygen impurity in HLMC using mass exchange apparatus with solid phase oxygen source; the required output of these apparatus is provided by HLMC flow rate specified by their own pumps; - to run efficient hydrogen cleaning of HLMC and circulation loop, it is necessary to provide, on the whole, the

  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. Application of integral codes for simulation of WWER type reactor off-normal modes

    International Nuclear Information System (INIS)

    The article concerns the experience in developing of calculating schemes, formation of input data set and its qualification using simulation of WWER-440 reactor installations off-normal modes by MELCOR-1.8.5 code as an example. Good reproducibility with experimental data is obtained and adequacy of suggested WWER-440 type reactor models to corresponding real installations is confirmed. The acceptability of non-emergency circulation circuit loops merging in calculating model is shown. The necessity of simulation of steam generator tube bundles along length on the first circuit side no less than by three calculating volumes is substantiated. The need of hydraulic lock and surge line simulation in view of three calculating volumes (horizontal and two vertical) is shown. The parameters of communications between calculating volumes are chose on the base of experimental results on corresponding reactor installations or their models

  13. Automatic control of neutron flux in experimental channels of the WWR-M type reactors

    International Nuclear Information System (INIS)

    The flowsheet of the neutron flux local regulator intended for maintaining the given level of neutron flux distribution in experimental channels of the WWR-M type reactor under stationary and transition modes, is suggested. The functional diagram of the electron regulation block (ERB) in considered. The regulator is tested when the reactor operates with the capacity of 13 MWt along with the staff system of automated regulation and without it. The experiments carried out demonstrate the stable operation of the entire control system and good performance characteristics of the ERR block. The conclusion is made that the suggested method of neutron flux automated regulation in experimental channels can be successfully extended to a higher number of experimental channels and applied at other research reactors. Small size fission chambers and direct charging detectors can be used in local systems as sensors

  14. 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)

  15. Channel blockage accident analysis for research reactors with MTR- type fuel elements

    International Nuclear Information System (INIS)

    It is the purpose of this study to investigate the feasibility of removing the residual decay heat from core of TR-2 ,which is a pool-type research reactor, after a channel blockage accident event and to identify the principal factors involved in cooling process. To analyze this accident scenery, THEAP-I computer code, which is a single phase transient 3-D structure/1-D flow thermal hydraulics code developed with the aim to contribute mainly to the safety analysis of the open pool research reactors, was modified and used. All of the analysis results figured out the fact that the core melting was inevitable in case of an uninterrupted operation (continuous operation) preceding a channel blockage accident of the TR-2 Reactor. Such a result will even be met if the blockage occurs only in a single fuel element. The results of analysis are expressed in terms of temperature field distribution as a function of time

  16. 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.)

  17. Avaliação do contato pneu-solo em três modelos de pneus agrícolas Evaluation of foot print for three agricultural tire types

    Directory of Open Access Journals (Sweden)

    Flávio R. Mazetto

    2004-12-01

    Full Text Available Este trabalho teve como objetivo avaliar as áreas de contato, as deformações elásticas dos pneus, a resistência do solo à penetração e os perfis do recalque no solo de três modelos de rodados pneumáticos. O ensaio seguiu um delineamento experimental casualizado, constituído por 12 tratamentos e quatro repetições, nos quais se avaliaram os modelos de pneus de carcaça diagonal, radial e o terceiro de configuração mista, denominado BPAF, inflados com as pressões ideais e submetidos a cargas radiais de 5; 10; 15 e 20 kN, simulando o que ocorre no campo. O ensaio dos pneus agrícolas foi realizado no Núcleo de Ensaio de Máquinas e Pneus Agrícolas - NEMPA, da Faculdade de Ciências Agronômicas - UNESP, Câmpus de Botucatu - SP, em uma prensa hidráulica sobre um tanque de solo. Os resultados relativos às áreas de contato e deformações elásticas mostraram valores maiores para o pneu BPAF. Os recalques do pneu BPAF no solo foram menores comparados aos outros rodados, e com os pneus radial e BPAF houve menores resistências do solo à penetração.The aim of this study was to evaluate the tire/ground contact areas, tire elastic deformation, cone index and foot printed on the ground of tire models. The experiment was done in 12 treatments and four repetitions with three tire models: bias ply, radial and bias belt. The tires were inflated with right inflation pressures and four radial loads were applied on the wheels: 5; 10; 15 and 20 kN, simulating what happens in the field. The agricultural tire tests were carried out at Sao Paulo State University, Botucatu - SP, Brazil, using a hydraulic press over the soil bin. The contact area and elastic deformation results showed highest values for bias belt tire. The soil sinkage of bias belt tire was smaller than other pneumatic tires. The radial and bias belt tires exercised smaller cone index than bias ply tire.

  18. Safety and environment aspects of Tokamak-type fusion power reactor - an overview

    International Nuclear Information System (INIS)

    Naturally occurring thermonuclear fusion reaction (of light atoms to form a heavier nucleus) in the sun and every star in the universe, releases incredible amounts of energy. Demonstrating the controlled and sustained reaction of deuterium-tritium plasma should enable the development of fusion as an energy source here on Earth. The promising fusion power reactors could be operated on the deuterium-tritium fuel cycle with fuel self-sufficiency. The potential impact of fusion power on the environment and the possible risks associated with operating large-scale fusion power plants is being studied by different countries. The results show that fusion can be a very safe and sustainable energy source. A fusion power plant possesses not only intrinsic advantages with respect to safety compared to other sources of energy, but also a negligible long term impact on the environment provided certain precautions are taken in its design. One of the important considerations is in the selection of low activation structural materials for reactor vessel. Selection of the materials for first wall and breeding blanket components is also important from safety issues. It is possible to fully benefit from the advantages of fusion energy if safety and environmental concerns are taken into account when considering the conceptual studies of a reactor design. The significant safety hazards are due to the tritium inventory and energetic neutron fluence induced activity in the reactor vessel, first wall components, blanket system etc. The potential of release of radioactivity under operational and accident conditions needs attention while designing the fusion reactor. This paper describes an overview of safety and environmental merits of fusion power reactor, issues and design considerations and need for R and D on safety and environmental aspects of Tokamak type fusion reactor. (author)

  19. Manufacture of steam generator units and components for the AGR power stations at Heysham II and Torness

    International Nuclear Information System (INIS)

    The current AGR Steam Generator is a development of the successful once-through units supplied for the Oldbury Magnox and Hinkley B/Hunterston B AGR power stations. In this paper a brief outline of the evolution of the steam generator design from the earlier gas cooled reactor stations is presented. A description of the main items of fabrication development is given. The production facilities for the manufacture of the units are described. Reference is also made to some of the work on associated components. The early experience on the construction site of installation of the steam generators is briefly outlined. (author)

  20. Design for reactor core safety in nuclear power plants

    International Nuclear Information System (INIS)

    This Guide covers the neutronic, thermal, hydraulic, mechanical, chemical and irradiation considerations important to the safe design of a nuclear reactor core. The Guide applies to the types of thermal neutron reactor power plants that are now in common use and fuelled with oxide fuels: advanced gas cooled reactor (AGR), boiling water reactor (BWR), pressurized heavy water reactor (PHWR) (pressure tube and pressure vessel type) and pressurized water reactor (PWR). It deals with the individual components and systems that make up the core and associated equipment and with design provisions for the safe operation of the core and safe handling of the fuel and other core components. The Guide discusses the reactor vessel internals and the reactivity control and shutdown devices mounted on the vessel. Possible effects on requirements for the reactor coolant, the reactor coolant system and its pressure boundary (including the pressure vessel) are considered only as far as necessary to clarify the interface with the Safety Guide on Reactor Coolant and Associated Systems in Nuclear Power Plants (IAEA Safety Series No. 50-SG-D13) and other Guides. In relation to instrumentation and control systems the guidance is mainly limited to functional requirements

  1. 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)

  2. Fuel cycle incentives for market introduction of new reactor types in Europe

    International Nuclear Information System (INIS)

    While all current new nuclear plant orders in the world, both in the Far East, Finland and France, are light water reactors, the question is under what circumstances non-LWR will be favourable from a customer (utility) point of view. Different reactor types offer different choices for fuel cycle strategy, and generate different quantities and compositions of spent fuel. Two replacement alternatives for the European reactor park were identified for this study: first, the evolutionary LWR with the possibility for use of MOX and Inert Matrix Fuel, second, the pebble bed HTR able to be loaded partially with plutonium/minor actinide fuel ('Deep Burn'). Inert Matrix Fuel offers the utility a possibility to keep the LWR while reducing transuranics production. Compared to LWR, the HTR offers enhanced plutonium burning capability. Plutonium and waste production is shown as a function of time for three LWR (EPR) and three HTR (PBMR) cases. With an HTR park actual plutonium consumption can be achieved, including the legacy plutonium of today's reactor park. Minimization of the total amount of waste can be achieved best with IMF waste in EPRs. Finally, the timescales of the major effects on waste quantities and composition are discussed: at the time of order decision of the replacement reactors, the geological waste storage facility will not be operational for several decades. (authors)

  3. Thermal Hydraulic Fortran Program for Steady State Calculations of Plate Type Fuel Research Reactors

    International Nuclear Information System (INIS)

    The safety assessment of Research and Power Reactors is a continuous process over their life and that requires verified and validated codes. Power Reactor codes all over the world are well established and qualified against a real measuring data and qualified experimental facilities. These codes are usually sophisticated, require special skills and consume much more running time. On the other hand, most of the Research Reactor codes still requiring more data for validation and qualification. Therefore it is benefit for a regulatory body and the companies working in the area of Research Reactor assessment and design to have their own program that give them a quick judgment. The present paper introduces a simple one dimensional Fortran program called THDSN for steady state best estimate Thermal Hydraulic (TH) calculations of plate type fuel RRs. Beside calculating the fuel and coolant temperature distribution and pressure gradient in an average and hot channel the program calculates the safety limits and margins against the critical phenomena encountered in RR such as the burnout heat flux and the onset of flow instability. Well known TH correlations for calculating the safety parameters are used. THDSN program is verified by comparing its results for 2 and 10 MW benchmark reactors with that published in IAEA publications and good agreement is found. Also the program results are compared with those published for other programs such as PARET and TERMIC. An extension for this program is underway to cover the transient TH calculations

  4. 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.

  5. Radiation damage structure in irradiated and annealed 440 WWER-Type reactor pressure vessel steels

    International Nuclear Information System (INIS)

    A review of irradiation damages in WWER-type RPV steels based on conventional Transmission Electron Microscopy investigations in a power reactor and a research reactor, is presented; the samples consist in Cr-Mo-V ferritic steel (15Kh2MFA type). The visible part of radiation-induced defects consists of very fine vanadium carbide precipitates, small dislocation loops and black dots (presumably corresponding to clusters and particle embryos formed from vacancies and solute-atoms (vanadium, copper, phosphorus) and carbon associated with vanadium. Radiation-induced defects are concentrated at dislocation substructure during irradiation in a power reactor, revealing the role of radiation-enhanced diffusion in damage structure forming process. Contrarily, the distribution of defects resulting from annealing of specimens irradiated in the research reactor is pre-determined by an homogenous distribution of radiation-induced defects prior to annealing. Increasing the number of re-irradiation and annealing cycles, the amount of dislocation loops among all defects seems to be growing. Simultaneously, the dislocation substructure recovers considerably. (authors). 14 refs., 11 figs., 3 tabs

  6. 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.)

  7. Data Compilation for AGR-1 Baseline Coated Particle Composite LEU01-46T

    Energy Technology Data Exchange (ETDEWEB)

    Hunn, John D [ORNL; Lowden, Richard Andrew [ORNL

    2006-04-01

    This document is a compilation of characterization data for the AGR-1 baseline coated particle composite LEU01-46T, a composite of four batches of TRISO-coated 350 {micro}m 19.7% low enrichment uranium oxide/uranium carbide kernels (LEUCO). The AGR-1 TRISO-coated particles consist of a spherical kernel coated with a {approx} 50% dense carbon buffer layer (100 {micro}m nominal thickness) followed by a dense inner pyrocarbonlayer (40 {micro}m nominal thickness) followed by a SiC layer (35 {micro}m nominal thickness) followed by another dense outer pyrocarbon layer (40 {micro}m nominal thickness). The coated particles, were produced by ORNL for the Advanced Gas Reactor Fuel Development and Qualification (AGR) program to be put into compacts for insertion in the first irradiation test capsule, AGR-1. The kernels were obtained from BWXT and identified as composite (G73D-20-69302). The BWXT kernel lot G73D-20-69302 was riffled into sublots for characterization and coating by ORNL and identified as LEU01-?? (where ?? is a series of integers beginning with 01). Additional particle batches were coated with only buffer or buffer plus inner pyrocarbon (IPyC) layers using similar process conditions as used for the full TRISO batches comprising the LEU01-46T composite. These batches were fabricated in order to qualify that the process conditions used for buffer and IPyC would produce acceptable densities, as described in sections 8 and 9. These qualifying batches used 350 {micro}m natural uranium oxide/uranium carbide kernels (NUCO). The kernels were obtained from BWXT and identified as composite G73B-NU-69300. The use of NUCO surrogate kernels is not expected to significantly effect the densities of the buffer and IPyC coatings. Confirmatory batches using LEUCO kernels from G73D-20-69302 were coated and characterized to verify this assumption. The AGR-1 Fuel Product Specification and Characterization Guidance (INL EDF-4380, Rev. 6) provides the requirements necessary for

  8. AGR-2 irradiation test final as-run report, Rev. 1

    Energy Technology Data Exchange (ETDEWEB)

    Collin, Blaise

    2014-08-01

    This document presents the as-run analysis of the AGR-2 irradiation experiment. AGR-2 is the second of the planned irradiations for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. Funding for this program is provided by the U.S. Department of Energy as part of the Very High Temperature Reactor (VHTR) Technical Development Office (TDO) program. The objectives of the AGR-2 experiment are to: (a) Irradiate UCO (uranium oxycarbide) and UO2 (uranium dioxide) fuel produced in a large coater. Fuel attributes are based on results obtained from the AGR-1 test and other project activities; (b) Provide irradiated fuel samples for post-irradiation experiment (PIE) and safety testing; and, (c) Support the development of an understanding of the relationship between fuel fabrication processes, fuel product properties, and irradiation performance. The primary objective of the test was to irradiate both UCO and UO2 TRISO (tri-structural isotropic) fuel produced from prototypic scale equipment to obtain normal operation and accident condition fuel performance data. The UCO compacts were subjected to a range of burnups and temperatures typical of anticipated prismatic reactor service conditions in three capsules. The test train also includes compacts containing UO2 particles produced independently by the United States, South Africa, and France in three separate capsules. The range of burnups and temperatures in these capsules were typical of anticipated pebble bed reactor service conditions. The results discussed in this report pertain only to U.S. produced fuel. In order to achieve the test objectives, the AGR-2 experiment was irradiated in the B-12 position of the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) for a total irradiation duration of 559.2 effective full power days (EFPD). Irradiation began on June 22, 2010, and ended on October 16, 2013, spanning 12 ATR power cycles and approximately three and a half calendar years. The test

  9. AGR-2 Irradiation Test Final As-Run Report, Rev 2

    Energy Technology Data Exchange (ETDEWEB)

    Blaise Collin

    2014-08-01

    This document presents the as-run analysis of the AGR-2 irradiation experiment. AGR-2 is the second of the planned irradiations for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. Funding for this program is provided by the U.S. Department of Energy as part of the Very High Temperature Reactor (VHTR) Technical Development Office (TDO) program. The objectives of the AGR-2 experiment are to: (a) Irradiate UCO (uranium oxycarbide) and UO2 (uranium dioxide) fuel produced in a large coater. Fuel attributes are based on results obtained from the AGR-1 test and other project activities. (b) Provide irradiated fuel samples for post-irradiation experiment (PIE) and safety testing. (c) Support the development of an understanding of the relationship between fuel fabrication processes, fuel product properties, and irradiation performance. The primary objective of the test was to irradiate both UCO and UO2 TRISO (tri-structural isotropic) fuel produced from prototypic scale equipment to obtain normal operation and accident condition fuel performance data. The UCO compacts were subjected to a range of burnups and temperatures typical of anticipated prismatic reactor service conditions in three capsules. The test train also includes compacts containing UO2 particles produced independently by the United States, South Africa, and France in three separate capsules. The range of burnups and temperatures in these capsules were typical of anticipated pebble bed reactor service conditions. The results discussed in this report pertain only to U.S. produced fuel. In order to achieve the test objectives, the AGR-2 experiment was irradiated in the B-12 position of the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) for a total irradiation duration of 559.2 effective full power days (EFPD). Irradiation began on June 22, 2010, and ended on October 16, 2013, spanning 12 ATR power cycles and approximately three and a half calendar years. The test

  10. 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)

  11. Development and validation of thermal-hydraulic analysis code for plate type fuel research reactors

    International Nuclear Information System (INIS)

    A thermal-hydraulic analysis code has been developed with Visual Fortran 6.5 for the investigation of plate type fuel reactors, based on the fundamental conservation of mass, momentum and energy, and some proper constitutive correlations of flow friction factor, heat transfer and property. The Reactivity Insertion Accident(RIA) and Loss Of Flow Accident (LOFA), which have been defined in the IAEA 10 MW MTR Benchmark transients, were analyzed with this developed program. The comparison of some key parameters, such as the core power at scram, the maximum fuel temperature, the maximum clad temperature and the maximum outlet coolant temperature demonstrated that the results were consistent with that in the literature, which indicated that the model of this developed code is proper for thermal-hydraulic analysis of plate type research reactors. (authors)

  12. 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.

  13. Influence of accident management strategies on source terms of VVER-1000-type reactors

    International Nuclear Information System (INIS)

    The source term can be mitigated by effective accident management. The goal of this work is the investigation of the influence of a number of accident management strategies on the source term of a VVER-1000-type reactor. This work is one of a series of studies investigating the behavior of a VVER-1000-type reactor during severe accidents. In particular, it is based on the study in which the pressure rise in the containment and the melt-through of the cavity bottom was investigated, indicating potential mitigation strategies. To rate the usefulness of these strategies, the source terms of selected scenarios are also calculated in the present work. All the calculations were performed using the Source Term Code Package; hydrogen explosions are not considered. For the first time, the source term behavior of these scenarios was simulated up to the very end of the accident the solidification of the melt

  14. Reduced-size LHD-type fusion reactor with D-shaped magnetic surface

    International Nuclear Information System (INIS)

    A new winding law for the continuous helical coils is proposed for Large Helical Device (LHD) type fusion reactors to satisfy the requirements for a wide blanket space and large plasma volume. Helical coils wound along the geodesic line of a torus with an elongated cross section can produce a magnetic configuration having a D-shaped magnetic surface with a magnetic well in the core region and high magnetic shear in the peripheral regions. The DT alpha particle confinement performance is greatly improved by increasing the elongation factor κ of the cross section of the winding frame for the helical coils. The results suggest that a smaller LHD-type fusion reactor can be realized. (author)

  15. Numerical effects in the neutron flux calculations into WWER type reactors vessels by Monte Carlo Method

    International Nuclear Information System (INIS)

    The calculation of neutron fluxes and fluence into reactor pressure vessel is a regulatory requirement in the stages of the design, operation and plan lifetime extension. Being the reactor vessel a part of the primary circuit, its integrity should be preserved under all operation regimes. The reactor vessel is considered a unique and non-substitutable part of the NPP that undergoes degradation. The main source of the aging comes from the fast neutron damage induced in the steel crystalline lattice. In the case of the WWER-type reactors, the vessel fragilization has been identified as one of the main problems concerning the safety of NPPs. Due to the proximity of the core edge to the vessel inner surface; the vessel steel is exposed to high fast neutron fluence. The effect of this irradiation on the mechanical properties becomes more acute because of the impurities measured in the current Russian steel alloys. In the present paper, a PC version of the Monte Carlo 3-D HEXANN-EVALU system is used for the estimation of the WWER reactor pressure vessel irradiation. It was selected on the basis of its flexible options that on the other hand need to be quantified in connection with the desired magnitudes. The parameters that control the random walk of neutrons as well as the efficiency increasing options included in the code are studied in order to identify their impact in the final results for fluxes and fluence in the reactor pressure vessel. As a result an optimal set of parameters is suggested

  16. 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.

  17. 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

  18. Natural vibrations of a core banel of a PWR type reactor by elements of revolution shell

    International Nuclear Information System (INIS)

    Aim to estimate the behavior of the cove barrel of PWR type reactors, submitted to several load conditions, their dynamic characteristic, were determined. In order to obtain the natural modes and frequencies of the core barrel, the CYLDYFE comprete code based in the finite element method, was developed. The obtained results are compared with results obtained by other programs such as SAP, ASKA and STRUDL/DYNAL and by other analytical methods. (M.C.K.)

  19. 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

  20. The influence of accident measures on accident scenarios for VVER-1000-Type reactors

    International Nuclear Information System (INIS)

    For VVER-1000-type reactors severe accident scenarios and possible mitigation strategies are investigated. The Station blackout sequence is chosen as reference case. At first a comparison between the cases with and without working spray systems is discussed. Afterwards the results of a parametric study investigating the influence of different water volumes on the course of the accident are presented. It can be shown that most of these accident mitigation measures will maintain the containment integrity and reduce the source term. (author)

  1. 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 ...

  2. Reactor noise analysis applications and systems in WWER-440 and WWER-1000 type PWRs

    International Nuclear Information System (INIS)

    This paper presents an introduction on different types of well selected noise diagnostic methods with their occurrence in WWER reactors with an analysis of their impact on operational safety and aging which affects the installations safety as well. The main objective is to attract the attention of NPP management staff dealing with safety, safety culture, maintenance, operation and quality assurance proving that such methods can give benefit not only to economy but impact safety of nuclear installations

  3. Propagation of cracks by stress corrosion in conditions of BWR type reactor

    International Nuclear Information System (INIS)

    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)

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

    OpenAIRE

    Staffan Qvist

    2014-01-01

    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 ...

  5. Synthesis of layered double hydroxide nanosheets by coprecipitation using a T-type microchannel reactor

    Energy Technology Data Exchange (ETDEWEB)

    Pang, Xiujiang; Sun, Meiyu; Ma, Xiuming [State Key Laboratory Base of Eco-chemical Engineering, College of Chemistry and Molecular Engineering, Qingdao University of Science and Technology, Qingdao 266042 (China); Hou, Wanguo, E-mail: wghou@sdu.edu.cn [Key Laboratory of Colloid and Interface Chemistry (Ministry of Education), Shandong University, Jinan 250100 (China)

    2014-02-15

    The synthesis of Mg{sub 2}Al–NO{sub 3} layered double hydroxide (LDH) nanosheets by coprecipitation using a T-type microchannel reactor is reported. Aqueous LDH nanosheet dispersions were obtained. The LDH nanosheets were characterized by X-ray diffraction, transmission electron microscopy, atomic force microscopy and particle size analysis, and the transmittance and viscosity of LDH nanosheet dispersions were examined. The two-dimensional LDH nanosheets consisted of 1–2 brucite-like layers and were stable for ca. 16 h at room temperature. In addition, the co-assembly between LDH nanosheets and dodecyl sulfate (DS) anions was carried out, and a DS intercalated LDH nanohybrid was obtained. To the best of our knowledge, this is the first report of LDH nanosheets being directly prepared in bulk aqueous solution. This simple, cheap method can provide naked LDH nanosheets in high quantities, which can be used as building blocks for functional materials. - Graphical abstract: Layered double hydroxide (LDH) nanosheets were synthesized by coprecipitation using a T-type microchannel reactor, and could be used as basic building blocks for LDH-based functional materials. Display Omitted - Highlights: • LDH nanosheets were synthesized by coprecipitation using a T-type microchannel reactor. • Naked LDH nanosheets were dispersed in aqueous media. • LDH nanosheets can be used as building blocks for functional materials.

  6. Study of advanced nuclear fuel cycles in Candu type power reactors

    International Nuclear Information System (INIS)

    The fuel burn up can be increased to a large extent, up to 14, 0000 MWD/te, by using the slightly enriched uranium or Pu mixed fuel in CANDU type power reactors. In the present study, the previous work was extended to compare the isotopic inventories and corresponding activities of important nuclides for different fuel cycles of a CANDU 600 type power reactor. The detail can be found in our studies. The calculations were performed using the computer code WIMSD4. The isotopic inventories and corresponding activities were calculated versus the fuel burn-up for the natural UO/sub 2/ fuel, 1.2 % enriched UO/sub 2/ fuel and 0.45 % PuO/sub 2/-UO/sub 2/ fuel. It was found that 1.2 % enriched uranium fuel has the lowest activity as compared to other two fuel cycles. It means that improvement in the fuel cycle technology of CANDU type power reactors can lead to high burn up which results in the reduction of actinide content in the spent fuel, and hence has a good environmental impact. (orig./A.B.)

  7. Analysis of the effect of the most dangerous failures of the reactor cooling system on the WWER-1000 type reactor operational safety

    International Nuclear Information System (INIS)

    The WWER-1000 system behaviour during a LOCA coincident with a mechanical failure of the main cooling pump is analysed using the codes RELAP-4/06 and SSIT-2. Data on the changes in the most important thermophysical parameters of the primary coolant cirquit, reactor core and the fuel elements under these conditions are obtained. The high reliability of the fuel elements of WWER-1000 type reactor during the above-mentioned regimes is proven. 6 figs., 4 refs

  8. Cost effective use of software in AGR safety applications

    International Nuclear Information System (INIS)

    Economic pressures on Scottish Nuclear have been significant since its retention in state ownership following the UK's Electrical Supply Industry (ESI) privatisation two years ago, and will continue to increase as other utilities' generation costs become more competitive. Software systems play a significant role in both the maintenance of safety claims and in the output of Scottish Nuclear's Advanced Gas-cooled Reactor (AGR) generating plant. This paper takes as its theme Scottish Nuclear's positive and negative experiences with high reliability software. A cost effective software life cycle for the maintenance of safety related software with unreliability demands in the range 1.0E-1 to 1.0E-2 pfd for reactor control and fuel handling applications based on a customer-software integrator-software supplier model is described. Experience from the evolution of this life cycle has led to the concept of an integrated software maintenance environment. The problems leading to Scottish Nuclear's move away from the use of software for safety critical (1.0E-4 pfd) applications and its substitution by non-processor technology on the fuel route are outlined. Finally the paper describes Scottish Nuclear's medium term policy for the engineering of safety critical applications. (Author) 2 refs., 3 figs

  9. 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)

  10. 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

  11. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Horhoianu, G. [Inst. for Nuclear Research (INR), Pitesti (Romania); Palleck, S. [Atomic Energy of Canada Limited., Mississauga, Ontario (Canada); Ionescu, D. [Inst. for Nuclear Research (INR), Pitesti (Romania)

    2010-07-01

    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)

  13. 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)

  14. Damage mechanism of piping welded joints made from austenitic Steel for the type RBMK reactor

    International Nuclear Information System (INIS)

    In the process of operation of RBMK reactors the damages were taking place on welded piping, produced from austenitic stainless steel of the type 08X18H10T. The inspection of damaged sections in piping has shown that in most cases crack-like defects are of corrosion and mechanical character. The paper considers in details the reasons of damages appearance and their development for this type of welded joints of downcomers 325xl6 mm, which were fabricated from austenitic stainless steel using TlG and MAW welding methods. (author)

  15. 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)

  16. Radiological consequences of a postulated cooling channel blockage incident at a pool-type research reactor

    International Nuclear Information System (INIS)

    An assessment of the radiological consequences of a postulated coolant flow blockage incident at the Hoger Onderwijs Reactor (HOR) is being presented. The HOR is a swimming-pool type research reactor with a maximum licensed power of 3 MW. Assuming a sudden blockage of cooling channels in the high power density region of the core, the source term for the release of radioactivity into the environment was calculated. The magnitude of this source term is required by actions of the HOR protection system as well as by physical processes acting on the fission products. Hence, almost 99% of the calculated release from the containment consists of noble gases; most of the aerosol-type activity set free in the environment results from the decay of these noble gases. The deposit of long-living radionuclides outside the reactor building is very low. Radio-iodine will be the main contributor to the environmental radiation dose, ingestion of contaminated food being the critical pathway. Despite the conservativeness of most assumptions used, the calculated thyroid dose for critical individuals at all distances from the site boundary remains well below the emergency reference levels recommended by national and international organizations and the national dose limit for members of the public

  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. Prediction of neutron energy spectrum in a typical MTR type research reactor using Monte Carlo simulations

    International Nuclear Information System (INIS)

    Highlights: ► The core of a typical MTR type research reactor was modelled, in three dimensions, using the Monte Carlo code, MCNP5. ► The neutron energy spectrum at the central irradiation site was determined. ► Results are compared with earlier calculations performed by the deterministic code CITATION. - Abstract: In previous work, determination of the neutron energy spectrum in a typical pool type Material Test research Reactor (MTR) was discussed. Solution of the neutron spectrum adjustment problem, which adjusts a theoretically calculated spectrum to a set of experimentally measured reaction rates, was also analyzed. The calculated spectrum was obtained through modelling the reactor core and the surroundings in three dimensions using the deterministic code CITATION. In this work, the same core configuration was modelled in three dimensions using the Monte Carlo code, MCNP5. The calculated spectrum by MCNP is compared to that calculated by CITATION. Both calculated spectra by CITATION and MCNP were also compared as input information in the experimental determination of the neutron spectrum through the use of experimentally measured reaction rates and the adjustment code MSITER. The good agreement between the calculated and adjusted spectra indicates that the MCNP approach can be used as pre-information in the experimental determination of the neutron spectrum as well as for the prediction of neutron spectrum at other locations

  19. 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

  20. 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)

  1. The use of the codes from MCU family for calculations of WWER type reactors

    International Nuclear Information System (INIS)

    The MCU-RFFI/A and MCU-REA codes developed within the framework of the long term MCU project are widely used for calculations of neutron physic characteristics of WWER type reactors. Complete descriptions of the codes are available in both Russian and English. The codes are verified and validated by means of the comparison of calculated results with experimental data and mathematical benchmarks. The codes are licensed by Russian Nuclear and Criticality Safety Regulatory Body (Gosatomnadzor RF) (Code Passports: N 61 of 17.10.1966 and N 115 of 02.03.2000 accordingly)). The report gives examples of WWER reactor physic tasks important for practice solved using the codes from the MCU family. Some calculational results are given too. (Authors)

  2. 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)

  3. 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...

  4. Development of Romanian SEU-43 fuel bundle for CANDU type reactors

    International Nuclear Information System (INIS)

    SEU-43 fuel bundle is a CANDU type fuel consisting of two element sizes, to reduce element ratings, while maintaining the same bundle power, and an uranium content very close to the uranium content of a standard 37-element bundle. In order to reduce the detrimental effects of the life limiting factors at extended burnup a set of solution have been adopted for fuel element design. As a part of the design verification program, experimental bundles have been fabricated and utilized in typical out of reactor tests conducted at the laboratories of INR, Pitesti. These tests simulated current CANDU-6 reactor normal operating conditions of flow, temperature and pressure. The results are in accordance with the specified acceptance criteria. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Horhoianu, G.; Ionescu, D.; Olteanu, G. [Inst. for Nuclear Research, Pitesti (Romania)

    2008-07-01

    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{sup -1} UO{sub 2}. 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)

  6. Reactivity initiated accident (RIA) type tests and annular core pulse reactor (ACPR) operational experience

    International Nuclear Information System (INIS)

    This paper describes the test conducted to investigate the failure threshold of the fuel when subject to RIA, accomplished in the TRIGA ACPR Nuclear Research Institute, Pitesti. The reactor facility, the capsule used in experiments and the experimental results are presented. The failure threshold was determined at 200 cal/g for an atmospheric gap pressure comparable with similar tests. The failure threshold decreases with increasing gap pressure. The tests proved useful for a better understanding of the fuel behavior in the transient conditions. As it is known RIA is not a common accident for the CANDU reactors, but the fuel failure mechanism can be similar to other type of accidents as LOCA and PCM. The program will be continued, with better instrumentation for the fuel sample and also independent instrumentation to measure pulse characteristics with better statistics. A new project for the experimental fuel elements must be considered to eliminate fuel-endcap interactions. (author)

  7. Strain components of nuclear-reactor-type concretes during first heat cycle

    International Nuclear Information System (INIS)

    Strains of three advanced-gas-cooled-reactor-type nuclear reactor concretes were measured during the first heat cycle and their relative thermal stability determined. It was possible to isolate for the first time the shrinkage component for the period during heating. Predictions of the residual strains for the loaded specimens can be made by simple superposition of creep and shrinkage components up to a certain critical temperature, which for basalt concrete is about 500 C and for limestone concrete is about 200-300 C. Above the critical temperature, an expansive ''cracking'' strain component is present. It is shown that the strain behaviour of concrete provides a sensitive indication of its thermal stability during heating and subsequent cooling. (orig.)

  8. Study of experiments with complex reactor core consisting of two types of fuel elements

    International Nuclear Information System (INIS)

    Criticality parameters and neutron flux distributions were measured at the heavy water RB reactor with the mixed reactor core. Cylindrical natural uranium fuel and tubular two-percent enriched metal fuel elements were used. Results were obtained under assumption of simple volume homogenization and procedure based on two-zone supercell estimation. Theoretical results obtained by homogenization are in good agreement with the experimental results, especially in case of criticality parameters of the system. This conclusion could not be generalized, because there is no significant difference in micro-distribution of neutron flux due to small difference in total thermal neutron absorption in each type of fuel elements, although the differences in multiplication and macroscopic cross sections are rather high. The obtained results are promising for studying the two-zone cell model in case of mixed core with significant mutual interference of different fuel elements, as well as for further improvement of this model

  9. Physical concept on the nuclear reactor with constellation type fissile fuels

    International Nuclear Information System (INIS)

    Under a prolong time neutron irradiation of 232Th by a strong constant neutron flux, a part of 232Th atoms will be converted into a series other nuclides as a result of successive neutron interaction. These neutron born daughter nuclides of 232Th consist of fissile nuclides such as 233U, 235U, 239Pu, 241Pu, ... and fertile nuclides, such as 233Pa, 2'34U, 236U, 237Np, 238Pu etc. As a rule the concentration of each daughter nuclide starts from zero and then increases gradually as the irradiation proceeds until reaching a saturated value and then decreases at a similar rate as the decreasing rate of 232Th. A reactor fueled with thorium and its whole family of neutron born daughter nuclides, with each daughter nuclide at its own saturated concentration, may possess some interesting properties. A preliminary study of the feasibility of such constellation type fissile fuels reactor is presented

  10. Optimized Core Design and Fuel Management of a Pebble-Bed Type Nuclear Reactor

    International Nuclear Information System (INIS)

    The Very High Temperature Reactor (VHTR) has been selected by the international Generation IV research initiative as one of the six most promising nuclear reactor concepts that are expected to enter service in the second half of the 21st century. The VHTR is characterized by a high plant efficiency and a high fuel discharge burnup level. More specifically, the (pebble-bed type) High Temperature Reactor (HTR) is known for its inherently safe characteristics, coming from a negative temperature reactivity feedback, a low power density and a large thermal inertia of the core. The core of a pebble-bed reactor consists of graphite spheres (pebbles) that form a randomly packed porous bed, which is cooled by high pressure helium. The pebbles contain thousands of fuel particles, which are coated with several pyrocarbon and silicon carbon layers that are designed to contain the fission products that are formed during operation of the reactor. The inherent safety concept has been demonstrated in small pebble-bed reactors in practice, but an increase in the reactor size and power is required for cost-effective power production. An increase of the power density in order to increase the helium coolant outlet temperature is attractive with regard to the efficiency and possible process heat applications. However, this increase leads in general to higher fuel temperatures, which could lead to a consequent increase of the fuel coating failure probability. This thesis deals with the pebble-bed type VHTR that aims at an increased coolant outlet temperature of 1000 degrees C and beyond. For the simulation of the neutronic and thermal-hydraulic behavior of the reactor the DALTON-THERMIX coupled code system has been developed and has been validated against experiments performed in the AVR and HTR-10 reactors. An analysis of the 400 MWth Pebble Bed Modular Reactor (PBMR) design shows that the inherent safety concept that has been demonstrated in practice in the smaller AVR and HTR-10

  11. Uncertainty Quantification of Calculated Temperatures for the U.S. Capsules in the AGR-2 Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Lybeck, Nancy [Idaho National Lab. (INL), Idaho Falls, ID (United States); Einerson, Jeffrey J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Pham, Binh T. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hawkes, Grant L. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-03-01

    A series of Advanced Gas Reactor (AGR) irradiation experiments are being conducted within the Advanced Reactor Technology (ART) Fuel Development and Qualification Program. The main objectives of the fuel experimental campaign are to provide the necessary data on fuel performance to support fuel process development, qualify a fuel design and fabrication process for normal operation and accident conditions, and support development and validation of fuel performance and fission product transport models and codes (PLN-3636). The AGR-2 test was inserted in the B-12 position in the Advanced Test Reactor (ATR) core at Idaho National Laboratory (INL) in June 2010 and successfully completed irradiation in October 2013, resulting in irradiation of the TRISO fuel for 559.2 effective full power days (EFPDs) during approximately 3.3 calendar years. The AGR-2 data, including the irradiation data and calculated results, were qualified and stored in the Nuclear Data Management and Analysis System (NDMAS) (Pham and Einerson 2014). To support the U.S. TRISO fuel performance assessment and to provide data for validation of fuel performance and fission product transport models and codes, the daily as-run thermal analysis has been performed separately on each of four AGR-2 U.S. capsules for the entire irradiation as discussed in (Hawkes 2014). The ABAQUS code’s finite element-based thermal model predicts the daily average volume-average fuel temperature and peak fuel temperature in each capsule. This thermal model involves complex physical mechanisms (e.g., graphite holder and fuel compact shrinkage) and properties (e.g., conductivity and density). Therefore, the thermal model predictions are affected by uncertainty in input parameters and by incomplete knowledge of the underlying physics leading to modeling assumptions. Therefore, alongside with the deterministic predictions from a set of input thermal conditions, information about prediction uncertainty is instrumental for the ART

  12. Supercritical Carbon Dioxide Brayton Power Conversion Cycle Design for Optimized Battery-Type Integral Reactor System

    International Nuclear Information System (INIS)

    Supercritical carbon dioxide (SCO2) promises a high power conversion efficiency of the recompression Brayton cycle due to its excellent compressibility reducing the compression work at the bottom of the cycle and to a higher density than helium or steam decreasing the component size. Therefore, the high SCO2 Brayton cycle efficiency as high as 45 % furnishes small sized nuclear reactors with economical benefits on the plant construction and maintenance. A 23 MWth BORIS (Battery Optimized Reactor Integral System) is being developed as a multipurpose reactor. BORIS, an integral-type optimized fast reactor with an ultra long life core, is coupled to the SCO2 Brayton cycle needing less room relative to the Rankine steam cycle because of its smaller components. The SCO2 Brayton cycle of BORIS consists of a 16 MW turbine, a 32 MW high temperature recuperator, a 14 MW low temperature recuperator, an 11 MW pre-cooler and 2 and 2.8 MW compressors. Entering six heat exchangers between primary and secondary system at 19.9 MPa and 663 K, the SCO2 leaves the heat exchangers at 19.9 MPa and 823 K. The promising secondary system efficiency of 45 % was calculated by a theoretical method in which the main parameters include pressure, temperature, heater power, the turbine's, recuperators' and compressors' efficiencies, and the flow split ratio of SCO2 going out from the low temperature recuperator. Test loop SOLOS (Shell-and-tube Overall Layout Optimization Study) is utilized to develop advanced techniques needed to adopt the shell-and-tube type heat exchanger in the secondary loop of BORIS by studying the SCO2 behavior from both thermal and hydrodynamic points of view. Concurrently, a computational fluid dynamics (CFD) code analysis is being conducted to develop an optimal analytical method of the SCO2 turbine efficiency having the parameters of flow characteristics of SCO2 passing through buckets of the turbine. These simultaneous experimental and analytical methods for designing

  13. Feasibility studies for LEU conversion of the WWR-SM reactor in Uzbekistan using pin-type and tubular fuels

    International Nuclear Information System (INIS)

    The 10 MW WWR-SM research reactor in Uzbekistan currently uses HEU (36%) IRT-3M 6-tube fuel assemblies manufactured by the Novosibirsk Chemical Concentrates Plant in Russia. This study compares the neutronic performance and preliminary thermal-hydraulic performance of the reactor and its experiments for several core sizes using various LEU pin-type and LEU tube-type fuel assembly designs with the performance of the current HEU (36%) reference fuel assembly and core. Several LEU fuel assembly designs are identified which would be suitable for conversion of the WWR-SM reactor if they are manufactured, successfully irradiation tested, and made commercially available in Russia. (author)

  14. 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)

  15. An improved assembly homogenization approach for plate-type research reactor

    International Nuclear Information System (INIS)

    Highlights: • An improved assembly homogenization approach is developed for plate-type research reactor. • The approach includes direct assembly calculation and SPH equivalence method. • The assembly environmental effect is considered with multi-assembly calculations and assembly classification. • The control rod worth can be calculated accurately by the improved approach. - Abstract: We have developed an improved assembly homogenization approach for plate-type research reactor. Compared with the conventional cell and assembly homogenization way, three advanced homogenization methods are used in our improved approach, including the direct single and multi-assembly calculations, treatment of the assembly environmental effect by multi-assembly calculations and assembly classification, and the superhomogenization (SPH) method. The approach is carried out with DRAGON v4 code. The neutronic analysis for fresh fueled core of JRR-3M with UAlx–Al dispersion type fuel is carried out by diffusion code CITATION with 3-group homogenized cross sections which are generated by the improved approach. Results by Monte Carlo code are taken as references, and results by the conventional three-step method are given for comparison purposes. The calculation results show that the improved assembly homogenization approach is effective to improve the accuracy of keff eigenvalue, control rod worth and neutron flux distributions for whole core calculations. Especially, the control rod worth can be calculated accurately by the improved approach

  16. 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.)

  17. Pressure drop calculation in a fuel element of a pool type reactor

    International Nuclear Information System (INIS)

    Even with the advances of hardware in computer sciences, sometimes it is necessary to simplify the simulation in order to optimize the results given the same calculation runtime. The object of this study is a thermodynamic analysis of the core of a pool type research reactor, focusing on natural circulation. Due to the high geometrical complexity of the core, the scale transfer process becomes an essential step to the thermodynamic study of the reactor. This process takes place by determining the effective equivalent properties obtained from a detailed simulation of the core and transferring them to a porous medium having a coarse mesh while preserving the overall characteristics. In this way, it will be able to obtain the quadratic resistance coefficient KQ by calculating the pressure drop inside the fuel element. To observe in detail the behavior of this flow, longitudinal and transversal cross sections will be made in different points, thereby observing the velocity and pressure distributions. The analysis will provide detailed data on the fluid flow between the fuel plates enabling the observation of possible critical points or undesired behavior. The whole analysis was made by using the commercial code ANSYS CFX ver. 12.1. This is study will provide data, as a first step to enable future simulations which will consider the entire reactor. (author)

  18. Organization and mechanization of maintenance operations at NPPs with the WWER type reactors

    International Nuclear Information System (INIS)

    The structure of capital investments defining organization and mechanization of maintepance operations at NPPs with the WWER type reactors is analyzed. The trends in development of optimum decisions for organization and mechanization of repair obs at NPPs being designed taking into account the prospects of nuclear powep enginerning development, the system of NPP maintenance servicing, as well as the structure of repair-productive capacities are discussed. On the basis of the analysis of the data obtained in designing the Zaporozhskaya NPP it is shown that the capital investments for organizing and mechanization of maintenance operations at the unified NPP site with four WWER-1000 reactors reach nearly 18 roubles/kW. A conclusion is drawn that at present the design of an NPP with the WWER-1000 reactor totally meets the requirements of realization of periodic maintenance operations. It is advisable to cooperate the NPP management with that of a thermal power station from the viewpoint of using manpower, which would improve the operating conditions and labour productivity of workers engaged in repair and, consequently, reduce the capital investments and repair expenditures

  19. Spatial fluxes and energy distributions of reactor fast neutrons in two types of heat resistant concretes

    International Nuclear Information System (INIS)

    Measurements have been carried out to study the spatial fluxes and energy distributions of reactor fast neutrons transmitted through two types of heat resistant concretes, serpentine concrete and magnetic lemonite concrete. The physical, chemical and mechanical properties of these concretes were checked by well known techniques. In addition, the effect of heating at temperatures up to 500deg C on the crystaline water content was checked by the method of differential thermal analysis. Measurements were performed using a collimated beam of reactor neutrons emitted from a 10 MW research reactor. The neutron spectra transmitted through concrete barriers of different thickness were measured by a scintillation spectrometer with NE-213 liquid organic scintillator. Discrimination against undesired pulses due to gamma-rays was achieved by a method based on pulse shape discrimination technique. The operating principle of this technique is based on the comparison of two weighted time integrals of the detector signal. The measured pulse amplitude distribution was converted to neutron energy distribution by a computational code based on double differentiation technique. The spectrometer workability and the accuracy of the unfolding technique were checked by measuring the neutron spectra of neutrons from Pu-α-Be and 252Cf neutron sources. The obtained neutron spectra for the two concretes were used to derive the total cross sections for neutrons of different energies. (orig.)

  20. CERREX Software Applied to a Research Reactor of the Generic Argonaut Type. Appendix IV

    International Nuclear Information System (INIS)

    This appendix presents the results of a cost calculation for the dismantling of a research reactor of the generic ARGONAUT type, performed using CERREX, and compares the results with a cost estimate obtained by the facility owners. The ARGONAUT reactor was first developed at the Argonne National Laboratories, USA. Typical roles and the main features of the ARGONAUT included: Participation in training activities in neutron flow and neutron activation analysis; Support for public research of irradiation of materials, implementation of analytical techniques using a neutron source and production of radioisotopes; Provision of similar services to companies under a contractual framework; A maximum power of 100 kW, with power typically between 1 and 100 W used for teaching purposes and up to 100 kW for irradiation. The reactor was dismantled between 2006 and 2008. The duration of the decommissioning site works estimated by the facility owners was 13 months, assuming 21 working days per month and 8 working hours per day (8 x 21 x 12 = 2016 working hours per year, see cell V2 in the 'ISDC' tab). Furthermore, the owners assumed six workers to be present on-site at any given time

  1. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Iqbal, Masood [Nuclear Engineering Division, PINSTECH, P.O. Nilore, Islamabad (Pakistan)], E-mail: masiqbal@hotmail.com; Mahmood, Tayyab; Pervez, Showket [Nuclear Engineering Division, PINSTECH, P.O. Nilore, Islamabad (Pakistan)

    2008-03-15

    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 {beta}{sub eff} decreases with 1.33 x 10{sup -6}/% burnup whereas prompt neutron generation time increases with 6.42 x 10{sup -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.

  3. 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

  4. Experience feedback examination in PWR type reactors operating for the 1997-1999 period

    International Nuclear Information System (INIS)

    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.)

  5. Analysis of the integrity of the pressure vessel of the BWR type nuclear reactor

    International Nuclear Information System (INIS)

    The presssure vessel of a BWR type reactor was monitored for cracking during alternating events of its in-service life. The monitoring was to determine criticality of fractures catastrophic fractures and the velocity of fracture propagation. Detected cracks were evaluated as specified in ASME code section XI, of a minimum wall thickness of 2.5% crack growths were compared a) of 1/10 of the critical maximum size and b) at in-service inspection intervals according to ASME recommendations to be established at the Laguna Verde nuclear plant. Finally conclusions are made and discussed. (author)

  6. Investigation on mass flux distribution and asymmetrical cooling in a plate-type fuel reactor

    International Nuclear Information System (INIS)

    A program was developed to calculate the mass flux distribution in the whole core and the asymmetrical cooling of fuel pins for a plate type fuel reactor by applying the proper physical model. Three iterative algorithms for mass flux distribution calculation and two iterative algorithms for temperature field calculation of plate fuel element under asymmetrical cooling condition were proposed and compared by applying in a subassembly. The results showed that the flow distribution is mainly determined by the core structure, although it is also impacted by the power distribution in the core. The asymmetrical cooling condition seriously impacts the temperature field and power distribution of the fuel pins. (authors)

  7. 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)

  8. Quantitative analysis of gases in fuels. Applications to PWR type reactors fuels

    International Nuclear Information System (INIS)

    The different methods used in Saclay and Cadarache to determine the quantity of gases which are present in fuels and fuel cans of PWR type reactors are described. These gases are fission gases (Xe, Kr), pollutant gases (hydrocarbons, N2, O2, H2O, CO, CO2), filling gases (He) or hydrides. A description of the equipment and the operation mode used are given. The obtained results on uranium oxides and mixed oxide fuels are compared with the measures of gases released in the whole rod. (O.M.)

  9. ACT-1000. Group activation cross-section library for WWER-1000 type reactors

    International Nuclear Information System (INIS)

    The ACT-1000, a problem-oriented library of group-averaged activation cross-sections for WWER-1000 type reactors, is based on evaluated microscopic cross-section data files. The ACT-1000 data library was designed for calculating induced activity for the main dose-generated nuclides contained in WWER-1000 structural materials. In preparing the ACT-1000 library, 47 group-averaged cross-section data for the 10-9-17.33 MeV energy range were used to calculate the spatial-energy neutron flux distribution. (author)

  10. LHD type proton-boron reactor and the control of its peripheral potential structure

    International Nuclear Information System (INIS)

    An advanced Large Helical Device (LHD) type proton-boron reactor, in which the minority protons are heated by ICRF, is proposed. The ratio of the fusion power to the RF input power is evaluated. Numerical computation of particle orbits shows that the ICRF of LHD can accelerate protons in the p-11 B fusion relevant energy. Numerical results also show that the LHD magnetic configuration can confine the high energy 4He well. An active peripheral potential control method and an active 4He ash exhaust scheme are discussed. (author)

  11. 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)

  12. 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

  13. 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

  14. Improvement of nuclear ship engineering simulation system. Hardware renewal and interface improvement of the integral type reactor

    International Nuclear Information System (INIS)

    JAERI had carried out the design study about a lightweight and compact integral type reactor (an advanced marine reactor) with passive safety equipment as a power source for the future nuclear ships, and completed an engineering design. We have developed the simulator for the integral type reactor to confirm the design and operation performance and to utilize the study of automation of the reactor operation. The simulator can be used also for future research and development of a compact reactor. However, the improvement in a performance of hardware and a human machine interface of software of the simulator were needed for future research and development. Therefore, renewal of hardware and improvement of software have been conducted. The operability of the integral-reactor simulator has been improved. Furthermore, this improvement with the hardware and software on the market brought about better versatility, maintainability, extendibility and transfer of the system. This report mainly focuses on contents of the enhancement in a human machine interface, and describes hardware renewal and the interface improvement of the integral type reactor simulator. (author)

  15. Identification of the agr Peptide of Listeria monocytogenes

    Science.gov (United States)

    Zetzmann, Marion; Sánchez-Kopper, Andrés; Waidmann, Mark S.; Blombach, Bastian; Riedel, Christian U.

    2016-01-01

    Listeria monocytogenes (Lm) is an important food-borne human pathogen that is able to strive under a wide range of environmental conditions. Its accessory gene regulator (agr) system was shown to impact on biofilm formation and virulence and has been proposed as one of the regulatory mechanisms involved in adaptation to these changing environments. The Lm agr operon is homologous to the Staphylococcus aureus system, which includes an agrD-encoded autoinducing peptide that stimulates expression of the agr genes via the AgrCA two-component system and is required for regulation of target genes. The aim of the present study was to identify the native autoinducing peptide (AIP) of Lm using a luciferase reporter system in wildtype and agrD deficient strains, rational design of synthetic peptides and mass spectrometry. Upon deletion of agrD, luciferase reporter activity driven by the PII promoter of the agr operon was completely abolished and this defect was restored by co-cultivation of the agrD-negative reporter strain with a producer strain. Based on the sequence and structures of known AIPs of other organisms, a set of potential Lm AIPs was designed and tested for PII-activation. This led to the identification of a cyclic pentapeptide that was able to induce PII-driven luciferase reporter activity and restore defective invasion of the agrD deletion mutant into Caco-2 cells. Analysis of supernatants of a recombinant Escherichia coli strain expressing AgrBD identified a peptide identical in mass and charge to the cyclic pentapeptide. The Lm agr system is specific for this pentapeptide since the AIP of Lactobacillus plantarum, which also is a pentapeptide yet with different amino acid sequence, did not induce PII activity. In summary, the presented results provide further evidence for the hypothesis that the agrD gene of Lm encodes a secreted AIP responsible for autoregulation of the agr system of Lm. Additionally, the structure of the native Lm AIP was identified.

  16. Determination of break size based on pressurizer water level in VVER-440 type reactors

    Energy Technology Data Exchange (ETDEWEB)

    Petofi, G.; Horvath, K. [Hungarian Atomic Energy Authority, Badapest (Hungary)

    2001-07-01

    The correct estimation of break size is an important task during the diagnosis phase of a severe accident. The evaluation of the total mass of water in the primary circuit, the estimation of the water level in the reactor vessel, and of the time of core uncover are all based on the supposed break size influencing the primary (and secondary) pressure and temperature and the primary inlet and outlet flow rates. Estimation of the water level is of primary importance, because the reactor-vessel level measurement is not available in the VVER-440 type reactors (except at the Loviisa NPP Finland). The break size estimation can be based either on the change of the pressurizer water level or the change of the primary pressure by taking into account the core inlet and outlet flow rates. In addition the Hungarian Atomic Energy Research Institute has developed a more sophisticated model. Usually at the early phase of an accidental situation the thermo-hydraulic experts have not enough time and data to make such estimations, thus the first diagnosis of the situation has to be based on simpler methods. Such a method can be based on a comparison of some significant parameters to experimental data. For this reason the authors of this article have tried to find some correlation between the change of the pressurizer water level (the time while the pressurizer becomes empty) and the size of the break. The analyses were made by running scenarios on the full scope simulator of Paks NPP and on the basis of MELCOR and ADAM simulation results. Two types of scenarios are assessed in this paper: the LOCA (Loss Of Coolant Accident) and the PRISE (PRImary leakage to Secondary circuit) cases. (author)

  17. 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

  18. A novel reactor type for autothermal reforming of diesel fuel and kerosene

    International Nuclear Information System (INIS)

    Highlights: • Development and experimental evaluation of Juelich’s novel ATR reactor type. • Constructive integration of steam generation chamber and nozzle for water injection. • Internal steam generator modified to reduce pressure drop to approx. a thirtieth. • Novel concept for ATR heat management proven to be suitable for fuel cell systems. • Reaction conditions during shut-down and start-up optimized to reduce byproducts. - Abstract: This paper describes the development and experimental evaluation of Juelich’s novel reactor type ATR AH2 for autothermal reforming of diesel fuel and kerosene. ATR AH2 overcomes the disadvantages of Juelich’s former reactor generations from the perspective of the fuel cell system by constructively integrating an additional pressure swirl nozzle for the injection of cold water and a steam generation chamber. As a consequence, ATR AH2 eliminates the need for external process configurations for steam supply. Additionally, the internal steam generator has been modified by increasing its cross-sectional area and by decreasing its length. This measure reduces the pressure drop of the steam generator from approx. 500 mbar to roughly a thirtieth. The experimental evaluation of ATR AH2 at steady state revealed that the novel concept for heat management applied in ATR AH2 is suitable for fuel cell systems at any reformer load point between 20% and 120% when the mass fractions of cold water to the newly integrated nozzle are set to values between 40% and 50%. The experimental evaluation of ATR AH2 during start-up and shut-down showed that slight modifications of the reaction conditions during these transient phases greatly reduced the concentrations of ethene, ethane, propene and benzene in the reformate. From the fuel cell system perspective, these improvements provide a very beneficial contribution to longer stabilities for the catalysts and adsorption materials

  19. 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

  20. Reactors

    International Nuclear Information System (INIS)

    Purpose: To provide a spray cooling structure wherein the steam phase in a bwr reactor vessel can sufficiently be cooled and the upper cap and flanges in the vessel can be cooled rapidly which kept from direct contaction with cold water. Constitution: An apertured shielding is provided in parallel spaced apart from the inner wall surface at the upper portion of a reactor vessel equipped with a spray nozzle, and the lower end of the shielding and the inner wall of the vessel are closed to each other so as to store the cooling water. Upon spray cooling, cooling water jetting out from the nozzle cools the vapor phase in the vessel and then hits against the shielding. Then the cooling water mostly falls as it is, while partially enters through the apertures to the back of the shielding plate, abuts against stoppers and falls down. The stoppers are formed in an inverted L shape so that the spray water may not in direct contaction with the inner wall of the vessel. (Horiuchi, T.)

  1. Nuclear knowledge preservation of Atucha type Reactor: Practical approaches and lessons learned

    International Nuclear Information System (INIS)

    Full text: The nuclear option is born in Argentina in 1950. Since then and after half century, the Argentine nuclear sector reaches a hierarchy and an important technological - scientific dimension in all the issues related to activities linked to the nuclear sector. The export of research reactors, the nuclear fuel production and the installation of nucleoelectrical power stations constitute one of the most outstanding achievements of this activity. The Atucha I Nuclear Power Plant, of Siemens-KWU technology with 30 years of operation and Embalse (Candu type) with 20 years of operation, register an efficient and safe performance, within the worldwide standards. In addition, and with Siemens technology the Atucha II NPP is in construction with a 82% of work advance. At the moment Siemens, has transferred his place as designer of reactors LWR to the Framatom ANP Company and consequently Argentina will have to make all the efforts to assure the 'knowledge preservation' of the technology of this line of reactors, if it hopes that Atucha I extends its useful life and Atucha II is finalized. Likewise the aging of the personnel in the sector is worrisome being numerous the great number of people who are close to retire what it worsens if we think that there are a few young people able in this specialty by the lack of the entrance of candidates in compatible races This situation generated the necessity to implement a system of Knowledge Management to capture and to capitalize the tacit and explicit knowledge, to spread it and to share by using techniques and tools adapted through the organization. Practical Approaches and Lessons Learned: A - Analysis of the strategy: The strategy was based on recognizing the critical knowledge by means of the use of a methodology that incorporates the technique of the map of the critical knowledge. That is to say, to recognize the patrimony of the knowledge around the reactor. The layout of the knowledge map of reactor together with the

  2. 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.

  3. Containment filter system for WWER type reactors in case of beyond design basis accident

    International Nuclear Information System (INIS)

    The papers presents a scheme of a filter design for WWER type reactors for the releasing of the pressurized vapour-air mixture from the containment and filtering of the radioactive products (fine particles and aerosols) in case of a beyond design basis accident,. The first main component of the filter is a bubbling type mixer where a condensations of the vapour component takes part and the non-condensed (air ) component has been filtered. Experiment have been performed for various values of the weight ratio between the steam and air components, corresponding to the conditions in case of a severe accident with a leakage in the containment. The tested bubbling mixer has 80 cm2 effective flow section. The results show a full vapour component condensation in case of an effective phase mixing. Analyses, made for the WWER-1000 reactors of the Kozloduy NPP, show that the expected vapour consumption can be assured by using of 1-3 parallel units. The second element of this unit - separator - consists of two elements - a demister, which presents a patch of metal wool separated by a metal net with a large effective section, and a droplets turbo-separator. The combination of a demister and turbo-separator can ensure up to 99.5% cleaning. The turbo-separator is tested is introduced in production with very good results

  4. 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.)

  5. Using the CompGen code for developing new MMS modules specific to CANDU type reactors

    International Nuclear Information System (INIS)

    The Modular Modeling System, MMS, is an advanced tool for thermal hydraulic analyses. The MMS code is built upon a modular philosophy allowing users to construct their own schemes from the existing MMS modules. Also, it is possible to develop new MMS modules by using the CompGen sub-code. Developing new modules in the frame of MMS program was necessary because the standard modules of the program libraries are customized to the American reactors, particularly to PWR type reactors. As known the Cernavoda NPP is of PHWR type having equipment and components different from the structure point of view, hence differ from the point of view of the equation describing the transient process. An example is the module of low pressure part of the turbine in the secondary circuit of the plant, where the moisture extraction coefficients measured from equipment operation are different from the coefficient introduced in the module of the MMS library. The new modules developed in this work were compared with data from the thermal balance supplied by General Electric, with functional data from operation, and with the data from the Commissioning Report of the Cernavoda NPP Unit 1 commissioning. In this work new modules which or developed are presented as well as two examples of using the steam generator modules, high pressure turbine, low pressure turbine, and superheater separator

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

    Science.gov (United States)

    Jolibois, J.; Takashima, K.; Mizuno, A.

    2011-06-01

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Horhoianu, Grigore [Institute for Nuclear Research, Pitesti (Romania). Nuclear Fuel Engineering Lab.; Sorescu, Ion [Institute for Nuclear Research, Pitesti (Romania). TRIGA Reactor Loop Facility; Parvan, Marcel [Institute for Nuclear Research, Pitesti (Romania). Hot Cells Lab.

    2012-12-15

    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 UO{sub 2} 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.)

  8. 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.

  9. New inverted hydride fuel design concept for pressure tube type super critical water reactors

    International Nuclear Information System (INIS)

    In this study, an innovative core design having inverted configuration has been proposed for pressure tube type supercritical water reactors. In this design the relative positions of fuel and coolant have been inverted and U-Th-Zr-hydride fuel has been used. A coupled neutronics and thermal hydraulics analysis was done for the proposed Inverted Pressure Tube Type (IPTT) SCWR. The neutronics analysis was carried out by using a 3D fine mesh diffusion theory code and thermal hydraulics calculations were done by using single channel model. These two codes were coupled with each other by a link code. The average outlet temperature for the proposed IPTT-SCWR was found to be 625degC with maximum clad surface temperature (MCST) under the design limits i.e. below 850degC. Moreover a core loading pattern has also been proposed to achieve uniform radial power distribution and lower cladding surface temperature. (author)

  10. Improving thermal model prediction through statistical analysis of irradiation and post-irradiation data from AGR experiments

    International Nuclear Information System (INIS)

    As part of the High Temperature Reactors (HTR) R and D program, a series of irradiation tests, designated as Advanced Gas-cooled Reactor (AGR), have been defined to support development and qualification of fuel design, fabrication process, and fuel performance under normal operation and accident conditions. The AGR tests employ fuel compacts placed in a graphite cylinder shrouded by a steel capsule and instrumented with thermocouples (TC) embedded in graphite blocks enabling temperature control. While not possible to obtain by direct measurements in the tests, crucial fuel conditions (e.g., temperature, neutron fast fluence, and burnup) are calculated using core physics and thermal modeling codes. This paper is focused on AGR test fuel temperature predicted by the ABAQUS code's finite element-based thermal models. The work follows up on a previous study, in which several statistical analysis methods were adapted, implemented in the NGNP Data Management and Analysis System (NDMAS), and applied for qualification of AGR-1 thermocouple data. Abnormal trends in measured data revealed by the statistical analysis are traced to either measuring instrument deterioration or physical mechanisms in capsules that may have shifted the system thermal response. The main thrust of this work is to exploit the variety of data obtained in irradiation and post-irradiation examination (PIE) for assessment of modeling assumptions. As an example, the uneven reduction of the control gas gap in Capsule 5 found in the capsule metrology measurements in PIE helps identify mechanisms other than TC drift causing the decrease in TC readings. This suggests a more physics-based modification of the thermal model that leads to a better fit with experimental data, thus reducing model uncertainty and increasing confidence in the calculated fuel temperatures of the AGR-1 test

  11. A novel extension of chord length sampling method for TRISO-type fueled reactor applications

    International Nuclear Information System (INIS)

    Highlights: • A new chord length distribution model is derived and developed for CLS methods. • The new model is specifically for MC analyses of reactors with TRISO-type fuel. • Fuel kernels can be sampled on the fly using the new model in MC simulations. • The model is verified to be accurate in analyzing simple and realistic systems. • The new model shows a higher efficiency than the old model in MC simulations. - Abstract: A new chord length distribution model is proposed to characterize the stochastic distribution of TRISO fuel particles in nuclear reactor systems and is used in the chord length sampling (CLS) method to analyze the neutronic behavior of TRISO fuel systems. In this model, the coating layers of fuel particles are homogenized with the background matrix region. The probability density function (PDF) of the chord length between fuel kernels, instead of fuel particles, is developed and is used in the CLS method. We first apply the new CLS model to solving one-group eigenvalue problems in a simplified 3-D stochastic medium system. Good accuracy is obtained in predicting the multiplication factor and fission density distribution. The relative differences are within 1.0% for both the multiplication factor and the total fission density in all the studied scenarios. We then apply the new CLS model to analyzing three realistic reactor designs: two Very High Temperature Gas-cooled Reactor unit cells and one fuel pin unit cell of an innovative light water reactor design with accident tolerant fuel. Infinite multiplication factor and intra-Dancoff factor are evaluated for the three unit cells respectively. Compared with the reference results, predictions from CLS simulations show a relative difference of less than 0.26% for infinite multiplication factors and less than 1.0% for intra-Dancoff factors in all the studied cases. Meanwhile, a significant improvement in the computational efficiency has been observed for the CLS new model compared with the

  12. 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

  13. 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

  14. Research reactors - an overview

    Energy Technology Data Exchange (ETDEWEB)

    West, C.D.

    1997-03-01

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

  15. 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

  16. Development of high temperature fission counter-chamber(FC)S for a top entry loop type fast breeder reactor

    International Nuclear Information System (INIS)

    Prototype high temperature fission counter-chambers have been made as neutron detectors for installation in the reactor vessel of the 600MWe-class top entry loop type fast breeder reactor. Using these prototypes as samples, a high-temperature endurance test has been conducted. The validity of the prototypes has been established by the test results, which show that the prototypes nearly satisfy the design performance. (author)

  17. Progress of the Russian RERTR program: Development of new-type fuel elements for Russian-built research reactors

    International Nuclear Information System (INIS)

    The new design of pin-type fuel elements and fuel assembly on their basis for Russian research reactors has been developed. The number of following activities has been performed: computational and experimental substantiation of fuel element design; development of fabrication process of fuel elements; manufacturing of experimental assembly for lifetime in-pile tests. The relevant fuel assemblies are considered to be perspective for usage as low-enriched fuel for Russian research reactors. (author)

  18. 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

  19. Gas Reactor International Cooperative Program. Interim report. Construction and operating experience of selected European Gas-Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    1978-09-01

    The construction and operating experience of selected European Gas-Cooled Reactors is summarized along with technical descriptions of the plants. Included in the report are the AVR Experimental Pebble Bed Reactor, the Dragon Reactor, AGR Reactors, and the Thorium High Temperature Reactor (THTR). The study demonstrates that the European experience has been favorable and forms a good foundation for the development of Advanced High Temperature Reactors.

  20. Gas Reactor International Cooperative Program. Interim report. Construction and operating experience of selected European Gas-Cooled Reactors

    International Nuclear Information System (INIS)

    The construction and operating experience of selected European Gas-Cooled Reactors is summarized along with technical descriptions of the plants. Included in the report are the AVR Experimental Pebble Bed Reactor, the Dragon Reactor, AGR Reactors, and the Thorium High Temperature Reactor (THTR). The study demonstrates that the European experience has been favorable and forms a good foundation for the development of Advanced High Temperature Reactors

  1. Development of an innovative plate-type SG for sodium cooled fast reactor

    International Nuclear Information System (INIS)

    The concept of an innovative plate type SG for sodium cooled small fast reactor fabricated by using the Hot Isostatic Pressing (HIP) 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. (1) It has found that the volume of the innovative plate type SG becomes 60% of the helical coil double-wall tube type SG by the sizing estimation study. (2) The optimum conditions of HIP for the modified 9Cr1Mo steel expected as a primary candidate of the SG material of the LMFR were found by the Hip tests, tensile tests. As a result, the optimum condition was found. A solder was selected as a joining material to laminate the heat transfer tube plates. The construction feasibility was confirmed through the SG partial model's trial manufacturing. (3) The structural integrity under the system condition of LMFR was confirmed by the thermal stress analysis. The leakage detection groove for the one side breakage detection is set up in the proposed SG concept. As a result of the crack extension analysis under the repetition stress that assumes start/stop of reactor, it was clarified that there is enough time to detect the one-side breakage before the sodium water reaction accident happens. The requirements of the eliminating of the secondary system by using this SG were arranged. As a result, the problem that becomes a big hurdle has not been extracted through it is necessary to confirm some issues. When the reduction of the Na handling area was taken into consideration, the effect of the cost reduction was estimated to be about 8.5%. Therefore, the prospect of the development of compact SG for small LMFR that has a possibility of the elimination of the secondary system was able to be obtained. (author)

  2. Development of a fuel failure monitoring method for a pool-type research reactor

    International Nuclear Information System (INIS)

    Studies on developing a sensitive monitoring method of possible release of fission products (FP) from a fuel element have been made for a pool-type research reactor. It consists of introducing gas bubbles into reactor coolant water to extract effectively the dissolved fission rare gases, 89Kr and 138Xe, produced somewhere in the core, and counting their respective daughter nuclides, 89Rb and 138Cs with high efficiency. The measurements were done by either method, (I) on a filter paper by sucking the bubbled gas and air covering water of the reactor tank, or (II) in the washing water of bubbled gas sampled into a bottle at the water surface. The followings are the summary of the results obtained. (1) DE increased as much as 30 times or more compared with no gas bubbling. (2) DE largely increased with increasing flow rate of introducing gas. (3) DE increased with increasing depth of the gas exit in the water. (4) DE at the same depth depended on the position of gas exit. It was larger for 'side' position than for 'center', due to the water convection in the tank. (5) DE largely depended on the condition of whether the primary cooling system was operated or not. (6) In Method II, DE depended on the time of standing, and it showed maximum at the theoretically predicted value. (7) The theoretical analysis for the effect of depth suggests that DE should be proportional to the value =(1+D /1033)2-(1+D /1033)5/3=, where D is the depth (in cm). The trend agreed at least partly with the observed data. (8) As a continuous mode experiment, we constructed an automatic fuel monitoring system for routine use by adopting Method II. It is composed of an intermittent sampling of the bubbling gas into bottle at the water surface, washing it with water after definite time of standing, and measuring the nuclides contained in the water. (J.P.N.)

  3. Locally manufactured films for neutron flux measurement in the MNSR type reactor

    International Nuclear Information System (INIS)

    Highlights: • Metal films deposited on Teflon are prepared to use as neutron monitors in the MNSR. • Ti, Al, V, and Ag films have been locally prepared by two different methods. • The thermal neutron flux was measured using Ti, Al, V, and Ag films. • V and Ag films were used as neutron monitors for the first time in MNSR type reactor. • With compared to published results in literature our neutron monitors are validated. - Abstract: Metal films deposited on Teflon are used in the Miniature Neutron Source Reactor (MNSR) for the first time to study their usability as neutron activation detectors for the thermal neutron flux measurements in the reactor. For this purpose Titanium, Aluminum, Vanadium, and Silver films deposited on Teflon have been locally prepared at room temperature using two methods: the vacuum arc deposition and DC Magnetron sputtering techniques. The thermal neutron flux in the MNSR inner irradiation site was measured using the prepared metal films. The results at the 95% level of confidence of the neutron flux using the metal films deposited on Teflon by the vacuum arc deposition for Titanium, Aluminum, and Vanadium were: (9.9 ± 0.3) × 1011, (1.4 ± 0.3) × 1012, (1.2 ± 0.2) × cm−2 s−1, respectively. The result at the same level of confidence of the neutron flux using the metal films deposited on Teflon by the DC Magnetron sputtering for Silver was: (1.5 ± 0.2) × 1011 cm−2 s−1. Good agreements are noticed between our obtained mean value (9.3 ± 0.9) × 1011 cm−2 s−1 and the previous published results

  4. 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)

  5. Minor actinide transmutation in a board type sodium cooled breed and burn reactor core

    International Nuclear Information System (INIS)

    Highlights: • A 1250 MWt board type sodium cooled breed and burn reactor core is further designed. • MCNP–ORIGEN coupled code MCORE is applied to perform neutronics and depletion calculation. • Transmutation efficiency and neutronic safety parameters are compared under different MA weight fraction. - Abstract: In this paper, a board type sodium cooled breed and burn reactor core is further designed and applied to perform minor actinide (MA) transmutation. MA is homogeneously loaded in all the fuel sub-assemblies with a weight fraction of 2.0 wt.%, 4.0 wt.%, 6.0 wt.%, 8.0 wt.%, 10.0 wt.% and 12.0 wt.%, respectively. The transmutation efficiency, transmutation amount, power density distribution, neutron fluence distribution and neutronic safety parameters, such as reactivity, Doppler feedback, void worth and delayed neutron fraction, are compared under different MA weight fraction. Neutronics and depletion calculations are performed based on the self-developed MCNP–ORIGEN coupled code with the ENDF/B-VII data library. In the breed and burn reactor core, a number of breeding sub-assemblies are arranged in the inner core in a board type way (scatter load) to breed, and a number of absorbing sub-assemblies are arranged in the inner side of the outer core to absorb neutrons and reduce power density in this area. All the fuel sub-assemblies (ignition and breeding sub-assemblies) are shuffled from outside in. The core reached asymptotically steady state after about 22 years, and the average and maximum discharged burn-up were about 17.0% and 35.3%, respectively. The transmutation amount increased linearly with the MA weight fraction, while the transmutation rate parabolically varied with the MA weight fraction. Power density in ignition sub-assembly positions increased with the MA weight fraction, while decreased in breeding sub-assembly positions. Neutron fluence decreased with the increase of MA weight fraction. Generally speaking, the core reactivity and void

  6. Fuel assemblies for use in high conversion type water cooled reactors

    International Nuclear Information System (INIS)

    Purpose: To enable to easily change the water/uranium ratio of identical fuel assemblies. Constitution: In the reactor core of a high conversion type water cooled reactor, it is necessary to accumulate Pu by decreasing the water/uranium ratio in a conversion region, while efficiently burn Pu and U by increasing the water/uranium ratio in the burner region. However, it has been difficult in the fuel assemblies of the prior structure to change the water/uranium ratio. In the present invention, water exclusion rods replaceable with water rods are detachably disposed. That is, in the conversion region, hollow or solid water exclusion rods made of zirconium alloy are inserted to set the water/uranium ratio lower. Then, in the case of charging the fuel assembly into the burner region, water rods are inserted instead of the water exclusion rods to set the water/uranium ratio higher. In this way, it is possible to easily change the water/uranium ratio by a simple method. (Kamimura, M.)

  7. Hot Water Layer and Thermal Stratification in an Open-pool type Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Yong-Seok; Yoon, Hyun-Gi; Choi, Jeongwoon; Kim, Seong-Hoon; Chi, Dae-Young [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    In many open-pool type research reactors, a hot water layer is introduced in the upper part of the pool as a shielding layer to reduce the radiation level on the pool top. By maintaining the hot water layer in a properly higher temperature than the lower part of the pool, a thermally stratified region is developed below the hot water layer and the flows in the lower part of the pool is successfully isolated from the upper part of the pool. This reduces a mass transport from the lower part of the pool to the pool upper part and consequently the radioactivity level on the pool top is also diminished. In this study, the characteristics of the hot water layer and the thermally stratified region in the pool of the KIJANG Research Reactor (KJRR) are investigated. Numerical simulation on a 3D simplified model of the pool of KJRR is conducted using the commercial CFD software ANSYS FLUENT 13.0. The results show initial time evolutions of the temperatures and the flow velocities in the pool toward each quasi steady state. For the shielding analysis in the pool which is required to estimate the radioactivity of the hot water layer, the mixing rate between the hot water layer and the lower part of the pool is important variable as well as the thicknesses of the hot water layer and the stratified region. In further study, the mixing rate will be estimated by post processing the obtained results.

  8. Water flow characteristics of Baumkuchen type fuel elements for Kyoto University high neutron flux reactor

    International Nuclear Information System (INIS)

    The Kyoto University high neutron flux reactor is a light water-moderated and cooled, divided core type reactor with heavy water reflector. In the core, six inside fuel elements and twelve outside fuel elements are arranged in double ring form, and two cylindrical, divided cores are placed at 15 cm distance. The flow rate distribution and pressure loss in the fuel elements constitute the base of the thermo-hydraulic design of the core, therefore the model fuel elements of full size were made, and the water flow experiment was carried out to examine their characteristics. It was found that the flow velocity in channels was strongly affected by the accuracy of channel gaps. The calculation of pressure loss in fuel elements, the experiments on inside fuel elements and outside fuel elements, and the results of experiments such as the calibration of the cooling channels in outside fuel elements, the relation between total flow rate and pressure loss, and the characteristics of flow at the time of reverse flow are reported. The general characteristics of flow in fuel elements were in good agreement with the prediction. In the pressure loss in fuel elements, the friction between fuel plates and the resistance of nozzles were the controlling factors under the rated operating conditions of the HFR. (Kako, I.)

  9. 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)

  10. The review of fuel types for Russian research reactors. Their fabrication and quality control

    International Nuclear Information System (INIS)

    The design of tubular fuel elements (FEs) for research reactor fuel assemblies (FAs) is considered. Commercial extrusion and annular-type technologies for tubular FE fabrication are described. 'Extrusion' technology is based on fabrication of tubular billet of fuel core by means of powder metallurgy followed by hot extrusion of fuel core tubular billet and tubular cladding billet. The process is completed with FE assembly operation. 'Annular' technology is based on fuel core fabrication using powder metallurgy followed by chemical treatment of fuel core surface and fuel core insertion into the cladding. The list of FE and FA control operations to check their conformance to the required quality level is given. The most common FA designs (WWR-M2, WWR-M5, IRT-2M, IRT-3M, MR, MIR, WWR-TS, IVV-2M, IVV-10, TWR-S, IR-100) for research reactors built according to the Russian projects are described. The Quality Assurance System in operation at 'Novosibirsk' Chemical Concentrates Plant' is presented. (author)

  11. 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)

  12. The Choice of thermal reactor systems. A report by the National Nuclear Corporation Limited

    International Nuclear Information System (INIS)

    The report to the Secretary of State in Great Britain by the National Nuclear Corporation following their assessment of the three thermal reactor systems, the AGR, PWR and SGHWR type reactors, which was performed in order to assist in the decision on the choice of thermal reactors for the U.K., is in three parts. Part I is an assessment of the three systems. It comprises: a description of the general method of assessment; a commentary in which are summarised discussions on the most important issues influencing reactor choice, i.e. safety, component failure, operational characteristics, development programme, construction programme; implications for the U.K. industry; costs; and reference design of each system. Part II consists of related questions and answers accompanied by commentaries on public acceptability and views from industry. Part III contains some conclusions including an analysis on the implications of the choices open and a summary of the main features of the assessment. (U.K.)

  13. Computation of gap conductance in different fuel assemblies in VVER-1000 type reactors

    International Nuclear Information System (INIS)

    In this paper, a calculation for fresh fuels gap conductance at different axial lengths of fuel assemblies of the VVER-1000 type reactors has been made using two models of Calza-Bini and Relap5. By applying these two models, the dependency of the fuel outer surface temperature and the clad inner surface temperature of the gap conductance has been determined upon using following procedures: Coupling gap conductance model computer programming to obtain temperature at different axial lengths of the fuel and clad; and coupling gap conductance model to the Cobra-En output code. The results of calculations and comparison with the final safety analysis report results showed that the Relap5 model is less accurate than the Calza-Bini model. The Calza-Bini model agrees well with the final safety analysis report results. By combining these two models, a new model with a better accuracy was proposed for the gap conductance.

  14. Synthesis of layered double hydroxide nanosheets by coprecipitation using a T-type microchannel reactor

    Science.gov (United States)

    Pang, Xiujiang; Sun, Meiyu; Ma, Xiuming; Hou, Wanguo

    2014-02-01

    The synthesis of Mg2Al-NO3 layered double hydroxide (LDH) nanosheets by coprecipitation using a T-type microchannel reactor is reported. Aqueous LDH nanosheet dispersions were obtained. The LDH nanosheets were characterized by X-ray diffraction, transmission electron microscopy, atomic force microscopy and particle size analysis, and the transmittance and viscosity of LDH nanosheet dispersions were examined. The two-dimensional LDH nanosheets consisted of 1-2 brucite-like layers and were stable for ca. 16 h at room temperature. In addition, the co-assembly between LDH nanosheets and dodecyl sulfate (DS) anions was carried out, and a DS intercalated LDH nanohybrid was obtained. To the best of our knowledge, this is the first report of LDH nanosheets being directly prepared in bulk aqueous solution. This simple, cheap method can provide naked LDH nanosheets in high quantities, which can be used as building blocks for functional materials.

  15. El microbioma de un suelo agrícola supresivo

    OpenAIRE

    Vida, Carmen; de Vicente, Antonio; Cazorla, Francisco M.

    2015-01-01

    La aplicación de enmiendas orgánicas es una estrategia agrícola que puede provocar la mejora del estado fitosanitario de los suelos agrícolas. Esta medida se incluye en el control integrado de la podredumbre blanca radicular del aguacate, causada por el hongo fitopatógeno Rosellinia necatrix en la zona Mediterránea. En este trabajo se evalúa la capacidad supresiva de suelos agrícolas del cultivo de aguacate, enmendados con cáscara de almendra compostada. Además se caracteriza la microbiota qu...

  16. Adsorption and transformation of PAHs from water by a laccase-loading spider-type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Niu, Junfeng, E-mail: junfengn@bnu.edu.cn [State Key Laboratory of Water Environment Simulation, School of Environment, Beijing Normal University, Beijing 100875 (China); Dai, Yunrong, E-mail: daiyunrong@mail.bnu.edu.cn [State Key Laboratory of Water Environment Simulation, School of Environment, Beijing Normal University, Beijing 100875 (China); Guo, Huiyuan, E-mail: hyguo0216@163.com [State Key Laboratory of Water Environment Simulation, School of Environment, Beijing Normal University, Beijing 100875 (China); Xu, Jiangjie, E-mail: 1993120hb@163.com [State Key Laboratory of Water Environment Simulation, School of Environment, Beijing Normal University, Beijing 100875 (China); Shen, Zhenyao, E-mail: zyshen@bnu.edu.cn [State Key Laboratory of Water Environment Simulation, School of Environment, Beijing Normal University, Beijing 100875 (China)

    2013-03-15

    Highlights: ► Laccase-loading spider-type reactor (LSTR) is got by emulsion electrospinning. ► LSTR consists of beads-in-string fibers with more laccase and higher activity. ► LSTR can achieve the rapid and efficient removal of PAHs from water. ► Aquatic environmental factors have little influence on the PAH removal by LSTR. ► A synergetic mechanism includes adsorption, directional migration and degradation. -- Abstract: The remediation of polycyclic aromatic hydrocarbons (PAHs) polluted waters has become a concern as a result of the widespread use of PAHs and their adverse impacts on water ecosystems and human health. To remove PAHs rapidly and efficiently in situ, an active fibrous membrane, laccase-loading spider-type reactor (LSTR) was fabricated by electrospinning a poly(D,L-lactide-co-glycolide) (PDLGA)/laccase emulsion. The LSTR is composed of beads-in-string structural core–shell fibers, with active laccase encapsulated inside the beads and nanoscale pores on the surface of the beads. This structure can load more laccase and retains higher activity than do linear structural core–shell fibers. The LSTR achieves the efficient removal/degradation of PAHs in water, which is attributed to not only the protection of the laccase activity by the core–shell structure but also the pre-concentration (adsorption) of PAHs on the surface of the LSTR and the concentration of laccase in the beads. Moreover, the effects of pH, temperature and dissolved organic matter (DOM) concentration on the removal of PAHs by the LSTR, in comparison with that by free laccase, have been taken into account. A synergetic mechanism including adsorption, directional migration and degradation for PAH removal is proposed.

  17. Thermal-hydraulic Fortran program for steady-state calculations of plate-type fuel research reactors

    Directory of Open Access Journals (Sweden)

    Khedr Ahmed

    2008-01-01

    Full Text Available The safety assessment of research and power reactors is a continuous process covering their lifespan and requiring verified and validated codes. Power reactor codes all over the world are well established and qualified against real measuring data and qualified experimental facilities. These codes are usually sophisticated, require special skills and consume a lot of running time. On the other hand, most research reactor codes still require much more data for validation and qualification. It is, therefore, of benefit to any regulatory body to develop its own codes for the review and assessment of research reactors. The present paper introduces a simple, one-dimensional Fortran program called THDSN for steady-state thermal-hydraulic calculations of plate-type fuel research reactors. Besides calculating the fuel and coolant temperature distributions and pressure gradients in an average and hot channel, the program calculates the safety limits and margins against the critical phenomena encountered in research reactors, such as the onset of nucleate boiling, critical heat flux and flow instability. Well known thermal-hydraulic correlations for calculating the safety parameters and several formulas for the heat transfer coefficient have been used. The THDSN program was verified by comparing its results for 2 and 10 MW benchmark reactors with those published in IAEA publications and a good agreement was found. Also, the results of the program are compared with those published for other programs, such as the PARET and TERMIC.

  18. Thermal-hydraulic Fortran program for steady-state calculations of plate-type fuel research reactors

    International Nuclear Information System (INIS)

    The safety assessment of research and power reactors is a continuous process covering their life span and requiring verified and validated codes. Power reactor codes all over the world are well established and qualified against real measuring data and qualified experimental facilities. These codes are usually sophisticated, require special skills and consume a lot of running time. On the other hand, most research reactor codes still require much more data for validation and qualification. It is, there fore, of benefit to any regulatory body to develop its own codes for the re view and assessment of research reactors. The present paper introduces a simple, one-dimensional Fortran program called THDSN for steady-state thermal hydraulic calculations of plate-type fuel research reactors. Besides calculating the fuel and coolant temperature distributions and pressure gradients in an average and hot channel, the program calculates the safety limits and margins against the critical phenomena encountered in research reactors, such as the on set of nucleate boiling, critical heat flux and flow in stability. Well known thermal-hydraulic correlations for calculating the safety parameters and several formulas for the heat transfer coefficient have been used. The THDSN program was verified by comparing its results for 2 and 10 MW bench mark reactors with those published in IAEA publications and a good agreement was found. Also, the results of the program are compared with those published for other programs, such as the PARET and TERMIC. (author)

  19. Floating nuclear heat. And power station 'Pevec' with KLT-40S type reactor plant for remote regions of Russia

    International Nuclear Information System (INIS)

    Floating small nuclear power plants power for local energy systems of littoral regions of Russia, located far from central energy system, open a new line in nuclear power development. Designing a floating power unit of a lead nuclear heat and power generating station for port Pevec at the Chuckchee national district is currently nearing completion. Most labor-intensive components are being manufactured. The co-generation NPP Pevec is to be created on the basis of a floating power unit with KLT-40S type reactor plant. KLT-40S reactor plant is based on similar propulsion plants, verified at operation of Russia's nuclear-powered civil ships, evolutionary improved by elimination of 'weak points' revealed during its prototypes operation or on the basis of safety analysis. KLT-40S reactor plant uses the most wide-spread and developed in the world practice PWR-type reactor. KLT-40S meets contemporary national and international requirements imposed to future reactor plants. The NHPS description, its main technical-economic data, environmental safety indices, basic characteristics of KLT-40S reactor plant are presented. Prospects of small NPPs utilization outside Russia, particularly as an energy source for sea water desalination, are considered. (author)

  20. Determination of neutronic fluxes in research nuclear reactor of Triga Mark I and WWRS types

    International Nuclear Information System (INIS)

    In this paper is presented the determination of the thermal, epithermal and fast neutron fluxes, using neutron activation analysis technique, for two research nuclear reactors of different design: the Triga Mark I reactor was designed by Gulf General Atomic Co in USA and the WWRS reactor was designed in the URSS, both in the 50's years. (Author)

  1. 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)

  2. The next generation type of nuclear reactor management system by using genetic algorithms. Coordinative optimization of reactor design and reactor control

    International Nuclear Information System (INIS)

    Two steps integrated optimization algorithm on the basis of the improvement genetic algorithm (GA) was developed for BWR core optimization. It showed good convergence performance keeping with global search capability. When the method was applied to 1356MWe BWR design, optimization was realized by the practical cost. An integer combinatorial optimization using MAA (Multi Agent Algorithm) was developed. MAA was introduced to the first-step part of two-step GA and the convergence performance increased. An idea of MAA proposed by us gets a hint from humane behavior in the group. The reactor design and reactor control of BWR, the coordinative optimization, its application to the practical plant and the next generation reactor control system are explained. (S.Y.)

  3. Validation of the Physics Analysis used to Characterize the AGR-1 TRISO Fuel Irradiation Test

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James W.; Harp, Jason M.; Demkowicz, Paul A.; Hawkes, Grant L.; Chang, Gray S.

    2015-05-01

    The results of a detailed physics depletion calculation used to characterize the AGR-1 TRISO-coated particle fuel test irradiated in the Advanced Test Reactor (ATR) at the Idaho National Laboratory are compared to measured data for the purpose of validation. The particle fuel was irradiated for 13 ATR power cycles over three calendar years. The physics analysis predicts compact burnups ranging from 11.30-19.56% FIMA and cumulative neutron fast fluence from 2.21?4.39E+25 n/m2 under simulated high-temperature gas-cooled reactor conditions in the ATR. The physics depletion calculation can provide a full characterization of all 72 irradiated TRISO-coated particle compacts during and post-irradiation, so validation of this physics calculation was a top priority. The validation of the physics analysis was done through comparisons with available measured experimental data which included: 1) high-resolution gamma scans for compact activity and burnup, 2) mass spectrometry for compact burnup, 3) flux wires for cumulative fast fluence, and 4) mass spectrometry for individual actinide and fission product concentrations. The measured data are generally in very good agreement with the calculated results, and therefore provide an adequate validation of the physics analysis and the results used to characterize the irradiated AGR-1 TRISO fuel.

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

    International Nuclear Information System (INIS)

    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

  5. 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.

  6. Two-channel model for dynamic analysis of GCR type reactor, Mathematical model

    International Nuclear Information System (INIS)

    A two-channel model for reactor dynamic analysis was developed. It enables representation of time dependent behaviour of a reactor as a whole and to obtain time and space dependent changes of temperature in any of the reactor channel. Model is suitable for follow-up of phenomena in limited time intervals up to few tens of minutes, since long term variations caused by fuel burnup and fission products are not taken into account in the model. Parameters are defined to cover the reactor power range from minimum to maximum. Model describes two main processes in the reactor: power generation dependent on the neutron flux and cooling

  7. Nasal Carriage as a Source of agr-Defective Staphylococcus aureus Bacteremia

    OpenAIRE

    Smyth, Davida S.; Kafer, Jared M.; Wasserman, Gregory A.; Velickovic, Lili; Mathema, Barun; Robert S Holzman; Knipe, Tiffany A.; Becker, Karsten; von Eiff, Christof; Peters, Georg; Chen, Liang; Kreiswirth, Barry N.; Novick, Richard P.; Shopsin, Bo

    2012-01-01

    Inactivating mutations in the Staphylococcus aureus virulence regulator agr are associated with worse outcomes in bacteremic patients. However, whether agr dysfunction is primarily a cause or a consequence of early bacteremia is unknown. Analysis of 158 paired S. aureus clones from blood and nasal carriage sites in individual patients revealed that recovery of an agr-defective mutant from blood was usually predicted by the agr functionality of carriage isolates. Many agr-positive blood isolat...

  8. The high temperature reactor - an interim balance of development and operation

    International Nuclear Information System (INIS)

    The high temperature reactor is the modern version of the gas cooled reactor. The interim balance presented in this report therefore refers to all gas cooled types of reactor, i.e. Magnox, AGR, and HTR. The period covered by the balance begins 28 years ago, when the British gas cooled reactor of Calder Hall went critical for the firt time. The experience accumulated with the German experimental HTR plant, the Juelich AVR reactor, has been extremely satisfactory, both with respect to the operating behaviour and to safety. The HTR fuel elements have a high safety margin against excessive operating temperatures. Although the dominating role played by the light water reactor line has so far prevented the commercial application of high temperature reactors, developments in recent years seem to indicate new market chances for the high temperature reactor line. In this connection, special importance attaches to the prototype THTR-300, which is about to be commissioned, and to the HTR-100 and HTR-500 conceptual design drafts and the modular reactor. As the design data of the THTR-300 and the HTR-500 are partly identical, the latter plant is characterized by foreseeable preparation and construction times; in addition, the licensability of the HTR-500 has already been confirmed. A medium sized reactor like this could be the link between electricity generation and the generation of process heat and space heat. (orig.)

  9. Determination of the Quantity of I-135 Released from the AGR Experiment Series

    Energy Technology Data Exchange (ETDEWEB)

    Scates, Dawn Marie [Idaho National Laboratory; Walter, John Bradley [Idaho National Laboratory; Reber, Edward Lawrence [Idaho National Laboratory; Sterbentz, James William [Idaho National Laboratory; Petti, David Andrew [Idaho National Laboratory

    2014-10-01

    A series of three Advanced Gas Reactor (AGR) experiments have been conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). From 2006 through 2014, these experiments supported the development and qualification of the new U.S. tri structural isotropic (TRISO) particle fuel for Very High Temperature Reactors (VHTR). Each AGR experiment consisted of multiple fueled capsules, each plumbed for independent temperature control using a mix of helium and neon gases. The gas leaving a capsule was routed to individual Fission Product Monitor (FPM) detectors. For intact fuel particles, the TRISO particle coatings provide a substantial barrier to fission product release. However, particles with failed coatings, whether because of a minute percentage of initially defective particles, those which fail during irradiation, or those designed to fail (DTF) particles, can release fission products to the flowing gas stream. Because reactive fission product elements like iodine and cesium quickly deposit on cooler capsule components and piping structures as the effluent gas leaves the reactor core, only the noble fission gas isotopes of Kr and Xe tend to reach FPM detectors. The FPM system utilizes High Purity Germanium (HPGe) detectors coupled with a thallium activated sodium iodide NaI(Tl) scintillator. The germanium detector provides individual isotopic information, while the NaI(Tl) scintillator is used as a gross count rate meter. During irradiation, the 135mXe concentration reaching the FPM detectors is from both direct fission and by decay of the accumulated 135I. About ~2.5 hours after irradiation (ten 15.3 minute 135mXe half lives) the directly produced 135mXe has decayed and only the longer lived 135I remains as a source. Decay systematics dictate that 135mXe will be in secular equilibrium with its 135I parent, such that it’s production rate very nearly equals the decay rate of the parent, and its concentration in the flowing gas stream will appear to

  10. The reliability of thermocouples in the Windscale AGR

    International Nuclear Information System (INIS)

    A brief description is given of two surveys carried out during 1977/78 on Windscale AGR thermocouples. One was concerned with fuel pin thermocouples, the other with channel gas inlet thermocouples. (author)

  11. AGR2 Predicts Tamoxifen Resistance in Postmenopausal Breast Cancer Patients

    Directory of Open Access Journals (Sweden)

    Roman Hrstka

    2013-01-01

    Full Text Available Endocrine resistance is a significant problem in breast cancer treatment. Thus identification and validation of novel resistance determinants is important to improve treatment efficacy and patient outcome. In our work, AGR2 expression was determined by qRT-PCR in Tru-Cut needle biopsies from tamoxifen-treated postmenopausal breast cancer patients. Our results showed inversed association of AGR2 mRNA levels with primary treatment response (P=0.0011 and progression-free survival (P=0.0366 in 61 ER-positive breast carcinomas. As shown by our experimental and clinical evaluations, elevated AGR2 expression predicts decreased efficacy of tamoxifen treatment. From this perspective, AGR2 is a potential predictive biomarker enabling selection of an optimal algorithm for adjuvant hormonal therapy in postmenopausal ER-positive breast cancer patients.

  12. Processing and microstructural characterisation of a UO{sub 2}-based ceramic for disposal studies on spent AGR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Hiezl, Z., E-mail: z.hiezl11@imperial.ac.uk [Department of Materials, Imperial College, London SW7 2AZ (United Kingdom); Hambley, D.I. [Spent Fuel Management and Disposal, UK National Nuclear Laboratory (NNL), Central Laboratory, Sellafield CA20 1PG (United Kingdom); Padovani, C. [Radioactive Waste Management Limited (formerly the Radioactive Waste Management Directorate of the Nuclear Decommissioning Authority), Harwell OX11 0RH (United Kingdom); Lee, W.E. [Department of Materials, Imperial College, London SW7 2AZ (United Kingdom)

    2015-01-15

    Preparation and characterisation of a Simulated Spent Nuclear Fuel (SIMFuel), which replicates the chemical state and microstructure of Spent Nuclear Fuel (SNF) discharged from a UK Advanced Gas-cooled Reactor (AGR) after a cooling time of 100 years is described. Given the relatively small differences in radionuclide inventory expected over longer time periods, the SIMFuel studied in this work is expected to be also representative of spent fuel after significantly longer periods (e.g. 1000 years). Thirteen stable elements were added to depleted UO{sub 2} and sintered to simulate the composition of fuel pellets after burn-ups of 25 and 43 GWd/tU and, as a reference, pure UO{sub 2} pellets were also investigated. The fission product distribution was calculated using the FISPIN code provided by the UK National Nuclear Laboratory. SIMFuel pellets were up to 92% dense and during the sintering process in H{sub 2} atmosphere Mo–Ru–Rh–Pd metallic precipitates and grey-phase ((Ba, Sr)(Zr, RE) O{sub 3} oxide precipitates) formed within the UO{sub 2} matrix. These secondary phases are present in real PWR and AGR SNF. Metallic precipitates are generally spherical and have submicron particle size (0.8 ± 0.7 μm). Spherical oxide precipitates in SIMFuel measured up to 30 μm in diameter, but no data were available in the public domain to compare this to AGR SNF. The grain size of actual AGR SNF (∼3–30 μm) is larger than that measured in AGR SIMFuel (∼2–5 μm)

  13. Processing and microstructural characterisation of a UO2-based ceramic for disposal studies on spent AGR fuel

    International Nuclear Information System (INIS)

    Preparation and characterisation of a Simulated Spent Nuclear Fuel (SIMFuel), which replicates the chemical state and microstructure of Spent Nuclear Fuel (SNF) discharged from a UK Advanced Gas-cooled Reactor (AGR) after a cooling time of 100 years is described. Given the relatively small differences in radionuclide inventory expected over longer time periods, the SIMFuel studied in this work is expected to be also representative of spent fuel after significantly longer periods (e.g. 1000 years). Thirteen stable elements were added to depleted UO2 and sintered to simulate the composition of fuel pellets after burn-ups of 25 and 43 GWd/tU and, as a reference, pure UO2 pellets were also investigated. The fission product distribution was calculated using the FISPIN code provided by the UK National Nuclear Laboratory. SIMFuel pellets were up to 92% dense and during the sintering process in H2 atmosphere Mo–Ru–Rh–Pd metallic precipitates and grey-phase ((Ba, Sr)(Zr, RE) O3 oxide precipitates) formed within the UO2 matrix. These secondary phases are present in real PWR and AGR SNF. Metallic precipitates are generally spherical and have submicron particle size (0.8 ± 0.7 μm). Spherical oxide precipitates in SIMFuel measured up to 30 μm in diameter, but no data were available in the public domain to compare this to AGR SNF. The grain size of actual AGR SNF (∼3–30 μm) is larger than that measured in AGR SIMFuel (∼2–5 μm)

  14. Claim criteria of significant events implying the safety of PWR type reactors

    International Nuclear Information System (INIS)

    There are ten criteria for the declaration of the significant events implying the safety for PWR type reactors. First criterion: Automatic stop of the reactor: manual or automatic, inconvenient starting or not, the function of automatic stop of the reactor, whatever is the state of the reactor, with the exception of the deliberate starting resulting from planned actions. Second criterion: Starting of one of the systems of protection, manual or automatic, inconvenient starting or not, of one of the systems of protection, with the exception of the deliberate starting resulting from planned actions. Third criterion: Disregard of the technical specifications of exploitation (S.T.E ), or an event which would have been able to lead to a disregard of the S.T.E., if the same event had occurred, the installation having been in a different state, any disregard of one or several permanent conditions defined in S.T.E., any disregard of the conditions of a dispensation in S.T.E., any overtaking of periods when it is not prescribed by state of fold, any unavailability provoked outside the conditions planned by the main rules of exploitation, not identified beforehand or identified but untreated according to the prescriptions of the S.T.E. fourth criterion: Internal or external aggression, happening of a natural external phenomenon or in relation with a human activity, or happening of an internal flooding, a fire or another phenomenon susceptible to affect the availability of the equipment important for the safety. Fifth criterion: Act or attempt of act of hostility susceptible to affect the safety of the installation. Sixth criterion: Passage in state of fold in application of the technical specifications of exploitation or the accidental procedures of driving following an unforeseen behavior of the installation. Seventh criterion: Event having cause or being able to cause multiple failures, unavailability of equipment due to the same failure either affecting all the ways of a

  15. AGR2 Predicts Tamoxifen Resistance in Postmenopausal Breast Cancer Patients

    OpenAIRE

    Roman Hrstka; Veronika Brychtova; Pavel Fabian; Borivoj Vojtesek; Marek Svoboda

    2013-01-01

    Endocrine resistance is a significant problem in breast cancer treatment. Thus identification and validation of novel resistance determinants is important to improve treatment efficacy and patient outcome. In our work, AGR2 expression was determined by qRT-PCR in Tru-Cut needle biopsies from tamoxifen-treated postmenopausal breast cancer patients. Our results showed inversed association of AGR2 mRNA levels with primary treatment response (P = 0.0011) and progression-free survival (P = 0.0366)...

  16. A strategic analysis of the development of structural materials for proto-type reactors for fusion

    International Nuclear Information System (INIS)

    Structural Materials Research and Development Subcommittee of Nuclear Materials Committee in Japan Atomic Energy Research Institute had made a study to propose a strategy how to expedite the research and development of structural materials for fusion reactors. This study was carried out along with the interim report of the Development of Structural Materials in Fusion Reactors proposed by Planning and Promotion Subcommittee of Fusion Council as well as with the Third Phase Basic Program of Fusion Research and Development settled by the Atomic Energy Commission. The present report was published to publicize the results of analyses of this study. In this report we focused mostly on the development of structural materials of blankets for tritium breeding because it is considered to be the most difficult task in the materials development due to severe conditions imposing on the blankets. We selected three candidate materials, namely, reduced low activation ferritic/ martensitic steel, SiC/SiC composites and Vanadium alloys, and elucidate the conditions in which these materials would be used as well as the design requirements for each material. Based on these conditions and requirements, we described the present status and the key issues of each material. For the development of the structural materials for the blankets, the keenest issue is the improvement and evaluation of radiation integrity and stability. Therefore, the necessity of radiation facilities, especially accelerator-type neutron sources with near fusion energy spectra was described. In addition the usage of fission reactors as irradiation facilities was also emphasized. In the processing of this reviewing we categorized reduced low activation ferritic/martensitic steel as advanced material, and SiC/SiC composites and Vanadium alloys as next-generation advanced material from the present status of developmental maturity. A periodical check and review in order to take the future progress in the development of

  17. Neutronic study on seed-blanket type reduced-moderation water reactor fuel assembly

    International Nuclear Information System (INIS)

    Parametric studies have been done for a PWR-type reduced-moderation water reactor (RMWR) with seed-blanket fuel assemblies to achieve a high conversion ratio, a negative void reactivity coefficient and a high burnup by using MOX, metal (Pu+U+Zr) or T-MOX (PuO2+ThO2) fuels. From the result of the assembly burnup calculation, it has been seen that 50% to 60% of seed in a seed-blanket (MOX-UO2) assembly has higher conversion ratio compared to the other combinations of seeds and blankets. And the recommended number of seed-blanket layers is 20, in which the number of seed layers is 15 (S15) and that of blanket layers is 5 (B5). It was found that the conversion ratio of a seed-blanket assembly decreases, when seed and blanket are arranged so as to look like a flower shape (Hanagara). By the optimization of different parameters, the S15B5 fuel assembly with the height of seed of 1,000x2 mm, internal blanket of 150 mm and axial blanket of 400x2 mm is recommended for a high conversion ratio. In this assembly, the gap of seed fuel rod is 1.0 mm and that of blanket fuel rod is 0.4 mm. In the S15B5 assembly, the conversion ratio is 1.0 and the average burnup in (seed + internal blanket + outer blanket) region is 38 GWd/t. The cycle length of the core is 16.5 effective full power in month (EFPM) by 6 batches refuelling scheme and the enrichment of fissile Pu is 14.6 wt%. The void coefficient is +22 pcm/%void, though, it is expected that the void coefficient will be negative if the radial neutron leakage is taken into account in the calculation. It is also possible to use the S15B5 fuel assembly as a high burnup reactor to achieve 45 GWd/t in (seed + internal blanket + outer blanket) region, but, it is necessary to decrease the height of seed to 500x2 mm to improve the void coefficient. In this reactor, the conversion ratio is 0.97 and void coefficient is +21 pcm/%void. The fuel temperature coefficient is negative for both of the cases. It is possible to improve the conversion

  18. 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)

  19. Deterministic Analysis of a Beyond Design Basis Accident in a Low Power, Pin-Type Fuel Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nagah Abdou, Hesham Mohammed [INVAP S. E., Bariloche (Argentina)

    2013-07-01

    A Beyond Design Basis Accident has been analyzed for a pool type research reactor with pin-type, Zry4 clad fuel. This is a low power research reactor (maximum power: 100kW) with neutron beam facilities. Two scenarios are considered: a neutron beam rapture that results in a fraction of the core submerged in water and a catastrophic failure that results in a fully uncovered core. The paper discusses the different cooling mechanisms for these two BDBAs and compares results for both scenarios, with predictions of no core damage in any situation. Core damage is defined as CHFR↔1.5 and/or Tclad→T start of breakaway oxidation temperature. In addition, the paper compares calculations with a thermalhydraulic code and an analytical model. This paper allows to analyze the applicability of regular thermalhydraulic codes to BDBA accident scenarios in low power research reactors.

  20. 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)

  1. Calculation analysis of TRIGA MARK II reactor core composed of two types of fuel elements

    International Nuclear Information System (INIS)

    The most important properties of mixed cores are treated for TRIGA MARK II reactor, composed of standard (20% enriched, 8.5w% U content) and FLIP (70% enriched, 8.5w% U content) fuel elements. Large difference in enrichment and presence of burnable poison in FLIP fuel have strong influence on the main core characteristics, such as: fuel temperature coefficient, power defect, Xe and Sm worth, power and flux distributions, etc. They are significantly different for both types of fuel. Optimal loading of mixed cores therefore strongly depends on the loading pattern of both types of fuel elements. Results of systematic calculational analysis of mixed cores are presented. Calculations on the level of fuel element are performed with WIMSD-4 computer code with extended cross-section library. Core calculations are performed with TRIGAP two-group 1-D diffusion code. Results are compared to measurements and physical explanation is provided. Special concern is devoted to realistic mixed cores, for which optimal in-core fuel management is derived. Refs, figs and tabs

  2. 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

  3. Some economic aspects of natural uranium graphite gas reactor types. Present status and trends of costs in France

    International Nuclear Information System (INIS)

    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

  4. 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)

  5. 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

  6. Sipping equipment for WWER-440 type reactors to make in-core studies on the tightness of fuel elements

    International Nuclear Information System (INIS)

    At the meeting of the Kraftwerk Union A.G. a survey on the sipping equipment originally designed to detect fuel element failures of BWRs was presented. Possibilities to apply the device for tightness inspection of fuel cladding in WWER-440 type reactors by means of 1 or 8 fuel assemby sipping bells were pointed out. (V.N.)

  7. Radiation damages on construction materials used for the vessels of reactor WWER type, after 40-years of exploitation

    International Nuclear Information System (INIS)

    In the paper the radiation damages in materials are described. The types of the damages are also discussed. On the example of the WWER reactors the influence of irradiation on the material properties of steels after 40 years of exploitation as well as the removal of the radiation embrittlement using the annealing method are described. (authors)

  8. 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

  9. 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)

  10. Gas mixing processes in nuclear AGR boilers

    International Nuclear Information System (INIS)

    To ensure the safe operation and control of Nuclear (A.G.R.) boilers, 2-D computational models are currently under development. The aim of these models is to predict the flow and temperature distribution of the gas and water side under different operational conditions. These models are based on numerical solutions of the 2-D flow and heat transfer equations for turbulent flow in boilers. Measurements on a closely pitched tube bank with water cooling have demonstrated considerable discrepancies between experimental results and computer predictions. This investigation is therefore being carried out to study theoretically and experimentally the flow and heat transfer process under such a boiler condition. A two dimensional computer model has been developed which incorporates the effects of gas mixing and the interactions between the gas and water side. To cover the complete heat exchanger the governing equations are written in the lumped parameter form. The governing equations have been solved by a computer code written in FORTRAN-77. To test the validity of this model, the computer predictions have been compared with experimental results. Results to date indicate reasonable agreement with experiment and a further refinement of the computer model is indicated. (author)

  11. Heysham II/Torness AGR steam generator

    International Nuclear Information System (INIS)

    The AGR Steam Generators for Heysham II and Torness Power Stations have been installed at site and are being operated in the initial low temperature commissioning plant engineering tests. In this paper a description of the high pressure once-through steam generators together with layout arrangements, materials employed, operating parameters, plant operating conditions and constraints is given. An outline of the development of the design through thermo-hydraulic considerations, mechanical design, instrumentation to component testing is presented. Special features of the design directed to accommodate such requirements as seismic loadings, waterside static and dynamic stability, gas flow induced vibration, thermal expansions are described in detail. The fabrication facilities employed and techniques selected and developed for the manufacture and assembly of the heating surfaces are presented. These include welding processes, tube manipulation and heat treatment with details of the automation applied to the processes. Operating experience in the early commissioning plant engineering tests at Site is described with an emphasis on those tests which provide the final confirmation of the design prior to operation at full load. The paper concludes with a description of the outstanding commissioning activities up to raise power. (author)

  12. 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.)

  13. Corrosion product deposits and its removal from heat-transfering surfaces of LWGR type reactors

    International Nuclear Information System (INIS)

    Data on the corrosion product concentration, quantity and composition of corrosion deposits in the RBMK-1000 reactor circulation circuit are given. It is shown that the most radical method for deposit prevention is the circulation curcuit decontamination. Oxalic acid solution is suggested for these purposes with respect to the RBMK-1000 reactor. Hydrogen peroxide is introduced for fast dissolving of oxalate deposits of bivalent iron. Economic expedience of premaintenance decontamination of RBMK-1000 reactors is shown

  14. Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Martens, Frederick H. [Argonne National Laboratory; Jacobson, Norman H.

    1968-09-01

    This booklet discusses research reactors - reactors designed to provide a source of neutrons and/or gamma radiation for research, or to aid in the investigation of the effects of radiation on any type of material.

  15. 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.

  16. 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

  17. Condensation nuclear power plants with water-cooled graphite-moderated channel type reactors and advances in their development

    International Nuclear Information System (INIS)

    Consideration is being given to results of technical and economical investigations of advisability of increasing unit power by elevating steam generating capacity as a result of inserting numerous of stereotype sectional structural elements of the reactor with similar thermodynamic parameters. It is concluded that construction of power units of condensation nuclear power plants with water-cooled graphite-moderated channel type reactors of 2400-3200 MWe and higher unit power capacity represents the real method for sharp growth of efficiency and labour productivity in power industry. It can also provide the required increase of the rate of putting electrogenerating powers into operation

  18. Calculation of steam content in a draught section of a tank-type boiling water cooled reactor

    International Nuclear Information System (INIS)

    Structural and hydrodynamic features of a two-phase flow in a draught section of a tank-type boiling water cooled reactor are considered. A calculated model of the steady flow and methods for determining steam content and phase rate profiles under the maximum steam content at the section axis and at some distance from it are proposed. Steam content distribution by height quantitatively agrees with experimental data for the VK-50 reactor. Calculation technique allows one to obtain steam content and phase rate profiles at the section outlet

  19. Analytical study of nuclear-coupled density-wave instability in a natural circulation pressure tube type boiling water reactor

    International Nuclear Information System (INIS)

    An analytical model has been developed to study the nuclear-coupled density-wave instability in the Indian advanced heavy water reactor (AHWR) which is a natural circulation pressure tube type boiling water reactor. The model considers a point kinetics model for the neutron dynamics and a lumped parameter model for the fuel thermal dynamics along with the conservation equations of mass, momentum and energy and equation of state for the coolant. In addition, to study the effect of neutron interactions between different parts of the core, the model considers a coupled multipoint kinetics equation in place of simple point kinetics equation. Linear stability theory was applied to reveal the instability of in-phase and out-of-phase modes in the boiling channels of the AHWR. The results indicate that the stability behavior of the reactor is greatly influenced by the void reactivity coefficient, fuel time constant, radial power distribution and channel inlet orificing. The delayed neutrons were found to have a strong influence on the Type I and Type II instabilities observed at low and high channel powers, respectively. Also, it was found that the coupled multipoint kinetics model and the modal point kinetics model predict the same threshold power for out-of-phase instability if the coupling coefficient in the former model is half the eigen value separation between the fundamental and the first harmonic mode in the latter model. Decay ratio maps were predicted considering various operating parameters of the reactor, which are useful for its design. (orig.)

  20. 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.)

  1. Temperature dependence of swelling in Type 316 stainless steel irradiated in HFIR [High Flux Isotope Reactor

    International Nuclear Information System (INIS)

    The temperature dependence of swelling was investigated in solution-annealed (SA) and 20% cold-worked (CW) type 316 stainless steel irradiated to 30 dpa at 300 to 6000C in the High Flux Isotope Reactor (HFIR). At irradiation temperatures ≤ 4000C, a high concentration (2 to 4 x 1023 m-3) of small bubbles (1.5 to 4.5 nm diam) formed uniformly in the matrix. Swelling was low (0C. At 5000C, there was a mixture of bubbles and voids, but at 6000C, most of the cavities were voids. Maximum swelling (∼5%) occurred at 5000C. By contrast, cavities in 20% CW specimens were much smaller, with diameters of 6 and 9 nm at 500 and 6000C, respectively, suggesting that they were primarily bubbles. The cavity number density in the CW 316 at both 500 and 6000C (∼1 x 1022 m-3) was about one order of magnitude less than at 4000C. Swelling increased slightly as irradiation temperature increased, peaking at 6000C (0.3%). These results indicate that SA 316 swells more than CW 316 at 500 and 6000C, but both SA and CW 316 are resistant to void swelling in HFIR at 4000C and below to 30 dpa. 15 refs

  2. Magnesium production from Asian Abe-Gram dolomite in pidgeon-type reactor

    International Nuclear Information System (INIS)

    Ore mineral characterization and various experimental test work were carried out on Asian Abe-Garm dolomite, Qazvin province, Iran. The test work consisted of calcining, chemical characterization, LOI determination, and reduction tests on the calcined dolomite (doloma), using Semnan ferrosilicon. Calcining of dolomite sample was carried out at about 1400degreeC in order to remove the contained CO2, moisture, and other easily volatilised impurities. The doloma was milled, thoroughly mixed with 21percentSemnan ferrosilicon and briquetted in hand press applying 30 MPa pressure. The briquettes were heated at 1125-1150degreeC and 500 Pa in a Pidgeon-type tube reactor for 10-12 hours to extract the magnesium. Ferrosilicon addition, relative to doloma, was determined based on the chemical analysis of the two reactants using Mintek's Pyrosim software package. Magnesium extraction calculated as 77.97percentand Mg purity of 96.35percent. The level of major impurities in the produced magnesium crown is similar to those in the crude metal production.

  3. A practical methodology of radiological protection for the reduction of hot particles in BWR type reactors

    International Nuclear Information System (INIS)

    The purpose of this work, in general form, is to describe a practical method for reduction of hot particles generated as consequence of the operational activities of BWR nuclear reactors. This methodology provides a description of the localizations and/or probable activities of finding particles highly radioactive denominated hot particles. For this purpose it was developed a strategy based on the decontamination lineaments, as well as the manipulation, gathering, registration, contention, documentation, control and final disposition of the hot particles. In addition, some recommendations are reiterated and alternative, in order to gathering the hot particles in a dynamic way given to the activities of the personal occupationally exposed in highly radioactive areas. The structure of the methodology of hot particles is supported in the radiological controls based on the Code of Federal Regulation 10 CFR 20 as well as the applicable regulatory documents. It provides an idea based on administrative controls of radiological protection, in order to suggesting the responsibilities and necessary directing for the control of the hot particles required in nuclear plants of the BWR type. (author)

  4. Method of start-up of rotary plug sealing devices in FBR type reactors

    International Nuclear Information System (INIS)

    Purpose: To rapidly and safely start-up the rotary plug sealing device by controlling to eliminate the pressure difference in the pressures of gases exerting on the liquid surfaces in the inner and the outer cylinders of a sealing alloy vessel in the rotary plug of a FBR type reactor. Method: In a case where an abnormal state results in the pressure difference of gases exerted on the liquid surfaces in the inner and the outer cylinders of a vessel charged with sealing alloy in a rotary plug and the sealing valve for the back-up gas supply tube is rapidly closed to seal the sealing portion, the pressure in the gas supply tube is controlled so that the pressure difference in the gases exerted on the liquid surfaces in the inner and outer cylinders while closing the sealing valve. Then, after conforming that the pressure is controlled to a predetermined level at which the pressure difference can be regarded to be zero, the sealing valve is gradually opened while regulating the pressure in the gas supply tube so as to maintain the pressure difference to a predetermined level. This prevents the occurrence of external disturbances upon opening of the sealing valve and enables rapid and safety start-up for the rotary plug sealing device. (Moriyama, K.)

  5. 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)

  6. 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

  7. Numerical simulation of natural circulation in a geometry simulating a Babcock and Wilcox type nuclear reactor

    International Nuclear Information System (INIS)

    In this paper, the authors present the results of numerical calculations for natural circulation in the facility called Once-Through Integral System (OTIS) Test Facility simulating a Babcock and Wilcox type nuclear reactor. The OTIS test facility was constructed to represent the main features of a Babcok and Wilcox raised loop plant. The computer code adopted for the study is RETRAN-02. A small break LOCA is simulated, and a number of important physical variables are calculated and compared with test data. These variables are temperature, pressure, void fraction, mass flow rate and liquid level in the steam generator secondary side. The analysis conducted indicates that the RETRAN-02 calculated response agrees reasonably well with the measured system response. Figure 1 shows cold leg fluid temperature during a two-phase natural circulation transient. Complex phenomena such as flow oscillations due to void generation are calculated well with RETRAN-02. Hot and cold fluid mixing near the HPI injection port is also well represented using RETRAN-02. The results do indicate, however, the need to account for piping heat losses to accurately represent the detailed phenomena occurring in the hot leg

  8. On modular stellarator coils of the W VII-AS type with reactor dimensions

    International Nuclear Information System (INIS)

    Modular Stellarator coil systems of the W VII-AS type with reactor dimensions are investigated. They comprise 5 field periods with 6 to 18 coils within each. Two groups of configurations are compared. The first group is characterized by the major torus radius of R0 = 25.5 m and a distance between plasma and coils of ccdcc 1.8 m. The field on plasma axis amounts to B3 = 5.3 T. In the second group the geometric dimensions are decreased. The major torus radius amounts to R3 = 15.2 m, and a distance of about 1.2 m is chosen between plasma and coils. The field on plasma axis is enlarged to B3 = 7 T. The total stored magnetic energy of the coil systems and the maximum magnetic flux density at the coils are calculated. Furthermore, the magnetic forces and mechanical stresses are evaluated, and two different support concepts for the coils are investigated: A single coil support scheme (outer and lateral support elements plus elastic padding), as well as a novel scheme of mutual coil support

  9. 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)

  10. 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.)

  11. Swelling, mechanical properties, and microstructure of Type 316 stainless steel at fusion reactor damage levels

    International Nuclear Information System (INIS)

    Alloys such as AISI 316 stainless steel exhibit more swelling and larger decreases in ductility when irradiated to produce fusion reactor He and dpa levels than at fast reactor He and dpa levels. For T approx. 0C to ensure adequate ductility for long-term service

  12. 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)

  13. DT fusion ignition of LHD-type helical reactor by joule heating associated with magnetic axis shift

    International Nuclear Information System (INIS)

    A new concept to achieve current drive with magnetic axis shift, which is caused by vertical magnetic field coil current change in LHD-type magnetic configuration, is proposed. It is confirmed numerically that an LHD-type helical fusion reactor can be ignited by high-current Joule heating. MHD stability of the plasma current in a helical system is analyzed theoretically. Large plasma current that flows in the opposite direction of the helical coil current is MHD stable. Currents with a hollow current profile are more stable than those with a flat-top profile. The central peak current profile will be redistributed to the hollow current profile. A new concept involving the current-driven and current-less hybrid operational scenario of an LHD-type helical reactor is discussed. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Bakhtyar, S.; Iqbal, M.; Israr, M.; Pervez, S.; Salahuddin, A. [Nuclear Engineering Division, Pakistan Institute of Nuclear Science and Technology, P.O. Nilore, Islamabad (Pakistan)

    2004-07-01

    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)

  15. 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)

  16. 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)

  17. Physics concept on the constellation type fissile fuels and its application to the prospective Th-232U Reactor

    International Nuclear Information System (INIS)

    In contrast with the conventional nuclear reactor which usually fuelled with on single fissile nuclide, a constellation type fissile fuels reactor consists of a parent nuclide such as 232Th or 238U and its whole family of neutron generated daughter nuclides. All of them are regarded as fissile fuels but of quite different fission ability. The concentration of each daughter nuclide is determined by its saturate concentration ratio with the parent nuclide. In such fuel system, the whole fuel consumed by neutron reaction almost completely results in fission products. In this article, some properties of such fuel system, determination of the saturate concentration of each daughter nuclide and applicability to Th-233U fueled reactor will be discussed. 3 refs., 1 tab., 2 figs

  18. 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}.

  19. 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)

  20. Utilization of a HTR type reactor as a heat source for the processing of pyrobituminous shale by the Petrosix method

    International Nuclear Information System (INIS)

    Some thermodynamics aspects of a system resulting from the coupling of a THTR nuclear power plant type (Thorium High Temperature Reactor) and a commercial shale oil processing plant are studied. The coupling is basically characterized by the application of all available energy from the nuclear reactor to the shale oil processing. The nuclear reactor employed is a PR-3000, with 2980,8 MW sub(t), developed in the Federal Republic of Germany for process heat applications (coal and steam reforming to produce reducers and products similar to the derivates of petroleum). The commercial shale oil plant considered (U.C.X.) uses the Petrosix process developed by the Superintendencia da Industrializacao do Xisto (S.I.X.) of Petrobras. Some flow diagrams are proposed for the coupling between the basic cycle of PR-3000 reactor with hot gas cycle of U.C.X. For a pre-determined flow diagram and boundary conditions, the thermodynamic parameters that lead to a maximum efficiency of the system are established. Also the main steam cycle parameters of PR-3000 reactor are determined, including those for the main heat exchanger, whose data are similar to the corresponding steam and coal reforming system used in process heat application of the PR-3000

  1. Physical characteristics of MBIR reactor with alternative types of oxide fuel

    International Nuclear Information System (INIS)

    Various variants of fuel compositions for research reactor MBIR: oxide uranium fuel (on the base of uranium-235), mixed oxide fuel pellets (depleted uranium dioxide and plutonium dioxide), combined vibroMOX-fuel are under consideration. The calculation results of physical characteristics of MBIR reactor core when using fuel compositions studied are given, their advantages and disadvantages are discussed. It is shown that the only acceptable variant is standard vibro-packed fuel from granulates of plutonium dioxide (38.8%) and depleted uranium dioxide. It is pointed out that the quality of neutron flux (its ability to radiation damage of materials) in MBIR is essentially higher than in power reactors

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

    Directory of Open Access Journals (Sweden)

    Devarshi Patel

    2014-05-01

    Full Text Available 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 requirement. For injecting the pellet we use the pellet injection system. Pellet injection system (PIS is the fuel injection system of the fusion reactor.

  3. 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.

  4. Operation of NPP with the WWER type reactor during lifetime prolongation using the power effect

    International Nuclear Information System (INIS)

    A method is described for improving fuel cycle at the expense of prolongation of reactor lifetime permits to have higher loading in the day time than at night. The reactivity balance is considered for WWER reactor in the regime of reactor life prolongation at continuous, stepwise and sinusoidal power changes. Calcu-- lation of reactivity gain at a sinusoidal change of loading is given. The use of sinusoidal law of power change in the reactor life time prolongation regime at ''Reinsberg'' and ''Greisfwald'' NPPsub(s) has revealed that the calculated reactivity gains may be attained in practice, in that case 1% loading variations of rated power do not affect the results. The advantage of sinusoidal regime, as compared with the other methods, is that the peaks correspond to the maximum loading and that one can exercise control of a unit at a constant pressure in the steam generator, which improves the temperature process

  5. Reactor feedwater flow rate control device in a BWR type power plant

    International Nuclear Information System (INIS)

    Purpose: To control reactor feedwater level stationarily in a case where steams are released from relief valves. Constitution: Flow rate of steams discharged from relief valves is determined by the reactor pressure and the number of opened relief valves, and the value is added to the main steam flow rate detected by a steam flow rate detector. Then, based on the sum and the feedwater flow rate, a flow rate deviation signal is obtained. Thus, in a case where steams are discharged from relief valves, the flow rate of steams discharged from the reactor can be estimated accurately with no negative errors, and reduction in the reactor water level can thereby be prevented. (Kamimura, M.)

  6. AGR-3/4 Data Qualification Report for ATR Cycles 151A, 151B, 152A, 152B, 154A, and 154B

    Energy Technology Data Exchange (ETDEWEB)

    Binh T. Pham

    2014-02-01

    This data report provides the qualification status of Advanced Gas Reactor-3/4 (AGR-3/4) fuel irradiation experimental data from Advanced Test Reactor (ATR) Cycles 151A, 151B, 152A, 152B, 154A, and 154B, as recorded in the Nuclear Data Management and Analysis System (NDMAS). Of these cycles, ATR Cycle 152A is a low power cycle that occurred when the ATR core was briefly at low power. The irradiation data are not used for physics and thermal calculation, but the qualification status of these cycle data is still covered in this report. On the other hand, during ATR Cycles 153A (unplanned Outage cycle) and 153B (Power Axial Locator Mechanism [PALM] cycle), the AGR-3/4 was pulled out from the ATR core and stored in the canal to avoid being overheated. Therefore, qualification of the AGR-3/4 irradiation data from these 2 cycles was excluded in this report. By the end of ATR Cycle 154B, AGR-3/4 was irradiated for a total of 264.1 effective full power days. The AGR-3/4 data streams addressed in this report include thermocouple (TC) temperatures, sweep gas data (flow rates, pressure, and moisture content), and Fission Product Monitoring System (FPMS) data (release rates and release-to-birth rate ratios [R/Bs]) for each of the twelve capsules in the AGR-3/4 experiment. The final data qualification status for these data streams is determined by a Data Review Committee (DRC) composed of AGR technical leads, Sitewide Quality Assurance (QA), and NDMAS analysts. The DRC convened on February 12, 2014, reviewed the data acquisition process, and considered whether the data met the requirements for data collection as specified in QA-approved Very High Temperature Reactor (VHTR) Technology Development Office (TDO) data collection plans. The DRC also examined the results of NDMAS data testing and statistical analyses, and confirmed the qualification status of the data as given in this report.

  7. Operational and reliability experience with reactor instrumentation

    International Nuclear Information System (INIS)

    In the last 15 years the CEGB has experienced progressive plant development, integration and changes in operating regime through nine nuclear (gas-cooled reactor) power stations with corresponding instrumentation advances leading towards more refined centralized control. Operation and reliability experience with reactor instrumentation is reported in this paper with reference to the progressive changes related to the early magnox, late magnox and AGR periods. Data on instrumentation reliability in terms of reactor forced outages are presented and show that the instrumentation contributions to loss of generating plant availability are small. Reactor safety circuits, neutron flux and temperature measurements, gas analysis and vibration monitoring are discussed. In reviewing the reactor instrumentation the emphasis is on reporting recent experience, particularly on AGR equipment, but overall performance and changes to magnox equipment are included so that some appreciation can be obtained of instrumentation requirements with respect to plant lifetimes. (author)

  8. Time constants and feedback transfer functions of EBR-II [Experimental Breeder Reactor] subassembly types

    International Nuclear Information System (INIS)

    Time constants, feedback reactivity transfer functions and power coefficients are calculated for stereotypical subassemblies in the EBR-II reactor. These quantities are calculated from nodal reactivities obtained from a reactor kinetic code analysis for a step change in power. Due to the multiplicity of eigenvalues, there are several time constants for each nodal position in a subassembly. Compared with these calculated values are analytically derived values for the initial node of a given channel

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

    OpenAIRE

    Hou-jun Gong; Xing-tuan Yang; Yan-ping Huang; Sheng-yao Jiang

    2015-01-01

    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 ...

  10. 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)

  11. 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

  12. Large-scale surface dielectric barrier discharge type reactor : effect of the electric wind on the conversion effectiveness

    Energy Technology Data Exchange (ETDEWEB)

    Jolibois, J. [Univ. de Poitiers, Poitiers (France). Centre national de la recherche scientifique, Laboratoire de Catalyse en Chimie Organique; Poitiers Univ., Futuroscope Chasseneuil Cedex (France). Centre national de la recherche scientifique, Inst. Pprime; Zouzou, N.; Moreau, E. [Poitiers Univ., Futuroscope Chasseneuil Cedex (France). Centre national de la recherche scientifique, Inst. Pprime; Tatibouet, J.M. [Univ. de Poitiers, Poitiers (France). Centre national de la recherche scientifique, Laboratoire de Catalyse en Chimie Organique

    2010-07-01

    Non-thermal plasma (NTP) techniques offer an innovative approach for air pollution reduction. Most studies in NTP techniques use volumetric discharge reactors with small dimensions and low flow rates at laboratory scale. The objective of this study was to develop an air pollution control plasma reactor at industrial scale with surface discharge. Propene (C{sub 3}H{sub 6}) was oxidized at high flow rates in a large-scale plasma reactor based on surface dielectric barrier discharge (DBD). Three different configurations of surface discharges were tested with 15 ppm of C{sub 3}H{sub 6} in air at ambient temperature for a flow rate of 50 m{sup 3} per hour. The properties of these different surface discharges were analyzed using chemical measurements and 3 component particle image velocimetry (PIV) measurements. PIV measurements were used characterize the effect of the electric wind on the polluted gas airflow inside the reactor and to explain the differences of effectiveness of the three tested plasma generators. For the three plasma generators, a propene oxidation of up to 45 percent was obtained at one J per liter. The electric wind produced by the surface discharge resulted in the formation of vortices inside the plasma reactor. This electric wind can increase gas mixing inside the plasma reactor and therefore plays a key role in conversion efficiency. It was concluded that the electric wind produced by surface discharges enables the use of this type of discharge for VOC elimination at high flow rate, with the same effectiveness of volumetric discharges. 5 refs., 10 figs.

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

    International Nuclear Information System (INIS)

    Highlights: ► Pyrolysis of aseptic packages was carried out in a laboratory flow reactor. ► Distribution of tetrapak into the product yields was obtained. ► Composition of the pyrolysis products was estimated. ► Secondary thermal and catalytic decomposition of tars was studied. ► Two types of catalysts (dolomite and red clay marked AFRC) were used. - Abstract: 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

  14. Modelling of the WWER-440 type reactor emergency conditions associated with a pipeline breakage while training equipment designing

    International Nuclear Information System (INIS)

    A theoretical investigation of the emergency conditions connected with the rupture of 500 mm dia main circulating pipe of the primary circuit of the WWER-440 type reactor nuclear power plant has been carried out. The calculations have been performed using the ''''Minsk-32'' digital computer. A method is described of data presentation in the form suitable for modelling of transients in conformity with the training equipment

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

    Energy Technology Data Exchange (ETDEWEB)

    Haydary, J., E-mail: juma.haydary@stuba.sk [Institute of Chemical and Environmental Engineering, Faculty of Chemical and Food Technology, Slovak University of Technology, Radlinského 9, 812 37 Bratislava (Slovakia); Susa, D.; Dudáš, J. [Institute of Chemical and Environmental Engineering, Faculty of Chemical and Food Technology, Slovak University of Technology, Radlinského 9, 812 37 Bratislava (Slovakia)

    2013-05-15

    Highlights: ► Pyrolysis of aseptic packages was carried out in a laboratory flow reactor. ► Distribution of tetrapak into the product yields was obtained. ► Composition of the pyrolysis products was estimated. ► Secondary thermal and catalytic decomposition of tars was studied. ► Two types of catalysts (dolomite and red clay marked AFRC) were used. - Abstract: 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 H{sub 2}, CO, CH{sub 4}, CO{sub 2} 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.

  16. 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

  17. Study of the thermal state of fuel elements of the VVER-1000 light-water type reactors

    International Nuclear Information System (INIS)

    Fuel assemblies of the VVER-1000 type reactors have been studied during 4500 hours with the ''no contact'' method of the thermal. The results of the studies show the instability of the fuel assemblies under irradiation. It has been pointed out that there is a sudden variation of the fuel temperature just after the accidental protection start-up. The fuel assembly, after a prolonged use, tends to a state in which the thermal resistance of the fuel can space is near zero

  18. New insight into transmembrane topology of Staphylococcus aureus histidine kinase AgrC.

    Science.gov (United States)

    Wang, Lina; Quan, Chunshan; Xiong, Wen; Qu, Xiaojing; Fan, Shengdi; Hu, Wenzhong

    2014-03-01

    Staphylococcus aureus accessory gene regulator (agr) locus controls the expression of virulence factors through a classical two-component signal transduction system that consists of a receptor histidine protein kinase AgrC and a cytoplasmic response regulator AgrA. An autoinducing peptide (AIP) encoded by agr locus activates AgrC, which transduces extracellular signals into the cytoplasm. Despite extensive investigations to identify AgrC-AIP interaction sites, precise signal recognition mechanisms remain unknown. This study aims to clarify the membrane topology of AgrC by applying the green fluorescent protein (GFP) fusion technique and the substituted cysteine accessibility method (SCAM). However, our findings were inconsistent with profile obtained previously by alkaline phosphatase. We report the topology of AgrC shows seven transmembrane segments, a periplasmic N-terminus, and a cytoplasmic C-terminus. PMID:24361366

  19. AGROCOOP Tienda Online Cooperativas Agrícolas

    OpenAIRE

    Segura Rama, José Ignacio

    2016-01-01

    El presente Trabajo Fin de Master versa sobre la elaboración, creación, programación, diseño y puesta en marcha de una tienda online AVANZADA para cooperativas agrícolas, queremos mostrar como poder crear una tienda e-comerce para cooperativas agrícolas de nuestro País. Con ello queremos así abrir el mercado de estas, ademas de mostrar como usar un gran compendio de tecnologías. Vamos a agrupar y poner en común, un conjunto amplio de distintas técnicas de programación web, que ...

  20. Producción agrícola controlada

    OpenAIRE

    Abraham Rojano A.; Raquel Salazar M.; Álvaro Llamas G.

    2004-01-01

    A diferencia de la producción de bienes industriales donde se puede trabajar día y noche y en diferentes lugares al mismo tiempo, la producción agrícola tiene componentes físicos y químicos en interacción con componentes vivos. Los factores físicos y químicos son factibles de trabajar con enfoques industriales de acuerdo a las reglas básicas del control; sin embargo, el componente vivo de la producción agrícola esta condicionado hasta la fecha por un proceso ineludible: la foto...

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

    International Nuclear Information System (INIS)

    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 H2O: 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, 13C-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

  2. Technical report: technical development on the silicide plate-type fuel experiment at nuclear safety research reactor

    International Nuclear Information System (INIS)

    According to a reduction of fuel enrichment from 45 w/o 235U to 20 w/o, an aluminide plate-type fuel used currently in the domestic research and material testing reactors will be replaced by a silicide plate-type one. One of the major concern arisen from this alternation is to understand the fuel behavior under simulated reactivity initiated accident (RIA) conditions, this is strongly necessary from the safety and licensing point of view. The in-core RIA experiments are, therefore, carried out at Nuclear Safety Research Reactor (NSRR) in Japan Atomic Energy Research Institute (JAERI). The silicide plate-type fuel consisted of the ternary alloy of U-Al-Si as a meat with uranium density up to 4.8 g/cm3 having thickness by 0.51 mm and the binary alloy of Al-3%Mg as a cladding by thickness of 0.38 mm. Comparison of the physical properties of this metallic plate fuel with the UO2-zircaloy fuel rod used conventionally in commercial light water reactors shows that the heat conductivity of the former is of the order of about 13 times greater than the latter, however the melting temperature is only one-half (1570degC). Prior to in-core RIA experiments, there were some difficulties lay in our technical path. This report summarized the technical achievements obtained through our four years work. (J.P.N.)

  3. 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.

  4. 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

  5. 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: <0.1 wt%, O: 46.8 wt% and H2O: 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, 13C-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.

  6. 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

  7. Heterogeneous catalysis in the different reactor types on the examples of ethyl benzene to styrene, methane dehydroaromatization and propylene carbonate, methanol transesterification

    OpenAIRE

    Mousko, Dimitri

    2009-01-01

    The dissertation is based on the three above mentioned projects. Different types of reactors for heterogeneous catalysis were applied in this work – fixed and fluidized bed rectors as well as a riser reactor. Many parameters as e.g. contact times, temperature control, catalyst regeneration procedure etc. can be varied and adjusted to the special reaction requirements. Correct reactor choice is the decisive factor for the optimal reaction performing. One of the objectives of this work is to fi...

  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. Operating characteristic analysis of a 400 mH class HTS DC reactor in connection with a laboratory scale LCC type HVDC system

    Science.gov (United States)

    Kim, Sung-Kyu; Kim, Kwangmin; Park, Minwon; Yu, In-Keun; Lee, Sangjin

    2015-11-01

    High temperature superconducting (HTS) devices are being developed due to their advantages. Most line commutated converter based high voltage direct current (HVDC) transmission systems for long-distance transmission require large inductance of DC reactor; however, generally, copper-based reactors cause a lot of electrical losses during the system operation. This is driving researchers to develop a new type of DC reactor using HTS wire. The authors have developed a 400 mH class HTS DC reactor and a laboratory scale test-bed for line-commutated converter type HVDC system and applied the HTS DC reactor to the HVDC system to investigate their operating characteristics. The 400 mH class HTS DC reactor is designed using a toroid type magnet. The HVDC system is designed in the form of a mono-pole system with thyristor-based 12-pulse power converters. In this paper, the investigation results of the HTS DC reactor in connection with the HVDC system are described. The operating characteristics of the HTS DC reactor are analyzed under various operating conditions of the system. Through the results, applicability of an HTS DC reactor in an HVDC system is discussed in detail.

  10. Water quality control method and device in BWR type reactor primary coolant system

    International Nuclear Information System (INIS)

    A specimen is sampled through a sampling line disposed to the upstream of a reactor cleanup device, to measure the concentration of iron and copper in a reactor water by using a concentration measuring device. Then, the ratio of Fe/Cu concentration is calculated based on the concentration of iron and copper in the measured reactor water. If the value is smaller than 2, the ratio of Fe/Cu concentration in the reactor water is controlled to greater than 2 by increasing the amount of iron in feedwater injected from an iron forming device to the side of the exit of an condensator desalter. With such procedures, copper deposited on the surface of fuel rods can be fixed to a stable chemical form of ferrite oxide. Accordingly, since the valence of Cu ions in fuel rod deposits can be kept identical with that of the valence of Cu ions in reactor water, accelerated corrosion of zirconium, as the main constituent element of the fuel cladding tube, can be prevented. With such procedures, integrity of the fuel cladding tube can be maintained throughout the use of the fuel. (I.N.)

  11. Drop impact analysis of plate-type fuel assembly in research reactor

    International Nuclear Information System (INIS)

    In this research, a drop impact analysis of a fuel assembly in a research reactor is carried out to determine whether the fuel plate integrity is maintained in a drop accident. A fuel assembly drop accident is classified based on where the accident occurs, i.e., inside or outside the reactor, since each occasion results in a different impact load on the fuel assembly. An analysis procedure suitable for each drop situation is systematically established. For an accident occurring outside the reactor, the direct impact of a fuel assembly on the pool bottom is analyzed using implicit and explicit approaches. The effects of the key parameters, such as the impact velocity and structural damping ratios, are also studied. For an accident occurring inside the reactor, the falling fuel assembly may first hit the fixing bar at the upper part of the standing fuel assembly. To confirm the fuel plate integrity, a fracture of the fixing bar should be investigated, since the fixing bar plays a role in protecting the fuel plate from the external impact force. Through such an analysis, the suitability of an impact analysis procedure associated with the drop situation in the research reactor is shown.

  12. Development of techniques to dispose of the Windscale AGR heat exchangers

    International Nuclear Information System (INIS)

    In a gas-cooled nuclear power plant the gas side of the heat exchanger tubes becomes contaminated with radioactive deposits carried from the reactor in the coolant stream. In order to dispose of the heat exchangers in the safest and most cost-effective way during plant decommissioning, the deposits have to be removed. In situ chemical decontamination is considered to be the only viable method. This paper describes the research and development of chemical decontamination methods for the Windscale AGR heat exchangers, and the testing of a selected method on an in situ superheater. The research involved characterization of tube corrosion and radioactivity deposits, laboratory testing of chemical reagents on actual tube samples, and the provision and operation of a plant to apply the selected reagent. Disposal of radioactive effluent is an important consideration in chemical decontamination and in the present case was the major factor in determining the process

  13. Steam generator materials constraints in UK design gas-cooled reactors

    International Nuclear Information System (INIS)

    A widely reported problem with Magnox-type reactors was the oxidation of carbon steel components in gas circuits and steam generators. The effects of temperature, pressure, gas composition and steel composition on oxidation kinetics have been determined, thus allowing the probabilities of failure of critical components to be predicted for a given set of operating conditions. This risk analysis, coupled with regular inspection of reactor and boiler internals, has allowed continued operation of all U.K. Magnox plant. The Advanced Gas Cooled Reactor (AGR) is a direct development of the Magnox design. The first four AGRs commenced operation in 1976, at Hinkley Point 'B' and at Hunterston 'B'. All known materials problems with the steam generators have been diagnosed and solved by the development of appropriate operational strategies, together with minor plant modifications. Materials constraints no longer impose any restrictions to full load performance from the steam generators throughout the predicted life of the plant. Problems discussed in detail are: 1. oxidation of the 9 Cr - 1 Mo superheater. 2. Stress corrosion of the austenitic superheater. 3. Creep of the transition joints between the 9 Cr - 1 Mo and austenitic sections. With the 9 Cr - 1 Mo oxidation maximum temperature restriction virtually removed and creep constraints properly quantified, boiler operation in now favourably placed. Stress corrosion research has allowed the risk of tube failure to be related to time, temperature, stress and chemistry. As a result, the rigorous 'no wetting' policy has been relaxed for the normally high quality AGR feedwater, and the superheat margin has been reduced to 23 deg. C. This has increased the size of the operating window and reduced the number of expensive, and potentially harmful, plant trips. (author)

  14. Tightening unit EZ 250 for VVER 1000 type reactor pressure vessel head flange joints

    International Nuclear Information System (INIS)

    The programme of flange joints tightening by seals made of expanded graphite for VVER 440 and VVER 1000 reactor head flange joints is highlighted, and tightening units of row EZ 650 and EZ 650 TK and KNI for VVER 440 reactor head flange joints and EZ 250 tightening unit for VVER 1000 reactor head flange joints are described in detail. The main advantages of electronically controlled tightening units include: Precise and uniform compression of the gasket during the tightening procedure; Automated solution to the graphite relaxing problem during tightening; Possibility of diagnosis of the thread status of the joints being tightened; Alleviation of operator's tough work; Shorter time of tensioning associated with a lower collective doses; Quick preparation of tightening procedure report from the data measured; Calibration renewal is possible in advance at time of unit storage without the need to place it on a real joint. (P.A.)

  15. 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

  16. Applied methods for mitigation of damage by stress corrosion in BWR type reactors

    International Nuclear Information System (INIS)

    The Boiling Water nuclear Reactors (BWR) have presented stress corrosion problems, mainly in components and pipes of the primary system, provoking negative impacts in the performance of energy generator plants, as well as the increasing in the radiation exposure to personnel involucred. This problem has caused development of research programs, which are guided to find solution alternatives for the phenomena control. Among results of greater relevance the control for the reactor water chemistry stands out particularly in the impurities concentration and oxidation of radiolysis products; as well as the supervision in the materials selection and the stresses levels reduction. The present work presents the methods which can be applied to diminish the problems of stress corrosion in BWR reactors. (Author)

  17. Basic CFD investigation of decay heat removal in a pool type research reactor

    International Nuclear Information System (INIS)

    Safety is one of the most important and desirable characteristic in a nuclear plant. Natural circulation cooling systems are noted for providing passive safety. These systems can be used as mechanism for removing the residual heat from the reactor, or even as the main cooling system for heated sections, such as the core. In this work, a computational fluid-dynamics (CFD) code is used to simulate the process of natural circulation in an open pool research reactor after its shutdown. The physical model studied is similar to the Open Pool Australian Light water reactor (OPAL), and contains the core, cooling pool, reflecting tank, circulation pipes and chimney. For best computing performance, the core region was modeled as a porous media, where the parameters were obtained from a separately detailed CFD analysis. (author)

  18. Studies on a Stellarator reactor of the Helias type: The modular coil system

    International Nuclear Information System (INIS)

    Helias Stellarator Reactors (HSR) are considered, focussing on the superconducting modular coil system which generates the magnetic field, aiming to clarify critical issues of such systems. The development of the coil system is presented and the properties of the vacuum magnetic field are discussed. Electromagnetic forces and the resulting mechanical stresses and strains inside the coils and the surrounding structure are calculated. Parameter studies are made varying the major radius R0 between 18 m and 24 m in order to investigate the engineering parameters for the superconducting coil system. The total mass and the fusion power output of HSR are compared with values evaluated for tokamak reactors. (orig.). 36 figs

  19. Model for analysis of hydrodynamic processes at the BHWR type reactors

    International Nuclear Information System (INIS)

    Two-phase flow of the coolant is defined by mass, energy and flow rate conservation laws. These laws define time dependence of the total steam quantity in the reactor which influences the reactivity changes. Changes of the fluid flow in the coolant system is determined as well. In fact the hydrodynamic model describes processes in the coolant system as a function of operating pressure, boiling limit and reactor thermal power. This paper deals with estimated presumptions in defining the mathematical model of hydrodynamic processes

  20. Computer software development of the in-core monitoring system for the WWER type reactor

    International Nuclear Information System (INIS)

    All of this has been led to the necessity of the requirement increase to the maintenance of the safety project limit guaranteeing during the exploitation of the fuel loading of the WWER reactor. It also causes the need of the functional possibility widening and computer software development of the in-core monitoring system for the WWER reactors. provided ways and fulfilment peculiarity of the corresponding work for the NPP Kola (WWER-440) and NPP Rostov (WWER-1000) are briefly discussed in the report (Authors)